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LONG ISl_AND' LIGHTING COMPANY l
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175 EAST OL D COUNTR Y R O A D . H i c K S V I L L E, N E W YORK 11001 ANTHUN C.SUODEN l
April 18, 1969 G r., Peter A. Morris, Director Division of Reactor Licensing U.S. Atomic Energy Commission 4915 St. Elmo Avenue ". -
Bethesda, Maryland 20014 Amendment No. 5 to License Application Shoreham Nuclear Power Station Unit 1 Docket No. 50-322 1
4
Dear Dr. Morris:
Pursuant to the Atomic Energy Act of 1954 as amended and the Commission's Rules and Regulations issued thereunder, Long Island Lighting Company hereby supplements its license application filed May 15, 1968 by providing responses to comments contained in the enclosure to your letter of January 19, 1969 .
Very truly yours, Oc iD A. C. Sugden S,'[ (p, ,,
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l SNPS-1 i TABLE OF CONTENTS
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SECTION PAGE ;
1.0 GENERAL. AS-1
.2.0 SITE & ENVIRONS AS-18 4
3.0 REACTOR CORE AS-36 .
4.0 -
REACTOR' COOLANT SYSTEM A5-38 5.0 CC;TAINMENT SYSTEMS A5-58 6.0 ENGINEERED SAFETY FEATURES A5-84 i
7.0 INSTRUMENTATION & CONTROL A5-110 6.O ELECTRICAL POWER SYSTEM AS-142-9.0 RADIOACTIVE WASTE SYSTEMS A5-148 10.0 AUXILIARY SYSTEMS AS-162 11.0 PuriER CONVERSION SYSTEM A5-163 12.0 STRUCTURES AND SHIELDING AS-165 13.O STATION OPERATIONS A5-206 14.0 ACCIDENT ANALYSIS A5-214 9
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SNPS-1 E
LIST OF TABLES TABLE NO. PAGE 6.2-1 AS-93 6.4-1 AS-96 9.7-1 .
AS-161 f
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I.IST OF FIGURES j FIGURE'NO. TITLE 4
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.I 1.3d-1 l Radioactive Waste Control Systems - Reactor Gas and Liquid Paths - Schematic.
-2.6A Maximum Tide Levels at New London.
2.9-1 Response Spectra - Operational Basis Earthquake.
2.9-2 . Response Spectra - Design Basis Earthquake.
D.*t Reactor Coolant Pressure Boundary.
5.17.C-1 Long Term Hydrogen Source Radiolytic Decomposition!
5.17.d-1 Flammability of 02 - H 2- N 2- Dry Mixtures.
6.2-1 Loss of Coolant Accident - RHR and Core Spray .
System Performance During ' Reactor Vesse. Blowdc - l 6.2-2 Loss of Coolant Accident - Estimated RER Syrtem Characteristics During Containment Spray / Cooling Operations Mode.
6.2-3 Loss of Coolant Accident - Primary Containment Temperature (1 Core Spray Pump & 1 RRRS Pump) .
6.2-4 Loss of Coolant Accident - NPSH Available for Core Spray and RHRS Pumps. (1 Core Spray Pmp-
& RHRS Pump)
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7.6-1 Control Rod Position Display Locations. 2 7.6-2 Reactor Control Board.
7.6-3 Reactor Control Board.
i 7.6-4 Typical Process Computer Printout. j 7.6-5 Input Signals to Four Rod Displays. ;
7.6-6 Control Rod 1osition Indication System.
7.6-7 Control Rod Drift Circuitry Preliminary Schematic f i
7.16-1 Reactor Recirculation Flow Reference Signals !
for Rod Block Monitoring System. j i
9.7-1 Station Structures. )
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' FIGURE NO.
. TITLE 9.8-1
(. Station Ventilation Systems and Associated Radiation Monitoring Provisions.
11.1A Electro-Hydraulic Diagram.
Control Systems - Block
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SNPS-1 EXHIBITS-EXHIBIT A
(Refers to Comment No. 2.9A)
B (Refers to Comment No. 2.7)
C (Refers to Comment No. 9.7)
D (Refers to Comment No. 4.1A,d) d (Refers to Comment No. 5.14) l I,
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LIST OF FIGURES FOR EXHIBITS EXHIBIT A FIGURE NO TITLE Soil Profile for Shoreham Site.
2 Soil Properties.
3 Effective Velocity and Damping as Function of Strain.
4 Acceleration Vs. Time at Rock and at Ground Surface: El Centro Input.
5 Velocity vs. Time at Rock and at Ground Surface:
El Centro Inpu'.. .
6 Acceleration Response Spectra at Rock and at Ground Surface El Centro Input.
I 7 Velocity Response Spectra at Rock and at Ground Surface: El Centro Input.
8 Ratio of Response Spectra for 5% Damping; El Centro Input.
9 Shear Input.
Stress Vs. Time for two layers: El Centro 10 Shear Stress Vs. Depth: El Centro Input.
11 l Calculation El 32.5 ft.
of Theoretical Shear Stress at 12 Response Spectra at Surface for 5% Damping: All Input.
A-1 Acceleration Surface: TafVs. t Input. Time at Rock and at. Ground A-2 Velocity Vs. Time at Rock and at Ground Surface: Taft Input.
A-3 Ratio Input.
of Response Spectra for 5% Damping: Taft f
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'SNPS-f EXHIBIT A- TITLE FIGURE NO A-4 Shear Stress vs. Timer Taf t Input.
A-5 Shear Stress vs. Depth: Taft Input.
A-6 Acceleration vs. Time at Rock and at-Grcund Surface: Helena Input.
A-7 V.:'ocity Vs.. Time at Rock and. at Ground
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Surface: Helena Input.
A-8 Ratio'of Response Spectra for 55 Damping: Helena Input.
A-9 Shear Stress vs. Time: Helena Input.
A-10 Shear Stress Vs. Depth: Helena Input.
A-11' Acceleration Vs. Time at Rock and Ground Surface: Artificial Input.
A-12 Velocitj' Vs. Time at Rock and Ground Surface: Artificial Input.
'A-13 Ratio Response ' Spectra for 5% i Damping: Artificial Input.
i A-14 Shear Stress Vs. Time Artificial Input. I A-15 Shear Stress vs. Depth: Artificial Input. l A-16 Acceleration Vs. Time at Rock and at Ground Surface Golden Gate Input.
A-17 Velocity Vs. Time at Rock and at Ground Surface: Golden Gate Input.
A-18 Ratio of Response Spectra for 55 Damping: Golden Gate Input.
A-19 Shear Stress vs. Time: Golden Gate Input.
A-20 Shear Stress Vs. Depth: Golden Gate Input i
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SNSP-1 I EXHIBIT B TITLE FIGURE NO B-1 through b-31 Boring Logs EXHIBIT C FIGURE NO C-1 Smoke Plume Section for Wind Approach Angle of 00 C-2 * " " " " " " "
300 C-3 * " " " " " * "
450 C-4 * ." " " " " " "
900 C-5 " " " " " " " "
1800 C-6 * * " " " " " "
2700 C-7 3150 C-8 " " " " " " " "
3300 i C-9 Smoke Plume Sections for Wir.d Speed of 1.0 m/see from Critical Wind Direction. !
C-10 Smoke Plume Sections for Wind Speed of 2.5 m/sec from Critical Wind Direction.
C-11 Smoke Plume Sections for Wind Speed of 4.0 m/sec from Critical Wind Direction.
C-12 Smoke Plume Sections for Wind Speed of 5.5 m/sec from Critical Wind Direction.
C-13 Smoke Plume Sections for Wind Speed of 10.0 m/sec
. from Critical Wind Direction.
C-14 Pressure Tap Locatirans for Turbine Building. i l
C-15 Pressure Tap Locations for Reactor Buildina. '
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- EXilIBIT B TITLE
-FIGURE NO B-1 through b-31 Boring Logs EXHIBIT C FIGURE
_N O_
C-1 Smoke Plume Section for Wind Approach Angle of 00 C-2 " " " " " " " "
300 C-3 " " " " " " " "
450 C-4 " " " " " " " "
90 0 ,
i C-5 " " " " " " "
1800 C-6 " " " " " " " "
2700 C-7 " " " " " " " "
3150 l C-8 * * * " " " " "
3300 C-9 Smoke Plume Sections for Wind Speed of 1.0 m/sec from Critical Wind Direction.
C-10 Smoke Plume Sections for Wind Speed of 2.5 m/see from Critical Wind Direction.
C-11 Smoke Plume Sections for Wind Speed of 4.0 m/sec from Critical Wind Direction.
C-12 Smoke Plume Sections for Wind Speed of 5.5 m/sec from Critical Wind Direction.
C-13 Smoke Plume Sections for Wind Speed of 10.0 m/see from Critical Wind Direction.
C-14 Pressure Tap Locations for Turbine Building.
C-15 Pressure Tap Locations for Reactor Building. q h
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1 SNPS-1 EXHIBIT D TITLE FIGURE NO D-1 Simplified Interpetration of Drop - Weight Test.
D-2 Fracture Toughness of A212 B Plate at NDT 200 F.
D-3 Thickness Toughness Relationship.
D-4 Modified Criterion for Brittle Fracture Prevention.
D-5 Effect of Local Increase in Thickness.
D-6 Prediction of Fracture Toughness from Approximate
] General Relationships.
D-7 Transition Temperatures Vs. Thickness Comp' rison of Theory and Date.
D-8 Comparison of Theoretical Relationship to Code Rule for Impact Test Exemption. !
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SNPS L EXHIBIT E !
FIGURE.
NO TITLE PAGE 1
Pressure with Suppression Multiple Vents Test Facility 2 I 2
Arrangement of Test Facility 3 3
Suppression Pool Baffle Arrangement for Quarter Scale Tests 4 4
Reactor Vessel Pressure Traces 6 5
Suppression Chamber Pressure Traces 8 6
Results of Quarter Scale Test #10H 11 7
Results of Quarter Scale Test #23H 12 8
Results of Quarter Scale Test #36 13 s
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- 1. O GEIGlRAL
. COMMENT 1.1: In your application, and in recent' technical meetings ' you have indicated that LIIro will take a significant pset in the design of the Shoreham Station. Please define participation in the design the extent of' LIIEC's effort and provide additional information including on - the LILCO organization, qualifications of key personnel. To what *ztent will specialized training be provided to the LIIro engineering staff to assist it in its design ;
functions during plant operation? function and its engineer ,
RESPONSEt ]
I-6.1 The response to this comment will be fond in Section i Safety Analysis Report (SNPS PSAR).of Amendment 4, Shoreham Nuc !
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AS-1
SNPS-1 COMMENT 1.2: Considering your definition of class I structures and equipment. for seismic design (pg XII-2-6) ,
please discuss your reasons for not designating the following structures or equipment class I.
- a. The steam piping beyond the first isolation valve external :to .the drywell and the main turbine.
- b. Reactor building above the crane rails.
- c. station , service water system, including the discharge piping.
- d. Turbine building cooling water system.
- e. Radwaste system.
- f. station auxiliary power buses.
- g. station service air system.
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- h. Valves and associated instrumentation and control systems of the reactor building normal ventilation system which must f unction - in order to switch to the standby gas treatment {
system.
RESPONSE: a. The definitions of Class I and class II structures, systems, or components, are cited in Section XII-2.1 and Appendix D, Section 2.0 of Amendment 4, Shoreham Nuclear Power Station Preliminary Safecy Analysis Report.
Radiologically with consideration to seismic events this is intended to define. .. .
(1) class I structures, systens, or components are those equipments, the failure of which, during a low probability event (e.g. , design basis earthquake), can result in the release of radiation with radiological dose consequences potentially exceeding 10CFR100 guideline limits.
(2) Class II structures, systems, or components are those equipments, the failure of which during a low probability event (e.g. , design basis earthquake) could not result in the release of radiation with radiological dose consequences in excess of 10CFR100 guideline limits. i i
In the GE-BWR direct cycle concept, all radiological sources of concern are contained within class I equipment and structures in order to comply with item 1 above. These sources remain within the class I - containment Systems since by design, and analysis, l
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/ SNPS-1 unique construction and erection techniques,- and rigid: testing
+ ' and maintenance requirements, the ." containment- integrity
- is guaranteed for all credible, postulated events. -
Refer to the station principal design criteria of-PSAR Amendment'
- 4, Section . I-2.0.
Per analysis of those sources of radioactivity which exist in class II items,-.
the radiological consequences in which a postulated class II item failure for a low probability event occurs have shown' that: 10cFR100 limits are not exceeded.
This 3.3, radiological analysis is described in detail in Section'XIV-Amendment 4, SNPS PSAR, as
' Accident". 'It should be clearly .noted the " Main that' this Steam event is only Line Break' postulated to occur but does, however, demonstrate principles of the concept of the source-containment philosophy. It illustrates that major contributing sources of any dose consequences to the public are, by design, immediately isolated and contained with integrity.
P1 ease- refer to Brunswick Steam Electric Plant, . Units 1 and- 2 (AEc Docket Nos. 50-324 and 50-325), Supplement 4,Section I.
- b. The response' to this comment will be found in Section III-2.1 of Amendment Analysis Report.4, Shoreham Nuclear Power Station Preliminary Safety
- c. The Station Service Water System and discharge piping up to and including the first block valve will be class I.
- d. The Turbine Building Cooling Water System is not considerei as class I equipment,- since this system is not' required for energency shutdown.
- e. The response to this comment will be found in Sections IX-2.2
-and XII-2.,1 of Amendment 4, Shoreham Nuclear Preliminary Safety Analysis Report. Power Station
- f. The station auxiliary power buses designated as class II l Equipment are the power buses required for operation of the station but which are not essential for safe shutdown. The !
standby diesel generators and emergency power buses which include all essential switchgear are designated as class *I Equipment.
All of this equipment will be located in a class I building.
(Refer to Sections VI11-4-1 and XII-2-8 MM. in Amendment 4, SNPS
- g. Please refei to PSAR Fig X-3-2. Class I design is not required for the !.tation service air system. The instruments and control equipment in the reactor building that require a small )
quantity of uninterruptible air for safe shutdown of the reactor, will have a separate backup supply from a reservoir located in a
,, . class I structure. This reservoir, which will be charged from A5-3
SNPS-1 the diesel generator sufficient surge capacitystarting air compressors, will have to requirements. during the time thesupply the uninterruptible air started. diesel-generators are being h.
and The response to this comument will be found in sections X-3.2 XII-2.1 of Amendment 4, Shoreham Nuclear Preliminary Safety Analysis Report. Power Station l
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i SNPS-1 COMMENT 1.3: By what means do you plan to obtain the !
information necessary to resolve each of the
,, following matters identified in 'CRS letters on !
other plants as being significant for all large !
water-cooled reactors?
- a. Verification of proposed fuel damage limits and ability of the fuel to withstand expected transients at the end of life (Browns Ferry, 3 i 67; Vermont Yankee, 6-15-67)
- b. Demonstration that fuel failures which could interfere with heat removal do not occur during a loss-of-coolant accident. (Browns Ferry, 3-14-67; Vermont Yankee, 6-15-67)
- c. Development of information to show that melting and subsequent disintegration of a portion of a fuel assembly will not lead to unacceptable conditions of fission 1 product release (Browns Ferry, 3 67; Vermont Yankee, 6-15-67)
- d. Development of instrumentation for f' more sensitive and expeditions detection of gross failure of a fuel element. (Diablo Canyon, 12 67; Fort Calhoun, 2-16-68) '
The information presented should include delineation of the responsible organization for each item, a description of the program for each I item, an analysis of the adequacy of the program to solve the problem, a schedule for each program J
and how this relates to the proposed facility construction schedule, final or interim results when available, and a discussion of alternatives in the event the program results do not demonstrate acceptable resolution of the probles.
RESPONSE: )
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- a. Information on this comment and on other ACRS-concern itees regarding water-cooled reactors has been previously submitted on the following:
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- 1) Pilgrim Nuclear Power Station, Unit 1 (AEC Docket i No. 50-293) Public Hearing Testimony,Section VII I
- 2) Browns Ferry Nuclear Power Station, Unit 3 (AEC Docket g No. 50-296) Public Hearing Testimony l AS-5
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- 3) . Bell Station, Unit '1.!(AEC Docket No. 50-319)
Amendment ' 1, C/R 1.1 f p:
- 4) Hatch Nuclear Plant, Unit 1 (AEC Docket No. 50-321)
. Amendment-. 2, C/R I-3.1. .f
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- 5) Brunswick Steam Electric Plant, Units 1 and 2 -(AEC
. Docket Nos. 50-324 and 50-325) Supplement 3, C/R .1.1 and Supplement 4, Part B, C/R 1.1.
The -identification of development programs related station to this in Section as I-5,well as to other Amendment BWRs 4, SNPS now under construction'is given PSAR. '
A detailed description of the programs referenced on page 'I-5-1 in Amendment 4, SNPS PAR.
document, i.e. APED-5608, can-General be found Electric in the Conpany referenced Analytical and
. Experimental Procrams for Resolution of ACRS Safety Concerns which has _been submitted to the AEC as a GE Topical Report. - This j Topical Report describes the technical information to be obtained for each program under the paragraph Interpretation of Concern M BWR's. A descriptive of each program is .in the General Electric Planned Action .to section 1 information is obtained, the AEC is kept ~ Resolve Concern. As topical reports. inforced by means ~ of-shown in APED 5608.Estimated completion dates for each program are
- b. Refer to Response 1.3.a I resolution.-
of this amendment for this
- c. Refer to Response 1.3.a of resolution.
this amendment for this I
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- d. 1. GENERAL 1
I Responses 1 and 7.5 and 9.4.2.1 of Brunswick Steam Electric Plant' Units 2,-(AEC Docket.Nos. 50-324 and 50-325) Supplement 3, have supplied a previous reference response to this comment.
The requiressubjectthatof failed fuel detection capability for this station the phenomenological aspects of the event and its detection and subsequent control action be thoroughly described in detail.
To accomplish this
-information purposes.
objective, the following is presented for l
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- 2. FUEL FAILURE'- NUMBER VERSUS SEVERITY
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. A. Qualitative i
The actual number of fuel rod defects that can be tolerated can only be. estimated -during normal power operation. Past-experience has shown that the BWR has operated with approximately. 200 defective fuel rods with defects ranging from pin holes to large portions of cladding missing. These defects were from prototype or experimental fuel. _Similar defects were observed at Dresden Nuclear Power Station, Unit 1,-(AEC Docket.50-010) during its ,
initial core load operating cycle. The offgas release rates l ranged from about 10 to 1,000 pci/sec per rod of a noble gas j mixture at 30 minutes decay. These defects were from stainless steel' clad fuel which is no longer being Jased in BWR design due to the more f avorable characteristics.of zirconium clad fuel in the BWR operating environment. The peak activity in the of'fgas i released at Dresden, Unit 1, with these deu. cts was approximately .]
85,000 pCi/sec. The peak activity in the offgas from Dresden j Unit 1 with the present Eircaloy fuel is approximately 10,000 J pti/sec.
B. Quantitative As an- analytical exercise, if it is assumed that a severely -
defected fuel pin (one with significant portions of cladding
{ missing) releases 1,000 pCi/sec offgas, then such an increase should be easily observable when' operating at 1,000 - 10,000 pCi/sec offgas. For the Dresden, Unit _ 1, fuel, it has been calculated that about 54 curies of the noble gas mixture resided in 'the stainless steel clad plenum. If this 54-curie plenus activity were expelled from the fuel rod defect in 100-sec., for example,- the offgas release rate would be 540,000 pci/sec average for this 100 sec. Such has never occurred even when large segments of cladding were missing. In fact " bursts" of 100 pci/sec or more have never been observed but rather the offgas release following such damage consisted of a slow gradual increase.
For the purposes of this analytical exercise, an assumed station vent release rate limit is made without exceeding the 10CFR20 annual average offsite dose of 500 mrem. Such a vent limit can be calculated only af ter sufficient meteorological site data are collected and analyzed then related through proper analysis techniques to the final vent design and characteristics of the gaseous effluent. If the vent limit were 0.5 ci/see af ter 30 l minutes decay ( or 0.018 ci/sec after 3 day decay), then the j following table qualitatively depicts possible conditions of the j fuel: '
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SNPS-10 TABLE
' Assumed Assumed Vent Percent of Number Severity Release Rate ofiFuel Release Total 1 Shorehas
, Defects
.of . per Defect. -Rate' Limit Fuel Rods Defect (uci/sec)~ (Ci/sec) Failed 30 minute 3 day Decay Decay 500 Very severe: 1,000 0.5 0.018.
portions of 1.8 cladding gone 5,000 Less sewere: 100 0.5 0.018 18 cracks , . 'many large pinholes 50,000 small pinholes 10 0.5 0.018 180 It should be reiterated that the' number of fuel rod -defects related to the release rate per defect in this example above is
'from stainless steel clad fuel as a result of Dresden, Unit 1 experience. If the station vent limit- were higher than_ 0.5 ci/sec,. for example,1 Ci/sec, then the number of tolerable - fuel defects in the _ table could be doubled or if the station vent-limit were only 50,000 pci/sec then the number of tolerable fuel
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defects 'in Table 1.3.b-1 could be divided by 10. i Actually, ,this postulated fuel defect analysis may consist ,of a spectrum of defects from the most ' severe to very small pinholes.
The number of fuel pins associated with such a spectrum is only conjecture.
It is emphasized that prior offgas release rate history is the important parameter to consider and not the number of fuel ro:1 perforations. Therefore, control over the state of the fuel should be based on this offgas measurement.
The following BWR plants ha%a vent release rate limits in terms
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of curies /second related to the 500 arem/ year dose limit in their
, FSAR analysis: ]
L Oyster No.53-219 Creek Nuclear Power Plant, Unit 1, Amendment 13, Docket Dresden Nuclear Power Station, Units 2 and 3, Appendix A, Docket Nos. 50-237 and 50-249 AS-8 o 1
SNPS-1 Quad-Cities Station, Units 1 and 2, Appendix A, Docket tios. 50-259 and 50-265 Millstone Nuclear Power Station, Unit 1, Appendix A, Docket No.
50-245 j
Monticello Nuclear Generating Plant, Unit 1, Appendix B, Docket No. 50-263 1
These vent release rate limits are all based on 'the site
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meteorology as relates to appropriate federal regulation.
Bowever, these vent limits do not represent a fuel performance requirement but represent margin between average or expected off- I gas releases . and the l 3
amount of available atmospheric dflution j over periods of one hour, one week, and one year. Each docket for the previously mentioned BWRs presents analyses on how such )
vent release limits were calculated.
The controlling regulatory criterion is dose limits set forth in 10CFR20 to control the station vent emission. Concentrations are look~ed at only when comparing biologically important isotopes such as Iodine-131. In such cases, concentration limits are not controlling. Again, analyses have been submitted in all of the previously mentioned BWR dockets.
- 3. GE-BWR FUEL FAILURE DETECTION CAPABILITY A. Qualitative Fuel failures result in the release of noble gases- and other fission products depending on the severity of the defect in the fuel cladding. The noble gases follow the same paths as the steam the from the reactor and are removed from the main condenser by air ejector as noncondensable gases and are in turn discharged through the otigas holdup and decay system and released out the station vent. Four main steam line monitors or the two vent monitors therefore serve as fuel leak detectors.
The deutiled capabilities of these instruments is presented in J Table VII-6-1 of the PSAR Amendment 4, and a detailed description of each system may be found in the PSAR Amendment 4, Section VII-6.0. Any change in offgas activity is expected to be detected within a few seconds to a few minutes to a few hours after the i action which caused the change depending on the severity of the failure and the subsequent offgas increase.
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Since the output of these monitors are recorded in the station j main control room, a continual record of any changes in offgas activity is available. l I
Supplemental methods for determining changes in activity in the l nuclear steam supply system which could be attributable to fuel leaks are by direct sampling and analysis of reactor water and laboratory analysis of the offgas boldup and decay syr; tem prior j to decay, during decay, and upon release to the station vent. j AS-9
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h These methods provide backup for determining.'the ' existence or-extent..of any. fuel : leaks, and are used to assure proper monitor functional operability. Analysis of liquid - and gas samples provides. information on the . composition ~of the isotopic mixtures-present in the reactor fluid and offgas. This information is useful-the nature in interpreting of the defect. the type of fuel failure and in assessing . y j
B. Quantitative-The ability of the GE-BWR cycle to detect the number of failed l
- fuel elements -)
2, above, of thisor fuel rods is related to the discussion in item response. i That is, the sensitivity to detect the number of failed rods by any means is a function of failure severity.
A small number of grossly f ailed rods are almost instantaneously detected as are a
- large number of pin-hole failures in many rods.
In the AEC issued and approved Technical Specifications for Oyster Creek Nuclear Power Plant, (AEC Docket No. 50-218), in discussing the GE-BWR failed fuel detection system capability, the AEC Staff evaluation statement " Limiting Conditions for-
. Operation" - is".....such capability provides the operator.with a prompt indication of any release of fission. products- from the fuel The gross to the reactor coolant above normal rated power background.
failure of any single- fuel rod -could release a sufficient amount of activity to background activity at normal- rated approximately power. This would be double the-indicative -of the onset reactor operator to the need of fuel failures and would alert the for the appro action, as defined in Section 6 of these Specifications."priate The applicant's evaluation of this concern is consistent with GE evaluations in Oyster Creek Nuclear Power Plant Unit 1, (AEC Docket No. 50-218) Amendment 44, Section 3.1. which states... .
"High main steamline radiation is an indication of excessive fuel failure. A scram is' provided to reduce the source of radiation so that site limits are not exceeded. This is accomplished by setting the scram at 10 times normal rated power background."
The following additional information clarify the basic principles involved in is submitted in order to the phenomena and to support the GE-BWR Fuel ElementFailure Detection capability. For the main steam line detectors as positioned in Fig. VII-6-1 of Amendment 4, SNPS PSAR, and with the following detector system minimum sensitivities . . . .
N-16 1.4 R/Hr for a 3 pCi/cc Noble Gases 4.8 R/Br for a 336 pCi/cc the fuel element failure detection system can detect, as an example 1 (49 fuel rods) to 2 (98 fuel rods) fuel element bundle n
A5-10
SNPS-1 failures with each fuel pCi/sec rod conservatively emitting 1,000 10 seconds with main steam line monitors.per failed rod instantaneous in 4 to Please refer to Fig. 1 within this amendment 3.d-1 for a generalized illustration of typical reactor gas and liquid path transportation times during normal operation.
The number of failure severity, normal backgroundactual of rods N-16failed
+
and detected is a functio i release, etc. The 0-19, rate of air ejectors offgas !
monitors would take conservatively if required 100 to 1000 seconds to detect the example event, !
as a backup detection l sensitivity for these monitors, as given system. The detector 7mmndment 4, SNPS PSAR, is 3 in Table VII-6-1 of x 1010 amp /R/hr. The actual subsystem sensitivity is determined later when air ejector leakage is established.
would assure that the stationEither' detection monitoring capability elevated exceeded per 10CFR20 regulations. release limit was not
- 4. CONCLUSIONS The abovecapability detection examples is demonstrate that the BWR fuel element failure both qualitatively and adequate for the ACRS concern item. quantitatively i
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COMMENT 1.49
. We understand plant that the R&D. effort for a number features,such of jet pumps, rod velocity limiter, in-coreas the rod worth monitor system, and neutron operability of the steam demonstration of the under accident conditionslinehas isolation valves been completed.
Please report the results from any such completed R&D effort of those notand indicate the status and schedule completed.
.)
RESPONSE
this facility.The items stated above are not R&D efforts assigned to I-5.0 as stated reflect the areas of AEC concern associ this facility. with The and RED effort items cited above have 15een successfully comp are or will shortly be reported to the AEC.
are referenced to a Topical Report. All items c.bove (
1 The above comment has been answered in detail previously following: on the i
a)
Bell Station, Unit 1 (AEC Docket No. 50-319)
Amendment 1, C/R 1.0 b)
Hatch Nuclear Amendment 2,Plant, Unit 1 (AEC Docket No. 50-321)
C/R I-3.0 c)
Pilgrim Nuclear Power Station, Unit 1 (AEC Docket No. 50-293)
Testimony, Section VIIAmendment 2, C/R 2.0, Public Hearing d)
Monticello Nuclear Generating Plant,
,No. 50-263), FSAR Section I-4.0 Unit 1 (AEC Docket The 'following topical listed future:above have been issued to the AEC or will be inreports conce the near APED-5446 - Control Rod Velocity Limiter APED-5449 - Control Rod Worth Minimizer APED-5555 - Rod Impact Testing on Collet Assembly for control Drive APED-5454 - CoolingMetal-Water SystemsReactions-Effects on Core Standby APED-5460 - Jet Pump Testing and EvaluationAPED-5458 APED-5706 - In-Core Neutron Monitoring System for GE-BWR AS-12
SV4PS-1 COMMENT 1.5: Please discuss the extent to which physical separation, shield walls, automatic fire fighting equipment and other design provisions will be used to minimize the probability of physical damage to more than one component of redundant critical equipment due to a failure of one of the components, a fire, or a natural disaster such as a tornado or earthquake. Include discussion of provisions for critical instrumentation and control equipment and electric power systems.
RESPONSE: Appropriate sections of Amendment 4, SNPS PSAR, have endeavored to deceribe in detail the design provisions to minimize physical damage to more than one component of redundant critical equipment. A brief summary of these conditions follows:
- 1. Critical instrument, control and power leads will be duplica-ted and run separately. (See Section VIII-3, Amendment 4, SNPS PSAR)
- 2. Critical systems, such as those associated with emergency core cooling and emergency power, will, as a minimum, satisfy the single f ailure criterion. This equipment will all be Class I and housed in Class I structures. All of it will survive all design basis accidents. This specifically includes tornadoes and earthquakes. (See Section II-11, Amandment 4, SNPS PSAR)
- 3. All class I structures for this unit will be reinforced concrete, and therefore fireproof. Equipment and system design features include careful selection of material to avoid a fire and to limit the consequences if fire should start. Provision has been made to control effectively and extinguish any fire which might occur due to the unavoidable presence of combustible material. (See Sections X-3 and X-4, Amendment 4, SNPS PSAR)
- 4. Physical separation of critical equipment is a design feature. Reference to the arrangement drawings (Figs. V-2-1 through V-2-5) will show that duplicate core cooling equipment is typically on opposite sides of the reactor building with the reactor between. The control room is in a class I ouilding which is separate from both the turbine building and the reactor building. The emergency diesel-generator units and associated switchgear are isolated from each other by firewalls. The ,
batteries, chargers and distribution boards are similarly '
separated.
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1 AS-13 l
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COMMENT 1.6: . Pleas'e provide a more complete discussion of- your quality assurance requirements. imposed on those materials, components, and. systems of, the shoreham plant essential to the prevention of-accidents which could affect the public health
. 'and safety, or to the mitigation of their-cons equences. Your discussion should include all activities which may affect the quality of these items including- _ designing, purchasing, fabricating, handling, shipping, storing, constructing, inspecting, testing, operating, maintaining, repairing , - and ' modifying. As a L<
minimumi the following information should be I discussed:
- a. overall Quality Assurance Planning -
The quality assurance- program encompassing all phases ' leading to a safe and reliable nuclear power plant-which -will be established, documented, executed and maintained by the applicant and his contractors and subcontractors.
- b. Organization - Your organization and personnel staffing to providei assurance that the authority and-responsibility" of persons and organizations performing quality.
functions will be clearly established and delineated in writing and that they will have sufficient j organizational freedom to. identify I quality problems and to ensure that j solutions are provided. Your <
application states that LILCO will.
establish a " Quality Assarance section" (ref. pg I-7-2) which will be responsible for auditing both the {
field and shop quality assurance l programs. Please provide additional I information which outlines the !
composition and functions of this I group and how it will relate to the ST,W and GE quality assurance
, organizations and functions described in your PSAR. This should include charts which indicate -the lines of I
communication, responsibility and l authority within and between each j l organization. 1 L
certain critical systems, including
.w the nuclear steam supply syster, scre i i
A5-14
SNPS-1 of the emergency s i systems, core. cooling and their associated instrumentation, and control systems, will
' apparently not be designed and/or provided entir~ely by one organization. Please explain how the
, design and quality assurance of such systems will be coordinated to assure j that the the integrated system will have intended performance capability.
- c. 1 Procedural quality. Documents - The extent that !
assurance program requirements will be implemented by written administrative policies, procedures, and instructions,
- d. Design Review independent, The extent that documented comprehensive, I assessments of the adequacy of design will be accomplished for major components and systems important to safety to assure compliance with criteria, codes, standards and requirements.
- e. Control of Specifications, Drawings,
( Procedures, and Instructions - Your systems prs.cedures, to assure that instructions, specifications, and drawings will be complete and current and will be readily available at job site. the
- f. Work instructions, Proced ures, and Drawings - The extent that all work affecting quality will be documented.
Include a tabulation of the procedures and instructions,
- g. Purchase Specifications - The extent that applicable criteria, codes, standards and requirements necessary to assure adequate
- quality and conformance to design characteristics i
will be properly included referenced or I in specifications for the procurement of materials, equipment ,
j systems, structures, and services,
- h. Control of
! Purchased Ma terial, l Equipment and Services - Your plans and systems 4 to assure that all I
purchased material, equipment and AS-15
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- v. services conform of purchase specifications.to the' requireme
- i. Control. . and lIdentificatt a Materials Parts,-
- Your sys, tem to assure ;controit thatand ts !
identification of materials. -parts and components will throughout be maintained with'their intended use.all operations con j.
Special Process : Control - Your plan and system to processes, includingassure ' that special' welding,: heat treating, and non-destructive testing 1 1
will be controlled applicable' codes, in accordance wi specifications standards,-and and
. accomplished by qualifiedwill be using qualified procedures, personnel k.
In-Process provisionsand Final' Inspection - Your .
to ensure documented, that planned,. t
'in progress 'and final 1 l inspections.at appropriate stages of.
fabrication, installation willconstruction, be and-with1 written in accordance-procedures instructions. and
- 1. Test Control - I assure that Your provisions-be. performed functional testing vill to- j I
under controlled I conditions test proceduresin accordance which incorporate with thewritten requirements contained and acceptance limits in specifications, applicable design documents. standards, and other m.
Calibration Equipment of Measurement and Test that Your system to assure tools, gages, and other measuring calibrated and testing devices will be in recognized standards accordance and procedures. with
- n. Handling, Storage, Shipping and Preservation - Your plans and systems for the provision and use of adequate work handling,and inspection instructions for storage, i preservation of equipment toshipping and !
damage or deterioration. prevent I
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- o. Site Cleanliness During Construction
, -- Your criteria and procedures to be invoked relative to establishing and maintaining site cleanliness during construction.
- p. Nonconforming Material, Parts, Components, or Workmanship - Your provisions for the identification and control of material, parts, components, and workmanship which do not comform to criteria, codes, standards and requirements.
- g. Corrective Action - Your provisions to assure that conditions adverse to quality will. be detected and reported, the cause of.each condition will be determined, and corrective action will be taken to preclude recurrence.
- r. Quality control Records - Your provisions to assure that complete i and reliable records suf ficient to furnish documentary evidence of
. product quality. will be maintained.
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- s. Audits - Your provision for audits i
. which will be established to assure compliance with all aspects of the quality assurance program and to detezzine the effectiveness of the program.
RESPONSE: The response to this comment is found in Appendix E, Amendment 4, SNPS PSAR.
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2.0 SITE AND ENVIRONS "
COMMENT 2.1 Since !
the proposed facility is almost location _of the Shoreham directly in line with the projected flight path from one of the runways of the Grumman Aircraft (Peconic River) Airport, what'special design provisions will be made in the design and/or operation of. the Shorehar facility in order that an aircraft crashing the facility sould not result in the releaseinto of excessive radioactivity to the environs? If you do. not plan to make any special design provisions for such a contingency, submit'in detail your basis for not doing so. In either case, the probable future use. of the airport should be considered.
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RESPONSE: The response to this comment constitutes Amendment 3 to the Shoreham Nuclear Power Station Preliminary Safety Analysis Report.
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COMMENT 2.2: Describe'the methods'used to make the population projections of the 1980 population distribution with distance around the proposed site. How do these projections compare- with the population growth. rate of 'similar areas of ' Long Island to {
the. west . of. the. site? Discuss the potential- 1 effect on the population growth of the-' area surrounding the site if, as has been proposed, a bridge linking Long Island to connecticut is constructed.
RESPONSE: Methods used to project population distribution in the vicinity of the site and the comparison of these projections with the growth rate of similar areas of Long' Island are discussed in Section II. 3, - Amiendment 4, SNPS PSAR. The- present response is addressed solely to the effect of additional bridges.
At least five (5) different bridge crossings from Long Island to New York and Connecticut have ~been considered, and studied by l
various public and private groups during recent years. Of these, the New York State Metropolitan Transportation Authority's recommendations to construct two crossings, one- from Oyster . Bay to Rye, New York and a second f rom Port Jefferson to Bridgeport, Connecticut, seem most likely to be carried out.
Considerable effort will be required to overcame public opposition, to enact legal authorization and to obtain right-of-t 'way'and funding. Bridge and approach road construction will require approximately five years from the ' time initial work is begun until the bridge is in service, based on time periods required to construct' the Throgs. Neck and Whitestone Bridges.
The LILCO Planning Department believes that both the Oyster Bay to Rye and the Port Jefferson to Bridgeport bridges will be constructed. The earliest reasonable starting dates could be 1975 for one and 1980. for the other. This is reflected in the projections for population growth shown in Table II-3-1.
These bridges have been assumed to be an inevitable aspect of the projected growth, and the prime effect will be to assure it's continuation. This growth will be influenced by many forces.
Long Island communities each have well defined zoning and land use restrictions, which are not easily revised. Large blocks of land already zoned for residential and industrial uses are vacant and await the population growth which is expected. But time and
, funding required to provide services such as water, highways, u sewers, schools and other community facilities will regulate f growth.
l In summary, it appears that Long Island population will continue to increase toward urban densities. The time period for this change will be measured in decades. If the two bridges are constructed, they will serve to assure the continued development 1
A5-19 l
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SNPS-1 of,Long Islands'but this will not make a' major change 'n i the
- projected population growth rate.
O AS-20
SNPS-1 COMMENT 2.3: Please provide additional details' on your preoperational environmental monitoring program including length of _ time the program will be in operation
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prior j to plant startup, sampling frequencies and and locations (preferably on a map),
the types and number of samples to be taken.
For those details not provide available at this time, a schedule information.
for submittal of the
RESPONSE
II-10.1 The response to this comment will be found in Section of Amendment 4, Shoreham Nuclear Power Station 1 Preliminary Safety Analysis Report. l
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l SNPS-1 I COMMENT 2.4: It is stated in the PSAR that the Population-Center Distance for this site is 18 miles;however, Table II-3-1, lists a projected 1980 population of approximately 30,500 between 5 and 10 miles of the site in the S to W sectors.
Considering these data and the guidelines of 10CFR100 regarding low population center distances, please discuss the reasons for your proposed Population Center and Low Population Zone distances for this site.
RESPONSE: The response to this comment is found in Amendment 4, Sections II-3.3 and II-3.4, Shoreham Nuclear Power Station Preliminary Safety Analysis Report.
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COMMENT 2.5: Please provide estimates of the number of temporary ' residents and the beaches in the area of thedaytime visitors along summer. site during the RESPONSE: Estimates Shoreham area may be obtained fromof the number of temporary residents in the the difference between the summer and winter population forecasts as - described in Ser: tion II-3.1.2.
Within a ten mile radius of the reactor there are approximately 5,500 additional sunmer residents.
The north ' shore site . consists - mostly ofof Long Island land private in the vicinity of the Shorehas beaches. There holdings and private there are no data available on daytime this area. Therefore is no commercial beach in beaches. It is a reasonable visitors along the category are largely limited to neighborsassumption and that persons in this acquaintances the owners. of have been included in the estimate of summer residents.The pe Since the station summer site have ofbeen 1967 the beach and parking facilities at the
. Town of Brookhaven. made available to the residents of the two used susmers the beachindicate that on the average approximately 35 Rec people season. per weekday, LIIco plans 150 per weekend day, and 8,000 per to rearrange facilities enring the the beach and parking permanent and use after station startup. construction Althoughperiodimproved to provide for their access planned that the beach will retain its naturalparking will be provid l rural character. I I
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i SNPS-1 COMMENT 2.6 Evaluate the potential flood levels that could result from a prolonged " northeaster" storm which has a probability of occurrence comparable to the Probable Maximum Hurricane.
RESPONSE: A study has level which reasonably could beenbemade to evaluate espected the at to occur maximum flood the Shoreham site as a result of a northeaster which has a probabilit occurrence of equal to that of the Probable Maximum Hurricane.y The floodcompared as levels which would occur at the site fromevaluation w the northeaster to. those already determined hurricane, and described in Section II-8.3 from the design of the PSAR. This Maximum Hurricane as delineated in Reference 1.later st The question of comparison of flood levels hurricanes and northeasters was considered by Nationalresulting from ;
Reasearch Project Report Hurricane No. 68 (Reference 2) .
that report types of compares the significant characteristics ofTable the No. 8twoof storms north of Cape Cod.
historically the greatest storm surges north of CapeThatCod from northeasters. report result also foun In spite of these findings, concluded that, the report decision Cod have asthe to whether hurricanes or northeasters northdue t of Cape greater damage potential must await surge computations based upon the comparison of isovel charts for the Standard Project Hurricane and Standard Project Northeaster.
It is our opinion comparing the surge effects of the two types ofthat a conclusive ans Island sound would require a storms on Long similar analysis.
Unfortunately, no Standard Project Northeaster hastype been of !
for this part of the North Atlantic ccast. defined l However, a qualitative storm-surge effects for analysis has been made of the relative Island northeasters and hurricanes in Long New England coast north of Cape Cod. Sound as compared to the eff levels at New London, Connecticut, an estimate has beenAlsomade the probable based on obse maximum stillwater of from a northeaster. This analysis leads to thelevel at this point resulting ;
the conclusion that '
from the Probable Maximum Hurricane ismaximum greater than water level predi reasonably could be expected from a northeaster with the same probability of occurrence.
conclusion. The following evidence is presented to support this Qualitative Comparison of Burricanes and Northeasters in Long Island Sound and North of Cape cod certain trends hurricanes andhave been observed which are characteristic of northeasters as they Atlantic coast. travel north These trends indicate that the surge producing along the AS-24
SNPS-1 1
b, features ' of hurricanes generally diminish as the storr moves '
northward while, conversely, northeasters tend to' intensify. j According to Reference 2, northeaster surge-producing potential -
l is . greatest for those. storms with . lowest central pressure,
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' strongest maxieum pressure gradient and longest ' fetch. It has ~
been .found that these ' critical- parameters often become more intense . as a northeaster proceeds northward. ' According to Appendix B of' Reference 2,' the central pressure for a storm transposed from 36 deg N to til deg N along the 68th meridian would decrease from 970 to 961.3 ab. . It was also noted that' the central pressure of all but one of 51 surge-producing cyclones studied decreased during the 12 hr period before the surge reached its peak at Boston. or Portland. The mean rate of
. deepening was _ about 9 mb/12 hrs. The more notable surge -
producing northeasters were also associated with the phenomenon of blocking high pressure located in advance of the storm center, which tended -to impede their forward motion, to tighten the isobaric spacing with a resulting increased. pressure gradient, and. to lengthen the radius of isobaric curvature, creating an unusually long fetch in the forward semicircle of the storm. All i of these factors tend to increase the surge producing potential
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of northeasters as they move northward.
Table- 7 of Reference 2 shows the change in the significant characteristics of hurricanes as they move northward. It is noted that the wind velocity decreases-, the forward speed increases and the central pressure index increases, all of which (
tend to reduce the surge producing potential of hurricanes. j l
A significant contrast was found.when tide records were examined to determine the type of storm which produced the highest' water levels in Long Island Sound as compared to north of Cape Cod.
Reference 2 states that only three hurricanes were included in the final selection of maximum annual storm surges used in the surge study at Portland and Boston, and that the surges from a number of northeasters were larger than from these hurricanes. {
On the other hand, the five highest tides recorded at the U.S. ;
I Coast & Geodetic Survey tide gage at New London were all associated with hurricanes.
The above comparison of trends in . characteristics of the two types of storms as they move north coupled with the historic data f on tide levels indicates quantitatively thac, in contrast to the .
findings of Reference 2 for north of Cape Cod, the maximum tide (
levels in Long Island Sound will result fros hurricanes.
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Storm Surge at New London, Connecticut, from a Northeaster The New Iondon tide gage records were analyzed to obtain a value for the probable maximas still water level at this location from a northeaster. The New London gage records, which have the l
longest period of record for a station close to the Shorehas
' site, were selected for study because a comparison can be made !
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SNPS-1 with the previously determined still water level at this location
_ from the Probable Maximum Hurricane.
Tide' gage records for New London, covering the period from June 1938 through October 1967, were examined, and the maximus annual readings, excluding hurricanes, were arranged in order of increasing magnitude.
The gage readings, converted to height above mean low water, were then plotted on probability paper using the distribution p , 100 (M - 0. 5 )
Y where P is the frequency of occurrence per 100 years, M is the rank of the gage reading, and Y is the number of years of record.
A curve was _ drawn through the plotted points as shown in Fig.
2.6A. This curve, which shous the frequency of occurrence of.a particular tide level per 100 yr, was extended to a frequency of 0.01 (a return period of 10,000 yr) to give the tide level of 9.5 ft MLW.
The observed tide gage readings used .in the above analysis include the total effect of the astronomical tide plus the storm surge effects at New London doe to the particular hydrography of the location. However, these surge effects probably did not all occur at the time of the extreme astronomical tide. In order to make an adjustment which would reflect the possibility of the simultaneous occurrence of a storm surge and the extreme astronomical tide, it was assumed that the observed tide readings occurred at the time when the astronomical tide was at mean high water, which is 2.6 ft MLW datum. From the tide tables, it was found that the maximum predicted astronomical tide at New London is about 3.7 ft MLW. Accordingly, the difference between these two values, 1.1 ft, was added to the previously projected level of 9.5 ft to give the extreme still water level at New Londor, of 10.6 ft MLW. This value, when compared to the previously determined Probable Maximum Burricane tide level of 13.8 f t MLW
'at New London, (Fig. II-8-4 of PSAR) indicates a maximum still water level of 3.2 Probable Maximum Burricane.
ft lower for the northeaster than for the Maximum Water Level at the Shoreham Site From a Northeaster In predicting the maximum water level at the shoreham site from the Probable Maximum Hurricane, that storm was positioned to produce the optimum surge effect from northerly crosswinds. In addition, wave runup from these considered. However, 'ferences 1 hurricane force winds was and 2 indicate that the maximum wind velociti- fram a northeaster would be less than from a hurricane. Also it can be shown (Ref. Fig. 1-7 of Reference 4) that at the wind velocities being considered, the wave heights would be limited by the fetch of 20 miles, measured across Long Island Sound in a southerly direction from New AS-26
SNPS-1 London.. Therefore, due to lower wind velocities in the 4
northeaster, the water level increase resulting from crosswinds from New London to the site probably would be less from the northeaster than from the Probable Maximum Hurricane. However, the wave run-up and surge effects from the cross winds are conservatively assumed to be the same as previously determined for the hurricane.
Tnis leads to the conclusion that the maximum flood level at the shoreham site resulting from a northeaster will be no'less than 3.2 ft lower than the water level resulting from the Probable Maximum Hurricane. Therefore, based on the maximum flood level of 20.5 f t predicted for the hurricane, the flood level for the northeaster whch has a similar probability of occurrence will not exceed 17.3 ft MLW.
REFERENCES
- 1. -Interim Report - Meteorological Characteristics of the Probable Maximum Hurricane, Atlantic and Gulf Coasts of the United States, HUR 7-97, Hydrometeorological Branch, U.S.
Weather Bureau, May 1968.
- 2. Peterson, K. R. and Goodyear, H. V. , Criteria for a Standard l Project Northeaster for New England North of Cape Cod, National Hurricane Research Project Report No. 68.
U.S. Weather Bureau, March 1964.
- 3. Tide Gage Readings, New London, Connecticut, U.S. Coast and Geodetic Survey.
- 4. Shore Protection Planning and Design. Technical Report No. 4, U.S. Army Coastal Engineering Rc7earch Center, Third Edition, 1966.
- 5. Tide Tables, High and Low w a*.rr Predictions, East Coast North and South America, U.S. Coast & Geodetic Survey.
h AS-27
SNPS-1 COMMENT'2.7:
Infor your application you provide boring data only a " typical" boring. Please provide the logs for all borings made at the site and include the standard penetration resistances and relative densities of.. all standard penetration tests taken-in sands below elevation 12.
RESPONSE
The response to this comment' will be found in Exhibit
"]!" of this Amendment.
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SNPS-1 COMMENT 2.8: Describe in detail the ' method
-ly of obtaining gs
" undisturbed" samples in the clean sand strata ha and explain how densification of the ~' sand was va Prevented while drilling, sampling, transporting, en and- testing the sand samples. Relative to the Plot of percent of relative density versus elevation, presented on Figtre II-5-3 of the PSAR, how was the solid line labeled " median of it' Penetration- tests" determined? Which borings.
were used in establishing this plot and was more weight. given to those borings nearer the containment building? Were the Raymond and Giles N values given equal weight in determining the "5-3?median" and
- lower quartile" lines in Figure II-RESPONSE: i Obtaining of ' undisturbed samples in borings was done using the Acker Dennison soil sampling barrel. This sampling i
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tool is designed of soils,- and for. taking undisturbed samples of a wide variety '
especially soils having some content of granular-materials, such~ as gravelly clay. The sample is retained by wall-friction in cohesive soils. A basket-type retainer is provided for use j site. with noncohesive soils, and was found necessary at this The sampler consists of two tubes. The outer tube rotates and has a cutter shoe, while the inner nonrotating tube is 1:Ined with very thin brass liners so that the sample can be recovered and preserved. 1 The outside diameter of the sampler was 4 in, and the nominal diameter of the sample was 2 13/16 in.
Taking a the sample involved a rotational process (100-140 rps) and steady downward force, with the result of pushing the tool in a continuous motion.
A bentonite slurry drilling fluid was used while drilling in order to prevent the hole from caving in and to carry cuttings to 4
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the surface. The penetration of the mud cake at the ends of the sample was limited.
Densification of the sand was avoided by:
- a. Avoiding, while drilling, any sudden changes of the rotational speed and any sudden changes of the pressure on the tool and of the pump. 4 J
- b. Having the rig equilment and the pumps regularly checked in order to make sure that they were working smoothly. ,
- c. Keeping low rotational speed and low pressure on the tool. I 4
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- d. Keeping- the cutter shoe at a distance of 1 in. above the cutting edge.of the . inner. nonrotating tube which takes the sample.
- e. Exercising special care when taking; out the liner containing the sample and lightly greasing the outer wall of the liner: 4 before its introduction, in order to assure its smooth removal withcut the need of tapping or using a has.aer.
f.- Filling completely with wax the liner where this was not completely filled with sample (recovery was in the range of 70-95 percent). ]
- g. Storing the samples immediately after waxing in heated barns with a continuous temperature'of 65-70 F. q q
- h. - ' Transporting from Shoreham, Long Island, to Boston laboratory of Stone & Webster, in a winterized panel truck rented for this job, with samples packed in foam-padded boxes. _ Loading, unloading, and transport were done under the direction of an I
experienced soils engineer.
i To determine the in-place dry density, sealing wax and mud cake .
were carefully removed and the ends of the soil trimmed approximately to an even plane by light scraping. Its volume was then determined by measuring the sample liner and deducting the distance from the end of the liner to the surface of the soil.
The entire sample, including the liner, was then weighed, following which the sample was ejected, dried, and the moisture content determined. By subtracting the tare weight of the liner and the weight of water from the total weight, the dry weight of soil was determined,and, from this and- the known volume, the in-place dry density determined, Data from the following borings were used for determinations of relative density from standard penetration test results: Borings 2, 3, 4,5,6,7, 10, 11, 14, 15, 102, 103, 104, 105, 106, 107, 110, 113, 117, and 118. These borings are located in the plant area or close to it. Borings B12, B13, and 101 were not included as being remote. Boring B16 is located in the edge of the marsh area and contained clay deposits in its upper portion. Berings B108, 109, 111, 112, 115, and 116 were drilled as undisturbed sample borings and, accordingly, standard penetration tests were not made in these borings. Borings B1, 8, and 9 were originally scheduled in the first series of investigations at locations well to the west of the plant area, but were never drilled.
. Experience with this same series of .outwash sands at Brookhaven National Laboratory has shown that, while they vary erratically in grain size and density within a few feet vertically and within a few tens of feet laterally, they are, over a large area, statistically homogeneous. Thus, in our opinion, a boring immediately at one location is not significantly more determinative of conditions at a distance of 50 ft from it than is a boring several hundred feet away. Under these conditions A5-30
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with sampling at. intervals, we consider that the results from c- number single of borings. give a better picture of conditions than any boring. Accordingly, additional weight was not given the borings- isusediately under the containment structure as compared with more remote borings.
To determine the median and lower quartile values of relative _ density, as shown on Figure II-5-3 and in Table II 2, standard penetration test results for all samples taken .in the borings listed .above between two specified elevations, as, for example, below El. -15 and above El. -25, this specific zone there were 45 tests. were tabulatedt for Other zones contained about the same number of items. The median was then selected by determining the "N" value than which half the values tabulated were smaller and half larger. For the lower quartile, the "N" values than which one quarter were smaller, were established in the same manner. The effective stress was then computed for the mid-point of the zone. Since a number of borings were included which were drilled from different elevations, an . assumed ground surface of El. +35 and ground water'1evel of El. +9 were used in computing the effective stress. This ground elevation is slightly higher than the average elevation of the borings, and this procedure, therefore, is slightly conservative. Using this effective stress and the "N" quartile, as determined above, the value for the median and lower relative density was then deterinined from the curves proposed for. " Average Sand, Dry or Moist," by Gibb and Holtz. The same procedure was then used the next increment. for depth were taken at 10 ft.
From. surface to El. -45, the increments of The increment from El. -45 to -60 was taken as one increment, and below El. -60, increments of 20 ft were used.
In making these . studies, no distinction was made between data from the Raymond and from the Giles borings.
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COMMENT 2.9: Based on submitted,. their
. ourreview . of the information consultants on geo?ogy..you- and seismology. have indicated that they will probably recommend 0.2g be~ that acceleration. values. of- 0.1g .and used and design basis for the operating basis earthquake earthquake . for. - the Shoreham plant, rather than the 0.07g and 0.15g values you proposed.
Please ' indicate ' your acceptance of these values or provide additional information to.
support the values you proposed.
DESPONSE: As a portion of the original _ PSAR s:.te , on. the Shoreham-a report was prepared by Dr. H. B.
earthquake motion through the _ overburden materials Seed on transmission of l bedrock for' this site, based above the on . certain _ - assumed earthquake acceleration levels at bedrock. This report was included as Appendix A.
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.since the time of this original study . analytica1' techniques and procedures for evaluating amplification of earthquake motion -to overburden materials have been extended and refined by studies undertaken by Dr. Robert V. Whitman of M.I.T. and his staff, and improved programs developed and )
incorporated library of Stone ' & Webster Engineering Corporation.in the Using computer these recently developed approaches and techniques, reanalysis of earthquake effects a thorough at the site was made. The assumptions, results, and conclusions of this analysis are contained in Dr. Whitman's report. (Exhibit A attached. ) t It was agreed with the consultants of the AEC that the design basis earthquake for this site strong Intensity VII on the Modifiedshould be characterized as a intensity scale Mercalli Scale. The MM a given location. is a subjective measure of structural- damage at Observations of structural damage due to i ground !
shock from blasting have indicated that particle velocity acceleration. isAccordingly, at the surface a better measure of structural damage than is in these later studies, it was decided to establish 8 ips as the maximum particle velocity at
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l the earth surface and to adjust input data at the rock surface to achieve this particle velocity. {
i Using Neuman's relation between surface velocity and MM intensity, MM a surface intensity of VIII+, particle velocity of 8 ips corresponds to an and thus would include all possible definitions of a strong intensity VII. Analyses were made for four earthquake records, normalized to produce surface velocities of 8 ips.
1935, These included the El Centro 1940, Taft 1952, Helena and an artificial record developed by random processes and having properties intermediate between Taft and Helena. In addition, the record of the 1957 Golden Gate earthquake was used.
For this earthquake, it was found that particle velocities at the surface i were rather suppressed and, accordingly, this earthquake was normalized to an input acceleration in the rock of 0.15 g, which is slightly larger than the maximus acceleration recorded AS-32
- _ _ -__ D
SNPS-1 for' this earthquake which had a magnitude. (Richter approximately. 5 1/2 on rock near the causative ' fault. Scale) of For each earthquake, time histories of the velocity at the earth surface, structural' response spectra for varying amounts of
-elevations, structural damping, time histories of shear stresses for selected maximum and average shear stresses for the eight largest pulses from ground surface to bedrock.- strains in- the soil mass ' at various elevations, . and ratios of response spectra at the rock surface to response spectra at the ground surface were computed.
. Whitman report. These data are illustrated and summarized in the In order to compare the various computations, it was spectra obtained in these tripartite paper.
- decided to plot the response spectra on To minimize confusion, only the . spectrum 5 percent structural damping was plotted. This plot is shown for on Fig.12 of the Whitman Housner Spectra for report. For comparison, the standard 5 percent structural damping normalized to ,
0.15 g and 0.2 g have also been plotted. It will be noted that !
the the 0.20 g standard several response spectrum provides a satisfactory envelope for spectra plotted, except for the velocity portica of the-spectrum for the El Centro earthquake.
Newmark, by 1.9 for in 1967, recommended that ground velocity be multiplied spectra. If 5 percent structural damping in developing response this is done,
' it results in a velocity portion of the envelopsspectrum the Elhaving Centroarecord pseudo-velocity of 15 ips which reasonably within the periods of interest structurally for this site.
the heavy dashed line in This adjusted spectrum is shown by Fig. 12. Accordingly, it has been decided to use for the design basis earthquake response spectra which coincide with the standard Housner Spectra, normalized to 0.20 g, for structural frequencies exceeding about 2 cps, but to increase the velocity portion of the spectra in accordance with the multiplier proposed by Newmark to account for a ground velocity of 8 ips.
Accordingly, the spectra proposed for design basis earthquake are slightly more conservative than the standard ;'
Housner spectrum, normalized to 0.20 g, within the portion of the spectrum where velocity is significant, that is, approximately 2 eps and slower., i These design basis earthquake spectra are shown in Fig. 2.9B. The operational basis earthquake will be taken as ;
one half the design basis earthquake and, consequently, will use spectra- which in the velocity range are slightly more conservative than the Housner Spectrum normalized to 0.1 g and coincide with the standard Housner Spectrum for frequencies exceeding about 2 cps. These operational basis earthquake spectra are shown in Fig. 2.9A.
In the PSAR as originally submitted, very conservative assumptions were used in determining possible shearing stresses in the soils underlying the site for purposes of analyzing the I underlying soils for safety against liquefaction potential.s As indicated previously, the analytical techniques developed and AS-33
SNPS-1 ,
used in this' study permit determining for each earthquake input the shearing stresses throughout. the depth of the soil mass. !
This includes the peak shearing stresses, the time history of chear a
stresces at any individual mass considered, and the average of selected number- of the largest pulses. Shear rtress distribution in the soil mass under earthquake conditions is discussed in detail in Dr. Whitman's report. The findings may be~
summarized, however, by considering the maximun shearing stress and the average of the five and ten largest peaks for each of the five earthquakes considered at an appropriate eleva tion which, for convenience, has been taken at El. -32.5, approximately 20 ft below founding grade of the containment structure. These results are tabulated as follows:
Earthquake Shear Stress, Psf 9 El. -32.5 Average, Average, i Peak 5 Largest 10 Largest l El Centro '40 450 343 252 Taft '52 499 284 199 Helena '35 {
698 270 * '
Artificial 388 196
- Golden Gate '57 297 158
- l l
- Re' cord did not contain ten large pulses Dr. Whitman has recommended, for purposes of analyzing liquef action potential, using an average shear value of }
400 psf J for the first five pulses at El. -32.5 (for the ground surface at El. 20) or 300 psf for the five largest pulses. These values are conservative, compared with the highest average for five pulses found in these analyses.
Evaluation of the effect of these ' revised shear stresses on liquefaction can be done by comparing these new values with the conservative values previously assumed for the yard area, as shown in Table II-5-4. By interpolation of estimated values, it '
is found that, for these previous studies, the average shear stress at El. -32.5 was assumed to be 625 psf. It may be noted that the elevation selected is in the depth range having the minimum factor of safety. This indicates the factor of safety against Initial Liquefaction for the sands underlying the site to i be of the order of 4.0 for the median of the penetration test i results. As indicated in Table II-5-4, the factor of safety against liquefaction under other structures is approximately l
equal to that which would exist in the yard area. Since Initial l Liquefaction is taken as that number of cycles of loading where i
I the pore pressures first become equal to the confining pressure, a number of additional cycles of loading would be necessary before significant strains developed in the soil mass. The f actor of saf ety against significant strains developing under earthquake loadings would exceed 4.0.
AS-34 l
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SNPS-1 I'
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These analys es, therefore, show that for a strong intensity VII earthquake at this site, there .is no hazard of liquefaction.
>- They' do show, however, that a response spectrum for structural analysis which coincides with the standard Housner, spectrum, normalized to. 0.20 g acceleration. for structural frequencies exceeding approximately 2 cps, is reasonable and will be used. l A somewhat more conservative spectrum will be used.for structures i having frequencies less than 2 cps.
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AS-35 l
SNPS-1
- 3.0 REACTOR CORE F
j COMMENT 3.1: How does the design reduce the probability of the capscrews that fasten the flow channels to the fuel assembly coming. loose and being released into the coolant water? What are the potential consequences of one or more of these capscrews getting into the coolant water stream?
RESPONSE: The Channel Fastener Assembly is attached to the Fuel Bundle upper tie plate by a 5/16 '18 UNC spline head capscrew of stainless steel. An Inconel spring type lockwasher is positioned under the head of the capscrew. The capscrew is installed with '.
70-80 in.-lbs torque which seats the lockwasher and produces the required torque-tension relationship to ensure a captive assembly. The Inconel lockwasher is not adversely affected by reactor operating or accident temperatures or environment. )
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This basic design has been used on over 3,400 fuel bundles installed in the following operating reactors: {
j a) Dr'esden Nuclear Power Station, Unit No. 1 (Docket 50-10) }
{
b) Humboldt Nuclear Power Plant, Unit No. 3 (Docket 50-133) l c) Kahl d) Senn e) JPDR f) KRB g) Tarapur There have been no incidents of loss of these parts during !
reactor operation.
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SNPS-1 COMMElft 3.2:
, Please describe the analytical method developed by 4
GE (pg. VII-2-11) which was used to arrive at the
~ conclusion- that xenon oscillations will be-strongly damped and .therefore will present no problem in the Shoreham reactor (pg. III-2-3, and VII 11) . What are the values of the significant nuclear parameters analysis? How have the results used in this given by this analytical technique been verified? In the event that a spatial flux oscillation did occur, how would it be detected by the operator and what could he do to suppress it?
RESPONSE: The entire concern of xenon in boiling water reactors is extensively Report discussed and resolved as reported in GE Topical GE-APED-5614 0, June, 1968, Class I, titled: " Xenon Considerations in Design of Large Boiling Water Reactors" by R.
L. Crowther, which has been submitted to the AEC.
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A5-37
l SNPS-1 4.0 REACTOR COOLANT SYSTEM COMMENT 4.1: On page G-2-38 of Appendix G of the PSAR you indicate that .you disagree with AEC General Design criterion 35 and that the Shoreham design does not conform to this criterion. To clarify your position 'on this subject, please' indicate your concurrence with responses previously -
provided by your nuclear steam system supplier, GE, .in Section 3.8 of Amendment 1 of the N. Y.
- State Electric and. Gas Corporation . Bell Station application (Docket No. 50-319) or, provide the following additional information.
- a. In Section IV-2.3 of your . PSAR, you state- that ferritic materials will behave in a ductile manner at temperatures 60 F above the ' NDT '
temperature. However, extensive work done at NRL (Pellini) indicates that fully ductile behavior. for these materials is only reached at temperatures higher than this.
Please discuss and support your position, in particular . with reference to the capacity of the mate 7:ials to . accommodate pressure pulcas associated with potential reactivity accidents.
]
- b. Your discussion of brittle fracture is limited to the reactor vessel, whereas the fracture of any component or piping in the reactor coolant pressure boundary can initiate a loss-of-coolant accident. To reduce the probability of . this type accident, efforts to prevent brittle fracture should be applied over the er. tire boundary. Please provide information on the preventive measures you intend to take in terms of material selections and operating stress and temperature controls.
- c. It is prudent to have material in the reactor coolant pressure boundary at a sufficiently high temperature that it is in a fully ductile condition before nuclear neans are used to increase temperature and pressure of the reactor coolant. What provisions are included in the shoreham design i
l A5-38 L - - - -_
SNPS-1 l
to raise the coolant temperature by non-nuclear means during startup? -
RESPONSE 4.1a: In Amendment'4, SNPS' PSAR, Appendix G, Section 1, while discussing. the : station's conformance to the AEC 70 Criteria, the applicant states: " Based on the applicant's current understanding of the , intent of the proposed criteria,. it is felt.
that the. SNPS-1 nuclear facility fully satisfies- the intent of the criteria."
In Amendment 4, SNPS PSAR, ' Appendix G, Section 2.6, the applicant specifically elaborates on his interpretation of the intent. of '
the group-individual criteria and in particular conformance to Criterion 35.
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This applicant is not in disagreement but in full agreement- with l previously reported conformance. to the intent of the proposed. AEC Criterion 35 as cited in the following: j i
a) Bell Station, Unit 1, (AEC Docket No. 50-319) Amendrent 1, !'
' C/R 3.8
Bell Station, Unit 1, (AEC Docket No. 50-319) Amendment 3,-
C/R 4.8, 3. 3 I
b) Brunswick Steam Electric Plant, Units 1 and 2, (AEC Docket Nos. 50-324 and 50-325) Supplement 3, C/R 4.8 1 g ,J c) Hatch Nuclear Plant, Unit 1, (AEC Docket No. 50-321) Amend-ment 2, C/R IV-1.0, 1.1 d) Browns Ferry Nuclear. Power Station, Unit 3, (AEC Docket 50-296) No. I Amendment 2, C/R 1.1 Amendment 3, C/R 1.1 e)- Pilgrim Nuclear Power Station, Unit 1, (AEC Docket No. 50-293)
Amendment 2, C/R 1. 0 Amendment 9, C/R 4.0 Amendment 10, C/R 1.0 i Public Hearing Testimony, Section VII-3.0 f)
Cooper Nuclear Station, Unit 1 (AEC Docket No. 50-298)
Amendment 2,Section I i' Amendment 3,Section I Public Hearing,Section VII To prevent possible further confusion as to the applicant's interpretation of the intent of the proposed criterion 35, the tollowing is included.
A. CRITERION 35 - INTERPRETATION OF INTENT I
It is recognized that component or piping rupture in the reactor coolant
{ pressure boundary could initiate a loss-of-coolant AS-39 i
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SNPS-1 j
accident. l In this response we describe our interpretation of AE(' {
Y Criterion 35; present the technical basis for our interpretation: '
and identify those parts of the reactor coolant pressure boundar )
which are affected by the use of our interpretation.
e
- 1. On-Record Proposed AEC Criterion 35 0
t.
t The states wording of the proposed criterion 35 as presently published if "Under conditions where reactor coolant pressure boundar) .
t system components constructed of ferritic. materials may bed
.f subjected to potential loadings, such as a reactivity-induced loading, service temr:eraturess shall be at least 120 F 3
the nil ductility component material transition (NDT) temperature of abov the i if the resulting energy release .is j expected to be t bsorbed ofby plastic deformation or 60 F above j b
the NDT temperature the component material O if the resulting energy release is expected to be absorbed within i the elastic strain energy range."
Refer to Federal Register, July 11, 1967.
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- 2. GE-APED Interpretation of AEC Criterion 35 b The wording of AEC Criterion 35 is presently under revision by the AEC. GE-APED had previously suggested the following criterion for brittle fracture prevention: .
"The fracture of notch toughness properties and the operating {
temperature of ferritic materials of the reactor coolant pressure boundary shall be controlled so as to assure adequate ductility at the time of significant pressurization.
Such control shall be provided for any condition in which significant pressure is applied to a pressure boundary component whose failure would require operation of emergency core cooling systems."
Referred to also in:
a) Bell- Station, Unit 1, (AEC Docket #50-319) Amendment 1, C/R 3.8 l
1 In this Amendment a rewording of Criterion 35 is proposed, which will be the basis for further discussion. The reworded statement is as follows:
"The toughness properties of ferritic material and the service temperature of the reactor coolant pressure boundary shall assure
- 1. Fully ductile behavior e . g . , in the energy absorption region of 100 percent shear fracture, whenever the boundary can be pressurized beyond the systems safety A5-40
SNPS-1 iEC . valve setting by operational transients in' postulated
- n't . accidents; and
. ry
- 2. - A ductile to brittle fracture transition' temperature at least 60 F below the service. . temperature whenever the boundary can be pressurized beyond 20 percent of' its ae'd.
design ~ pressure . by operational transients, hydrotests, and postulated accidents."
ry This .has been referred to also ins-ba' a) Bell station, Unit 1, (AEC Docket #50-319) Amendment 3, Jedl C/R 48, 33
.ve I ha b) Brunswick Steam Electric Plant, Units 1 & 2 is' (AEC Docket #50-324 and 50-325) . supplement 3, C/R 4.8-Va-
- ha a. Hydrostatic test conditions need not be considered except for in' those components whose failure results in a condition requiring l emergency core cooling.
- b. Significant pressure is interpreted to mean approximately 20 percent . of reactor vessel design pressere and. for the case of APED reactors this corresponds to 250 psi gage, Dy c. The criterion applies to pressure boundary parts and does' not 39 apply to related components such as anchors, hangers,- 1
,, suppressors, and restraints.
l ng d. Adequate ductility at will be provided by maintaining a material temperature at least 60 F above the nil doctility transition re (NDT) . temperature for thick sections of
'a material. For thinner sections of material (less than one inch), the effect of
. :h increasing fracture toughness will be considered as described 'in TY Fig. 4 of Exhibit D.
- y
- e. The NDT temperature will be established by means of the Charpy V-Notch impact test (AS'DI A370 type A), or drop weight tests per ASTN E208.
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- f. Impact tests on plain carbon steel products will be required only if an NDT temperature of less than 190 F is required. This h
minimum temperature may be lowered as additional data on the NDT range.of presently used carbon steel materials is collected and t
a statistical approach can be used to demonstrate confidence in a lower limit.
.e. B. CRITERION 35 - PRESSURE BOUNDARY DEFINITION 11 The requirements for brittle fracture control shall apply to those components and piping of the reactor coolant system and associated auxiliary systems and emergency core cooling systems which are within the following boundaries:
\ Y AS-41
SNPS !
.1. . Reactor Coolant System and Associated A_uxiliary Systems '(
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(1) The reactor vessel and its associated control rod drive i housings and all other reactor. coolant pressure-containing components and piping of the reactor ' coolant system - which '
are subjected to reactor coolant ' system conditions during normal operation (including startup, hot standby and shut-down operations) .- The reactor coolant system boundary shall extend to the second in-series containment isolation valve in the main steam ahd feedwater lines.
(2) Those portions of associated auxiliary systems
- which are connected to the reactor coolant system, and which ,
extend to {
(i)
The second in-series containment ' isolation valve in piping that penetrates the primary reactor -containment, and through wh.i.ch reactor coolant flows or may flow l
'during normal operation when reactor pressure _ is 250 psi gage or greater (including startup, hor. standby ,
and shutdown operations), or l l
(11) The first block valve in the section of the system i
piping which is located within primary reactor contain- {
ment, and which is maintained closed during normal j cperation when reactor pressure its 250 psi gage or _j greater (including operations).
startup, hot standby, and shutdown I (iii) Exempted are lines whose size is such that in the postulated event of severance, the reactor system normal make-up capability ** coolant is adequate to supply water to the reactor coolant system at a rate at least equal to the rate of coolant loss for the interval required to permit an orderly reactor shutdown (or scram) and cooldown without requiring the operation of emergency core cooling systems.
- These associated auxiliary systems are those systems through-which reactor coolant must flow either continuously or intermittently to permit normal reactor startup, hot standby, and shutdown). operation (including to Normal makeup capability includes those systems which serve provide reactor coolant makeup either continuously or intermittently during normal reactor operation (such as may be provided by the pump capacity of the RCIC System) .
I AS-42 ,
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1 SNPS-1
- 2. Emeroency Core Cooling Systems (ECCS)
These- . portions of emergency core cooling systems
- extending from the connection with the reactor coolant system to the first system isolation valve ** (normally closed during normal reactor operation) which is outside the primary containment.-
- 3. Piping Systems 1' The piping systems included boundary for an SNPS-1 are shown in the reactor coolant pressure in Fig. 1.4.a-1. This figure identifies gage pressurethe minimum temperatures associated with the . 250 psi level discussed-in the previous paragraph. Based on this diagram and the GE-APED interpretation of Criterion 35, control of material fracture properties will be required on ferr* tic steel components of the in11owing systems from the reactor .to their appropriate isolation valves: q 1
(a) Reactor Feedwater, (
(c) RER, (d) Main Steam Line(b) . Reactor Cleanup Demineralized, Drain (e) Reactor Vessel Bead Spray andand Reactor Vessel Drain, Head Vent, (f) Reactor Core Spray Cooling, (g) CRD Return. Line, and the Reactor Cleanup ;
Demineralized, (h) HPCI, and RCIC inlets to the feedwater lines.
These systems are tabulated in Table 1.4.a-1 to shows the effect of wall thickness incomplete ~ and on NDT temperature requirements. This table is !
this station. does not show all line sizes and temperatures in f
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The sizes.
values shown in Table 1.4.a-1 may vary for different reacter Also indicated in Table 1.4.a-1 are the minimum temperatures at which the components in these systems may be pressurized above 250 psi gage. A minimum temperature of 70 F is used for those systems located within the drywell which are e ^jected to pressure under no-flow conditions. A minimum temper.
is used for those systems which may be te of 40 F subjec..d to cold condensate flow during RCICS or HPCIS operation.
Certain components subjected to which are located outside the drywell and are the minimum reactor building temperature ofsystem 40 F. pressure under no-flow co Portions of the main steam drain line and the reactor cleanup demineralized system which are outside the drywell must meet this temperature requirement. minimum as Emergency core cooling systems are those systems which one have, of their functions, the requirement to inject water into !
the reactor ccre in the event of any pipe break in the reactor coolant pipe) system from (including which the double-ended the loss rupture of the largest of reactor coolant is beyond the i system's normal makeup capability. i
\ System isolation valves identify those valves which serve to isolate the system from the reactor coolant system. !
A5-43
SNPS-1 The high at 250 pei be required. The gage that impact tests on the materialother will not the hydrostatic test pressure needsteam line will be 406 F at 250 psi gage and not be considered since a steam line rupture at hydrostatic test require emergency core cooling. The safety and conditions would not the HPCIS and RCICS lines will have the same temperature relief valves and pressure and these lines will always drain to the trap. conditions as the main ste C. CRITERION 35 - CAPACITY OF REACTOR PRESSURE BOUNDARj(
ASSOCIATED WITH POTP.NTIAL REACTIVITY ACCIDENTS Refer also to Bell Station, Unit 1, Amendment 1, C/R 3.8. (AEC Docket No. 50-319)
It is our position practical that NDTT engineering plus 60 F is an adequate limit for design purposes with (conventional) loading. static type in the 60 F margin is It is also fully our position that a reduction justified material. for thin sections of of fracture toughness and the inability of theOur BWR to position is justified b significant dynamic loading condition on the cause a pressure boundary. nuclear systes
- 1. Fracture Touchness IOTT + 60of eactions F or NDTT plus another temperature margin for thin in:.euded to besaterials used asasadiscussed in the Response Exhibit D is control limiting temperature for fracture has been wlde ,4.n the nuclear system pressure boundary .
This criterion loa 3ed structar2accepted for fracture control in conventionally for Nuclear vessels., and is the basic requirement ~ of the ASME Code The addition of temperature margin above this value,mechanisms, reactivity provides unneeded conservatism for the design since no resulting in loading, can be postulated for the BWR. significant dynamic type most conservative It is estimated interpretation of available data thatfrom the pressure pulse the is less than 100 psi.resulting Such ain pressure the most severe reactivity accident pulse is estimated to result in no significant dynamic loading.
The basis Puzak.* for the 60 F temperature margin is found in Pellini and "The origin of the ' classical' brittle fracture problem is the <
decrease in fracture toughness resulting from a chcnge in l fracture mode from high energy absorption ductile tearing to low energy absorption cleavage fracture in a rather narrow
- W.
S. Pellini and P. P. Puzak, Applying Laboratory " Practical Consideration in Design of Pressure Vessels," USNRL Report 6030, November , 1963.
5 Fract h5-44 i
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ShPS-1 ranga of temparaturen.
the order of 120 F.
Tha full spen of the transition is in i However, the transition span from an
'intermed iate' to. a 'very low' level of fracture toughness, which is the transition. range of engineering interest for conventionally loaded structures, occurs over a' range of 10 F to 60 F depending on .the stress level."
Therefore it is evident that our position is in agreement with Pellini.
- 2. Pressure Pulses With regard to reactivity I dynamic loading mechanisms potentially leading to i 1
it is estimated,that no mechanism canof the reactor primary system coolant) be postulated which can result in an abrupt pressure rise as large as' 100 psi.
Reactivity the produce mechanisms largest when the reactor is in the " cold" condition pressure transients. The following !
reactivity mechanisms in the BWR relationship to NDT and dynamic loading have been identified and their {
ascertained. None of these is ' estimated to result in pressure pulses as large l i
100 psi. (Refer as 3.0.) to Amendment 4, SNPS PSAR, Section XIV-2.0 and {
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Control Rod Withdrawal Errors I Control by operating rod operation procedures. is limited to one notch movement at a time 3 ips. Maximum Continuous reactivity rod withdrawal insertion rates is limited to 0.0019 Ak/sec.
to- reach This radioactivity insertion rate isdo insufficient not exceed j fuel enthalpies and heat transfer rates to the coolant capable of reducing pressure pulses (reference 1). ] '
(n) Cold Water Insertion 1
Any reactivity generation addition due to cold water insertion prior to the {
l as follows: of steam voids in the core would be small for reasons
{ 4 (1)
The small magnitude of the temperature coefficient of the reactivity, particularly in the low temperature range. !
(2) The recirculation and mixing time of any cold water inserted into the core.
The maximum reactivity addition resulting from cold water insertion would be comparable to in-sequence control rod i withdrawal i.e., a fraction of one dollar per second. No {
j pressure pulses can be initiated by means of this mechanism. i (C)
Reactor Coolant Recirculation System Malfunctions When the reactor malfunctions is in the " cold" condition, recirculation systen result in minor reactivity additions due to the inherent natural circulation capability of the boiling water reactor.
A5-45
unsa e
-(D) Turbine control /Stop valva Closurn No turbine _ valve . manipulation occurs prior to boiling of the i reactor coolant.
(E) Control- Rod Drive Malfunctions or Failures The only reactivity insertion of interest, .by this mechanism, relates to separation of the drive housing from the reactor vessel. Control rod movement. 's afficient to reactivity. into the insert . measurable housing support.
core is prevented' by the control rod drive (F) Control Rod Drco Accident l The - control rod velocity limiter prevents " free f all" velocity of a rod separated from its drive from exceeding 5 fps. A . control j rod . drop velocity of 5 fps, even u
-* existinc, in the core at any time duringfor the maximum worth rod a normal startup, will I result in peak fuel enthalpies Experimental excursion data have below the fully molten fuel point.
[ demonstrated that the heat transfer rates from f uel rods to the surrounding moderator are p] not. sufficient to cause a significant pressure pulse (greater q than 100 psi) if the enthalpy is below the. molten value. A
'g control rod worth minimizer is provided as. a backup to reactor h operator procedures prohibiting withdrawal of high worth rods.
J (Refer to Amendment 4, SNPS PSAR, Sections VII-4.0, VII-9.0, and.
XIV-3.1.)
. 31 f
Uncertainties associated with the reactivity coefficients used in the analysis of reactivity transients are discussed in Dresden
' Nuclear Power Sta tion, Unit 3 (AEC Docket Amendment 3. No. 50-249) d cold, This report discusses hot standby conditions. At low
]j pressure conditions, there is less uncertainty in the value of of the coefficients due to availability of data. Examples 1 these. data are the cold SPERT I oxide core excursion data and J temperature coefficient measurements made at low temperature and pressure at all operating BWRs.
A complete discussion of the methods, criteria, uncertainty analysis and evaluation of the consequences of reactivity insertion by control
, listed below. rod motion is contained in the references 4
a e
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SNPS-1 REFERENCES l- .
(1) H. A. Brammer, J. E. Wood,. R. J.
McWhorter, Nuclear Excursicn Technology, APED-5528, September, 1965.
(2) J. E. ' Wood, Critical Reactivity Analysis Methods of Hypothetical Super-Prompt Transients ~ in {
APED-5448, April, 1968. Large Power Reactors, j (3)' J.
E. Boyden, ~ H. A. Brammer, J. E. Wood, Summary Memorandum on Excursion Analysis Uncertainties, Dresden Unit 3 PSAR, Amendment 3.
(4) R. A. Dietrich, The Mechanical 1
Effects of Reactivity -l Transients, APED-5455, January 1968.
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SNPS-1 TABLE 4.1.a-1 TEST TDtPERATURE REQUIREMENTS _. FOR PIPING i
Piping Wall Minimum Temperature 60 F Margin NDT System Thickness At Pressurization Equiva lent - Requirement Feedwater 0.843 40 F NDT + 44 F -4 F Steam Drain 0.315" 40 F NDT - 8 F. 48 F Clean-Up 0.366" 40 F NDT - 3F 43 F RHR 1.11" 40 F NDT + 60 F -20 F Core Spray 0.593" 40 7 NDT + 20 F 20 F Head Spray 0.432" 70 F NDT + 4F 66 F Head Vent 0.218" 70 F NDT - 18 88 F Bottom Drain 0.218" 70 F NDT - 18 88 F 1
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SNPS-1 RESPONSE 4.1b: Please refer to this Amendment, for this requested information. Response 4.1.a
[
RESPONSE 4.1c Please refer to this Amendment, Response 4.1a for reference information.
General Electric Boiling Water Reactors (BWR) are never pressurized- at temperatures below -the ductile ranges.
discussed in Amendment 4,Section IV. 2. 3. 3. 3, SNPS As temperature is based upon PS&R, this the Nil Ductility Transition Temperature data available.
(NDTT) of the material plus the worst case NDIT shift To the initial temperature and most pessimistic leTT shift data is added a margin of 60 F. This temperature margin is adequate for. statistically loaded structures.
In addition, increase it should be noted that a BWR inherently requires an in temperature, wall abcVe
- he range of concern with regard to this question, in order to achieve pressurization. No equipment is provided which could pressurize a BnM vessel during startup prior to attaining a temperature well above the NDTT limit. Ther=. fore, no provisions are required for raising the coolant temperature by non-nuclear means during a startup.
Normal startup from a cold condition would involve nuclear heat as well recirculation coolant as a mechanical heat addition to the coolant via reactor system pump operation. Reactor vessel temperature and pressure would follow the steam saturation curve.
( l The NDTT of those parts of the reactor coolant pressure boundary which do not follow the reactor vessel temperature is controlled so as to assure adequate ductility before significant pressure in applied.
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SNPS-1 1
COMMENT 4. 2 : j The importance of periodic inspections detecting and preventing impending failure of in 'j high pressure systems is generally acknowledged.
The design provisions and plans for in-service ,
j inspection of the Shorehaa primary coolant system are not presented in j any detail in the PSAR.
What areas cf the reactor coolant system wi.11 be inspectable periodically during the life C the plant and to what i
accessible? Do' cheseextent will these areas be areas include the more highly stressed areas in the system and the areas most reactor coolant subject to cyclic )
stresses? i l
Please list those areas of your, proposed inservice inspection j plan which do not meet the proposed
- requirements of the " Draft USA Standard, code for Inservice Inspection of Nuclear Reactor Coolant Systems, October 1968" (USASI-N.45). Discuss the i
reasons wh/ you do not propose to meet the. {
requirements, or why such inspections are not 1 feasible. !
RESPONSE
reactor Neutron vessel adjacent fluxtodosimeters. willatbe the ve'ssel wall theinstalled in the I level. The core midplane I dosimeters will be removed and tested after 1 yr of.
operation to verify 'l values experimentally and adjust the calculated of integrated neutron flux that are used to determine the j NDTT. i 1
' Inspections will be made as specified in Table 4. 2-1 during the first 10 years of operations. A preservice inspection will be done for after later installation at the site to provide a base reference inspections.
investigated, evaluated and recorded.
Indications of a defect will be A visual examination for 1 taks will be made with the reacter {
coolant or after system majorat repairs pressure during each scheduled refueling outage j system. have been made. to the reactor coolant j j
A minimum of one half of the safety valves will be bench checked or 1
I replaced to ensure setwith a bench points are as checked valve at each refueling outage, specified.
A sample of reactor coolant will be analyzed at least weekly to determine the content of chloride check the conductivity. ion and Iodine-131, and to Numerous '.ata are available relating integrated flux and the change in Nil Ductility Transition Temperature (NDTT) in various steels .
integrated Theflux mostat conservative data will be used. The measured the vessel wall physics data and will be measured using flux monitors rrom is calculated installed core i
AS-50 )
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inside the vessel. The measurements of 'the neutron finx at the vessel wall will be used to verify or correct the calculated data to determine an accurate NDTT.
Due to the design and construction techniques used, the reactor primary system will be free of gross defects prior to service and gross defects should not occur throughout the station life. This program is designed to detect gross defects in the unlikely event that unusual corrosion or stress cycling causes them to form.
The inspection for leaks will cover all points of all systems inspected. The nondestructive test program at sample locations
~
will detec' lipient defects that may occur generally throughout the systec. a inspection period is based on the growth rates for defe tr. 26 4erved in ABC sponsored fatigue studies which show that it .eio i thousands 'of stress cycles, at stresses beyond any cor. v '.v .st e in a reactor system, to propagate a crack. If a leak oce. /, c will require weeks for it to grow large enough to be a har.i? It is considered appropriate to evaluate the results obtained from compliance with this technical specification and the state of the art before establishing a longer-term inspection program.
The type of inspection planned for each component depends on location, accessibility, and type of expected defect. Direct visual examination is proposed wherever possible since it is sensitive, fast and positive. Ultrasonic testing or radiography are added where defects can occur on concealed surfaces.
conductivity instruments continuously monitor the reactor coolant. Experience indicates that a check of the conductivity instrumentation at least every 72 hrs is adequate to ensure accurate readings.
The reactor water sample will also be used to determine the chloride ion content and the Iodine-131 activity to assure that the limits of reactor coolaat qua13 ty are not exceeded. The chloride ion content and the Iodine-131 activity will not change rapidly nver a period of several days; therefore, the sampling frequency is adequate.
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I t SNPS-1 TABLE 4.2-1 (
EXAMINATION SCHEDULE OF REACTOR C00LAlfT SYSTEM l
l Inspection Inspection Coeponent Sample Extent Process Frequency See Note (1) See Note (2)
Reactor Vessel Flange Studs All Top end UT & VT b steam nozzle one a. entire nozzle to volumetric a shell weld & VT
- b. entire este.rior VT a nczzle surface
- c. entire nozzle to volumetric a pipe weld & VT Control Rod 10% of total Interior circumfer- UT c Drive penetra- selected ence opposite tions housing to stub tube weld Recirculation outlet nozzle one nozzle a. entire safe end volumetric a nozzle weld & VT
- b. entire safe end volumetric pipe weld & VT Circumferential one 10% of weld length volumetric a weld-head to including 2 & VT head flange intersec. with longitudinal welds Longitudinal one Entire length Volumetric a weld on head, & VT from flange weld to cap Piping System in Drywell Recircu- 105 of cir- Entire circumference Volumetric a lation, cumferential & VT main steam welds greater and core -
n 4" in spray utas in one loop of each of the 3
, systems A5-52
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Ultrastnic Examination
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Radi Jgraphic Examination' / '
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L' id Penetrant Examination i.
VT -
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Visual Examination ,
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volunetric - UT or RT ,4 7 Note (J) a. Inspect same sample twice during,firht 5 yrs of operaticu., '
b b.
100% inspedtepartial sample during at least two inspectic4n such that 100% of the studs are inspectec 6aring the first 5 yrs of operation.
[' r t c.
Inspect partial sample during at least two
' inspections such that 10% of the penetrations are
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, inspected during the first 5 yrs of operation.
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COMMENT 4.33 ; _ What' loading, combinations will be used- -in the
- design of . tne _ nuclear steam supply system, and what are the proposed. rtress and deformation -
limits 'for each combination?- How do' these- '
proposed design criteria compare' with- the:
criteria for_ emergency and fault _ conditions recently approved' by the ASME. Section III Committee?
1-L RESPONSE:
Please: refer to Amendment 4, SNPS PSAR, Appendix D for -
the " Loading Criteria" used is: the design of nuclear' steam supply system. (NESS)..
As indicated in the Appendix D (referred to the "GE Imading Criteria"), the si. cess and deformation limits given are quite similar to -those, recently approved - (Summer 1968) addenda to Section III of- the ASME Boiler & Pressure Vessel Code..
There are some minor differences. some new concepts used in. GE Loading Criteria do not appear in the ASME criteria. The principal additional GE concepts are (1) the use of a- variable safety (Pgo) of factor which depends the hypothesized uponorthe event 40 yr encounter probability events ao that a - rational procedure for event classification can be carried out, (2) the inclusion fatigue of specific deformation limits, (3) the inclusion of exemptionlielts in thewhich ASME III eliminate rules. the need for the 25 cycle The minimum safety factors for emergency and' fault conditions .are found by using (f aultad) respectively. . GE Loading Criteria indicated, Pg , = 10 8 and P40 *
- faulted conditions, P40 = 1.5 x 10-*. However, in that for the carrying out the load capability analysis, General Electric recommends-. load capability . evaluation at an unreliability index no greater than
! 10-1, so that even trough the numeric safety factor corresponds to that for P4 0 = 1.,5 10-* ,
is never less than that corresponding to Pthe cquivalent desian safely factor 4 0 = 1.5 10-5 As specific cxamples with Pgo = 10 > (emergency condition) under the GE lisit, a 50 percent increase primary stress is permitted, which is above ASME III elastic exactly the same as permitting 100 percent of the tabulated yield strength as is done in the recent ASME III revisions.
For stainless materials, ASME same III permits only a 20 percent increase while GE permits the j 50 percent increase as for other materials. This is based on attempting to preserve a consistent approach used in ASME III i of permitting higher allowable stresses relative to the
- 0. 2 percent offset yield for the strongly strain hardening materials.
should cause no conce n.
Since deflection limits are in the GE criteria, this ,
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The GE limits permit a 50 percent increase in buckling loads under emergency conditions while the ASME III limits are only 20 percent greater. The GE limit is 50 percent increase in primary stresses. The consistent with the GE limits for fault conditions permit higher loads when using limit analysis than the A5-54 i
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SNPS-1 I ASME 17.I revisions and lower limits when usi.ng the plastic inst. ability analysis. These changes were made becausc the suggested ASME III limits are inecnsistent with each other, the !
limit procedure being too conservative . and the plastic I instability analysis not being conservative Enough.
The GE limits also permit more optional me'Jods for both !
anergency and faulted conditions which are equivalent to the limits used in the ASME III procedure.
The General Electric Leadihg Criteria (PSAR Appendia D) are intended to supplement, not replace, the ASME Code Section III.
In the design of the reactor pressure vessel, the design limits used are those of ASME Code Section III or the Gener&l Electric Loading criteria, whichever .are more restrictive.
In the design of reactor internals, the General Electric Loading criteria are considered to be appropriate and adequate to assuee safety and structural reliability.
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T SetPS-1 COMMENT 4.4: ' You ' have indicated that - two isolation valves ,' in 1- '
series will be provided in the shutdown cooling lines to the RERS to separate the low ' design- 3
' pressure system outside of containment from the high pressure in the reactor coolant system during normal operation. If both of these valves are postulated to be opened -during reactor operation, due to a failure,- operator' error or
[ otherwise, the low design pressure system would almost certainly be ruptured,- resulting in an uncontained loss-of-coolant accident. What provisions, other than administrative control, will be included in the design in order to )
minimize the probability of such an accident?. .
1 RESPONSE: The ECCS are designed in conformance to the intent. of the AEC proposed 70 design critieria and in particular the RHRS in conjunction with the r other'- ECCS designed to is conform !
specifically to the intent of criteria-19, 20, 21, 22, 23, '24 ,
25, 26, 38, 4 0, 4 4, 45, and 46. -
The criteria stated above require analysis for a single (one) I component failure in order to assure design performance. Thus, (
the failure .of two independent isolation valves in the RERS (Reactor Shutdown Cooling Subsystem Mode) is not required .and will not be entertained in this response. ]
Examination of. the RHRS (Reactor fanutdown Cooling Subsystem) mode !
equipment in detail ensures conformance to the above criteria:
, A) Isolation valves - General l Please refer to Amendment 4, SNPS PSAR, Fig. VI-2-6. It should be noted that both the inner isolation valve (MO-F009) and the outer isolation valve (MO-F008) of the primary containment system are ' closed" during normal power operation and are designed for rated reactor primary syrtem pressures, temperatures, etc., and the drywell design of 48 psi gage, (296 F), 100 percent RH etc.
All control and status indit:ations for this subsystem are in the station main control room.
B) Isolation Valves - Control I Refer to Amendment 4, SNPS PSAR, Fig, VI-2-6B and VI-2-7A through I
- c. It should be noted that both valves are keylocked " closed" and have position indication on the reactor console of the keylock position. It should also be noted that the valves are interlocked to prevent their inadvertent manual "openino" by the reactor operator. Interlock signals from a redundant set of high reactor pressure instrumentation channels or from the edundant, two phenomena ECCS initiation signals of high drywell p. .ssure or low reactor water level or the keylock " closed" switch signal prevent the valve openings. The position of the valves while in l
l AS-56 )
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SNPS-1 the keylock "opon" switch position is also shown on console. the reactor C)
Isolation Valve Power and Control-Power Sources Refer to Amendment 4, SNPS PSAR, Fig. VI-2-7A through C, VIII 1 and VIII-4-1. It should also be noted that the isolation valves are supplied by two different-type independent power sources, Since each of which is, to some extent, redundant in itself..
the valves are closed, they fail-safe, any loss of either/or both, control or power energy, to valves, while the either/or both reactor is at normal power operation, would result in each valve remaining in its quiescent position, that is
" closed".
The above failure analysis shows that the present design far exceeds the requirements of the ABC.
of 4
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5.0 CONTAINMENT SYSTEMS 1
COMMENT 5.1: Will the negative pressure and leakage (0. 5 -'in. - rate volume perof- day) water and 50 percent of the building for the secondary containment building, (page XIV-3-9), as used in. the accident - analyses also be used as design.
specifications and acceptance criteria for the reactor building and standby gas treatment system? How will the secondary containment '
building leakage characteristics be verified by initial and periodic testing?
RESPONdE The design specification' for construction. of the secondary containment will be based upon an initial leakage _ rate much lower than the 50 percent per day value used in the accident analyses.
j The basis for these construction specifications will-1 be the leakage rates for _ construction systems as reported in i
" conventional Buildings for Reactor Containment," AEC Research and Development Report, Publication NAA-SR-10100.
It is . recognized that experience is meager in the design, construction and testing of low leakage secondary containment structures of this size and type.
to Furthermore,'it is difficult assess accurately the extent to which the completed structure will be below the maximum allowable leakage rate. As a result, the design criteria for all penetrations and joints of the .)
j secondary rate containment as practical. will be based on obtaining as low a leakage )
-1 The acceptance criteria for the secondary containment, as is' the case 'for the primary containment structure, will be based on demonstrating with an initial leakage rate test that the leakage rate does not exceed the design value used for accident calculations. For the secondary containment, the acceptance criteria correspond to a maximum 50 percent per day leakage rate at 0.5 in. water differential pressure and calm wind conditions.
The initial secondary containment leakage rate test methods, as presently envisioned, will not differ substantially from methods used for periodic test !
leak rate testing. The detailed procedure for testing the reactor building will be developed during the fiel design phase of the engineering for the station containment systems. The basic method of leak rate testing the reactor building will be as follows:
1
- 1. The test is initiated under conditions of low wind speeds.
Wind speeds up to 10 mph (measured up to the rooftop of the reactor building) will, at most, have a 10 percent effect on the l i
l-leakage rate since the velocity head for 10 mph is only 0.05 in.
water and the minimum reactor building internal pressure will be D.50 in. water (negative). Furthermore, it ,
conduct the test at low wind speeds since thisis desirable to 2 is the condition which corresponds to the maximum inleakage rate. i l
AS-58 1 ,
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The' reactor building normal ventilation system is and the secondary containment isolation valves are closed shut off 3.
The either: reactor building inleakage rate can then be determin (a) activating ..
e y- d b system and the measuring both reactor building standby ventilation stabilized the avera pressure which is out of the building; or (b)within the reactor building,ge for utilizing a and the flow rate of air separate blower system exhausting the reactor building to- a prensure level of
-0.5 in. water is' required forgage and measuring stabilizing the internalthe blower discharge rate whic this value. building pressure to-n A5-59 l
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SNPS-1 COMMENT 5.23 What 4,re the potential consequences of losing both coolers in any of the primary . containment hot pipe penetrations? Wha
- monitoring provisions will be included to inform the operator if any of these coolers fail? (you say wish to correlate this answer with that for item 12.13c) ,
RESPONSE: The response to this comment will be found ir. Section V-2.3 of Amendment 4, Shoreham Huclear Power Station Preliminary Safety Analysis Report.
AS-60
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4 SNP3-1
{' COMMENT 5.3:- In general, current power reactor designs' -provide l ~t two barriers against leakage for every _ primary
(
- containment penetration. These barriers usually provided either ty closed systems or by '
are automatically actuated isolation valves.
However, any pipeline which penetrates containanent always includes a valve- (not necessarily an ' automatically actuated ~ valve) ,
outside, and close to, the containment which can be closed by operator action. This means that .f such a valve is provided even if the- line is .j l
connected to " closed" systems both inside and outside containment and in those pipelines ubere two check valves could serve as isolation valves because flow is normally into the containment.
This ' is to provide the operator with capability for corrective action if a leak should develop through such a penetration during an accident.
11 ease discuss whether your design criteria - for containment isolation valves will follow this policy. If they do not, identify those penetrations which do not provide the operator with such 'capabili+y and discuss why you think these are acceptable.
RESPONSE: The general criteria governing automatically actuated
' primary containment isolation valves are presented in Amendment 4, SNPS PSAR, Section V-2.2.4 and described in Section V-2.3.4.
(
- Those lines listed as exceptions to the automatic closure criteria will have, however, capability for closure by operator action.
For cases in which two check valves are to be used for isolation, one of the valves will be of a stop check design or an additional, manually activated, valve will be installed outside of the primary containment. Additionally, as stated in Section V-2.2.4, manually activated isolation valves will be provided for primary containment penetrations of lines connected to " closed" systems.
In sum, the general policy governing primary containment isolation valves will enable any line connected to the primary containment to be closed by operator action. For automatic activated isolation valves, provisions will be made to ensure that the preprogrammed logic of the safety circuitry is not altered during an accident by an erroneous operator action.
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AS-61
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COMMENTi5.4:- What seisnic design criteria will the design be applied in '
and procurement (since this may have to precede final design) of primary ' containment isolation valves?
RESPONSE: The primary containment isolation valves designated as Class I equipment as stated are Section III-2.1. in PSAR Am'end . 4, The specific applit able design criteria for
-XII-2.1.
Class I equipment are listed in 2mendment 4, SNPS PSAR,Section I
i A S- 62
SNPS-1 COMMENT 5.5: Will all primary containment penetrations,.
.in- hatches, access locks, anchor stud seals or the vs cont.ainment - vessel head which have resilient at. seals, gaskets, sealant compounds or expansion bellows be designed to permit periodic local ;
leakage testing during reactor operation at 100 re percent containment design pressure? If not, 4, will the design of the primary containment permit 3r performing annual integrated leakage tests during an the anticipated life of the plant?
RESPONSE: The response to this comment will be found in Section V-2.5 of Amendment 4, Shoreham Nuclear Power Str~. ion Preliminary safety Analysis Report.
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AS-63
SNPS-1 COMMENT 5.6: What R & D effort will be - associated with completing the " preliminary" design of the primary containment drywell floor-to-wall seal?
To what extent will this seal be subject to preoperatior.al and periodic postoperational testing? Please provide additional detialls on the analysis of the consequences of drywell-to-suppression chamber leakage referred to on page V-2-43 of the PSAR. 'How would the results of this analysis be changed if steam rather than water were assumed to leak from the drywell into the suppression chamber?
RESPONSE: The response to this comment will be found in Section V-2.4, Amendment 4, Shoreham Nuclear Power Station Preliminary safety Analysis Report.
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A5-64
SNPS-1 i
k 1 l COMMENT 5.7:- What are the ' design criteria and preliminary- l I
( ) designs' for the primary containment electrical !
l penetrations and control rod drive hydraulic line 1 penetrations? ;
RESPONSE: The response to this comment will be found.in section
(- V-2. S(Fig. V-2-10) of Amendment 4, Shoreham Nuclear Power Station l'
Preliminary Safety Analysis Report.
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a SNPS-1 l COMMENT 5.8 Please provide an outline of the basic equations and assumptions for the LOCTVS computer code which was used to calculate the vapor suppression characteristics of *the Shoreham primary containment system.' What are the values of the input data to the code used for the design' basis loss-of-coolant accident calculation? What~are the significance and implications of using a
" homogeneous blowdown model' in correlating the LOCTVS code with the Bodega Bay test data and the
' Moody blowdown model" for the transient analysis for Shoreham? (pg. V-2-40) To what extent have results obtained using; the Ib.".rVS code been checked against the calculational results obtained using the corresponding GE code for a BWR vapor suppression system?
RESPONSE: The response to this comment will be found in Section V-2.4, Amendspent 4, Shoreham % clear Power Station Preliminary safety Analysis Report.
AS-66
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SNPS-1 I
i COMMENT 5.9: Please explain how you determined that 2.6 sq ft corresponds to the maximum potential break size i j
for the design basis loss-of-coolant accident. l i
RESPON5E: In order to establish a design basis for the i
suppression pressure containment with regard to pressure rating and steam condensing capability, the maximum break size of the reactor {;
primary system is defined. For this design, an instantaneous, circumferential break of one reactor coolant recirculation system )
line has been selected as a basis for determining the maximur {
gross drywell pressure and the condensing capability of the )'
pressure suppression system. For design purposes this failure condition has been selected to establish the containment parameters, but the failure mode and the magnitude of failure is still assessed as being incredible.
Also, in establishing the containment design, circumferential pipe breaks are assumed with sufficient distance separation to allow full potential flow from each end of the pipe. Pipeline flow restrictions are not considered in establishing break flow rates.
Since the assumed initial break rate and the accompanying reactor depressurization is so rapid, progressive f ailure of the piping is not a limiting factor in the containment design.
The break area assumed (for the purpose of calculating the ce.tainment peak transient pressure and establishing the break vent area ratio) was 4.34 fta. This is equivalent to the total area of the ten (10) jet pomp injection nozzles, and the recirculation suction line area. In calculating the peak pressures no credit has been taken for pipe friction or for the pump and flow nozzle resistances. 3 The 2.6 fta break corresponded to the maximum break for a i
standard 1665 MWt (540 MWe) BHR with no equalizer line. Further discussion and data for the present 2,436 MWt BWR will be found in Amendment 4, SNPS PSAR, Section V-2.4 and Table V-2-1. l I
l AS-67
l l SNPS-1 COMMENT 5.10: What are. the possibilities and potential consequences of local pressure buildups within (
cavities inside the containment drywell (i.e.
within the reactor shield wall or support l
pedestals) during a loss-of-coolant accident?
RESPONSE: The field welded ends of all pipes leaving the vessel l tie outside the thermal shield wall. Since pipe breaks are l l
considered only up to the field welds, blowdown into the annulus between the shield wall and pressure vessel cannot occur, therefore, a local pressure rise inside the annulus cannot occur.
l Below the vessel under the support pedestal there are no large pipes which could potentially blow down. Blowdown from instrument tubes on the control rod drive penetrations is insignificant pedestal.
relative to the vent area present in the support The drywell chamber will have no enclosed areas in which local pressure buildups could occur.
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SNPS-1 COMMENT 5.11: . ase- provide the details of how the
) flow-m e sistancer for the drywell to vapor suppression j
chamber downcome,r vent pipes.3 are' calculated for the Shoreham d'e sign, how this pressure drop for the. Shoreham design compages to other vapor suppression system designs and . how sensitive the peak containment pressure calculated is to this-parameter. ,
i' RESPONSE: The response to this comment will be found in Section V-2.4, Amendment 4, shoreham Nuclear Power gStation Preliminary Safety Analysis Report.
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COMMENT 5.12: How was the possibility of flooding and ch'oking of
- the drywell-to-suppression cha9h ar vent pipe i
inlets considered and provided for in the design?
What is the basis for the distance specified for the vent pipes' protrusion above the drywell.
floor?
, RESPONSE: The basis for the distance specified for the vent pipes' protrusion above the drywe13 floor was that beight considered adequate to preclude clogging of the vents by debris either inadvertently lying on the drywell floor or generated during the LOCA.
The possibility of flooding and choking of the vent pipe inlets .
by a directed stream of water is precluded by the vent deflector plates provided for each vent pipe. Accumulation of water on the drywell floor, due to unflashed reactor vessel effluz and condensing steam, is slow enough to preclude flooding of the vent pipes during blowdown.
Af ter the blowdown phase 'of the LOCA, water overflow from the reactor vessel due to ECCS operatien accumulates on the drywell floor. When the level of water on the drywell floor reaches the height of the vent pipe inlets, water will overflow the vents and return to the suppression pool. As a result of this recycle of '
water back to the suppression pool, a constant supply of water is assured for the ECCS pumps during the longterm phase of the LOCA.
Flooding of the downcomer vents during this time period is of no consequence since the flow of noncondensables or steam through the vents is negligible.
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A5-70
9 SNPS-1 COMMENT 5.13: What are the criteria and bases for the design of the suppression chamber to drywell vacuum breaker valves? What is the preliminary. design for these valves?
RESPORSE: For the Shoreham Station Primary Containment design, '
the valves are required primarily to limit the. net noward force on the drywell floor to below the structural' limits rather than to prevent excessive vacuum conditions in the drywell.. The major design criteria for these valves are as follows:
1, The valves will be capable of operating under the drywell design pressure and temperature conditions as well as the differential pressure (downward) rating for the drywell floor.
- 2. The flow characteristics and total flow area of- the valves will limit- the marimum upward force on the drywell floor to within the structural design limits.
- 3. Failure of any one valve, either in the closed or open position, will not permit the containment design conditions to be exceeded.
- 4. The valve design will prevent leakage of steam directly the suppression chamber.
The basis for design according to the above listed criteria will j be ' the loss-of-coolant accident utilizing the IOCTVS computer code. Analyses' will be made to ensure that the valve flow area- <
and pressure loss coefficient are adequate to limit, during long- !
term phase of the IOCA, the net upward force on the drywell floor to a value less than the dead weight of the floor.
The design of - drywell suppression chamber vacuum breakerc employed in the lightbulb-torus type of containment structures has been used as guidance for the shoreham preliminary design.
At present a final design of the vacuum breakers has not been adopted for the Shoreham Primary Containment.
i A 5- 71 I
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SNPS-1 COMMENT 5.14:
The discussion in the PSAR on the basis for not including baffles in the suppression chamber (pg.
V-2-44 and 45) refers to some "one-quarter scale tests." Please provide a summary description of these tests and the results that were obtained or indicate by reference where this information is available.
- i
RESPONSE
in Information on the 1/4 scale test was first submitted the Preliminary H;zards Summary Park, Unit 1, Pacific Gas and Electric Co. Report, Bodega Bay Appendix Atomic 1, (Docket 50-205), Dec. 28, 1962.
Further information was submitted on AEC approved designs for Quad-Cities Station Plant Design Analysis, Units 1 and 2, Commonwealth Edison Co., Amendment No. 5, (Docket 50-254) Public Power District, Amendment No. 1, (Docket 50-298) .
Additional information is hereby submitted in the form of a report, Exhibit E, from the Pacific Gas and Electric Co." Pressure Suppression Parton dated Tests Octoberwith 25,Multiple 1963. Vents"by C. P. Ashworth and D. B.
This additional Electric Plant,information Units 1 & 2, was also submitted on Brunswick Steam Supplement 3, C/R 5.14. (AEC Docket Nos. 50-324 and 50-325)
A5 -72
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t COMMEttr 5.15: The . discussion in the PSAR on the design against e damage, by missiles. (pg.
V-2-60) and the referenced Davenport , paper cite a 100 ft/see missile velocity as a " conservative value" . What i is the basis for this - judgment?
! RESPONSE: The type of- missile considered is a valve stem that fails and is accelerated through its packing tube. Based on the size of the valves to' be used and the driving pressure available, 100 fps is considered to be an upper boundary for the velocity of ^
- this missile.
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SNPS-1 COMMENT 5.16: Please' provide a more d3finitive description of the power sources to the primary containment
- isolation . valves to support your conclusion '(V -
Its, . subparagraph A. 3) that no single failure could interrupt power on both isolation valves in any penetration.
i RESPONSE: l Where two isolation valves in series are provided, the i
valve motive power system will be designed in accordance with one of the following:
- 1. valve operators requiring electric power to close the_ valve, will each have the operating power supplied from 'a separate source to j valve to maintain the required reliability. The power supply may be from separate standby diesel generators or it may be one a-c and one d-c source. - Direct current control power will be supplied from two separate batteries.
- 2. Spring loaded air operated isolation valves may be designed to fail closed be from separate in which sources.case the operating power does not have to AS-74 l
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SNPS-1 COMMENT 5.17: Provide your assessment. of the problem of hydrogen ! I g accumulation in the primary containment'following a loss-of-coolant' accident due to a : metal-water i
' reaction and/or that g merated' by the radiolytic '
decomposition of cooling water. Your analysis i
- should consider the effects ' discussed in BMI-1844 - i on the amounts of hydrogen and oxygen released due to . radiolysis and_ any experimental data-available for gas . evolution due to the radiolyt!.c decomposition ~ of water that_ is being sprayed.
Provide.a description of any method (s) proposed to control the post-accident accumulation or hydrogen in the Shoreham containment.
RESPONSE: A discussion 'of metal-water reactions is presented in section V-2.4.3, Amendment 4, SNPS PSAR. The analysis shows that the primary containment has the capability- to withstand a 67 percent metal-water _ reaction. However, the maxistun anticipated reaetion- is . orders of magnitude less.
A' complete evaluation of the various postulated -loss of coolant accidents indicates that metal-water reactions might result in a hydrogen concentration of 0.1 percent by volume (Refer to GE/ APED Topical Report No.'5454, " Metal Water Reactions--Effects on Core !
Cooling and containment," Merch 1968, and GE/ APED Topical Report No. 5654, " considerations Pertaining 'to Containment'Inerting,"
g-
' August 1968.) Preliminary estimates of the concentration which might develop from radiolytic decomposition are on the order of 4 percent by volume. The applicant is aware of recent oral ;
statements made by others on the subject of radiolytic '
decomposition and on the possibility that higher volume percent of hydrogen accumulation might occur in cercain reactor types.
Because the applicant is not aware of' the documentation of' this j i
information, the applicant is unable to make a statement of' I position of how this information might affect containment designs l for the facility. i d
The~ General Electric Company has pursued and will continue to pursue its investigations of the possibilities of hydrogen generation in containments following loss-of-coolant accidents by both analytical and experimental techniques. ,
t In response to questions on this subject recently in other dockets, the following responses are given in order to supply additional information to avert future requests in this area.
~
- 1. There are various potential sources of hydrogen following a postulated design basis loss-of-coolant accident, i.e., 00 2 -water reactions; Zircalcy-water reactions; radiolysis. Radiolysis is one on which the least information exists. Of the potentially significant sources, the 00 minimal.
2 - H 2O reaction has been shown to be The Zircaloy-water reaction, although a potentially large source of hydrogen, has been reduced to an insignificant
( quantity by the in-depth emergency core cooling systems networks AS-75 i
e I
SNPS-1 which have been provided on BHR's. Other sources such as corrosion and original reactor water inventories have been shown 4
to-be negligible.(1)
Theoretically, the quantity of _ radiolytic hydrogen should be limited to some equilibrium concentration by the gamma-induced back reaction resulting from the presence of hydrogen.<a>
i Experimental confirmation'of this has been obtained at ORNL from '
irradiation of capsules, but the exact value of the equilibrium has recently been found to vary over a wide range, depending ,
primarily on the solubles in the solution.ca) This has caused concern because the equilibrium values from these tests would be well in excess of the flammability limits if directly applied to a primary containment.
The pressure suppression concept, because of its very large heat sink and inherent ability to vent high energy rates into the pool, can accommodate- the energy release from l very large quantities of hydrogen. This is true whether it burns as evolved, does not burn at all, or burns after it has evolved if ;
burning occurs at a rate no faster than a few seconds. Below 4 percent to 6 percent by volume the hydrogen cannot burn, while
.above this region, it can burn, but should not detonate until considerably above this level. Primary containment inerting offers no solution to this concern because the radiolytic hydrogen is formed stoichiometrically with o::ygen, thus, il providing the necessary oxygen in sufficient quantities for the respective postulated reactions.
Therefore, the concern is that if hydrogen formed by radiolysis should equilibrate above the 4 percent to 6 percent concentration by volume, and if detonation were possible, the containment integrity could become questionable. ,
l The only published recent work on radiolysis is that by ORNL.C3)
This consists of capsule irradiations in which the hydrogen l 3
overpressure is measured as a function of integrated flux from a cobalt source. The amo'unt of over pressure at equilibrium ranges
{
from as low as 5 psig for pure water to over 30 psig for l solutions containing sodium hydroxide. There are two problems with these data. First is the key question as to whether capsule data such as this can be applied to a reactor containment system complex. Secondly, the conditions of interest to a BWR have not been tested, although it would be expected that the results for
)
pure water would be the more nearly applicable. The capsule :
data, if they apply, would indicate equilibrium hydrogen !
concentrations in excess of the flammability limits. Whether j this is correct remains a key unanswered question which should be j
, resolved. j l
i i
Tests were conducted on the Humboldt Bay reactor to determine the L evolution rate from a practical system. The equivalent G H2 was {
determined to lie between 0.0085 molecules /100 ev an(d )0.05 l
{
molecules /100 ev. Thermal calculations indicate that boiling l j I
AS-76 i
1 1
SNPS-1
~
)
lies-in the
.withinlthe corefordid not occur, so thisItrange of tG .p!iht the upper i 1
' expected range nonboiling systems. appears limit' for a boiling system is 0.45 molecules /100 ev absorbed.m Capsule data.
are the only other - experimental ' . information i available and they strongly cuggest 4
that, for BWR water, an
. equilibrium does exist and will te reached in a. short ' time.
Unfortunately, these data cannot . establish with the. accuracy-required,. whether the equilibrium value reached is above or below the flammability limits. - This is because the- capsules 'are 'too l much of an oversimplification of the actual system. Also, . the water chemistry employed, the oxygen . ratios, and the . fluid thermodynamic conditions tested do- not correspond to EBWR conditions. [Therefore, an experimental' program will be needed: ;
specifically. tailored to the BWR parameters .which have an important bearing on radiolysis.] Rate data extrapolated from
)
j these same capsule data indicate that for pure . water a G (H21- f- )
0.065 to 0.15 molecules /100 ev wonid be expected.
Please refer also to similar Respont.es to similar Comments on the 1
'following current facilities:
a) Bell Station, Unit 1, ( AEC Docket No. 50-319) ,
Amendment' 4,.
C/R 1, 2, 3, & 4.
l b) Batch Nuclear Plant, Unit 1, (AEC Docket Amendment.3.
No. 50-321), ;
I c) Brunswick Steam Electric Plant, Units 1'.& 2.(AEC Docket Nos.-
50-324 + 50-325) Supplement A, C/R 6.2.1-1 l
-1
- 2. An experimental program has been formulated by the General Electric Company to: I
)
a) Determine whether for BWR conditions the equilibrium value of hydrogen concentration is high enough to cause concern, and b) Determine the hydrogen evolution rate for conditions of interest to the BWR loss-of-coolant accident.
It is Loop proposed that this program could have several approaches.
tests could simulate the conditions expected after a postulated design basis LOCA. At the Vallecitos Nuclear Laboratory, a suitable radiation source, such as spent GETR . fuel or Co-60, could be used to simulate the core radiation. The key parameters affecting the pressure suppression system which - have an effect on radiolysis would be scaled.
Using this locp, the following parameters affecting hydrogen equilibrium and yield rate would be studied; i.e. , degree of boiling; core residence
. time; system temperatures, all for water chemistry of interest. to AS-77 l'
- ___ __ --_-_-___-- __ _ _ _ . 1
i ..
SNPS-1 BWR's.- The system would consist of a tank .to simulate ' the reactor vessel and a tank to represent the primary containment.-
4
-A second path ' would. consist of a test- to- substantiate ' the hydrogen yield rates on a BWR after shutdown. This test would be similar .to that conducted at~ the PGEE-Humboldt . Bay, Unit 3 reactor, but on a ' larger reactor. *
' Thirdly, data applicable to' BWR's may become available from 'other.
investigators,. including. the National Laboratories. Should such information. bccome available at- an early t
adequate 'as a basis for-design, then this woudate ld.be'eno prove to be an acceptable source.
The General Electric Company is; now initiating work in' allLthree of the above areas. Each path will be kept under constant review for satisfactory applicability. and progress status. . Continuation and completion of any of the paths will be dependent upon the progress of the entire experimental program as a whole.
An analytical ~ effort has been underway since late 1968. . The.
experimental program has been started.- Results abould .become available by late 1969 or early 1970.
Please refer also to similar responses to similar. comments on the following current facilities.
a) Bell station, Unit 1, (AEC Docket No. 50-319), Amendment 4, C/R 1, 2,'3 and 4.
b)- Batch Nuclear Plant, Unit 1, (AEC- Docket No. 50-321),
Amendment 3.
c) Brunswick Steam Electric Plant, Units 1 & 2 (AEC Docket Nos.
50-3 & 4 & 50-325) Supplement 4.
- 3. There are two cases of interest, i.e., large breaks under degraded ECCS conditions in which 0.1 percent meta 1-water reaction occurs and small breaks or large breaks in which all ECCS function and thus no metal-water reaction takes place.
, For breaks in which a small temperature excursion occurs, the metal-water contribution is insignificant. Thus, the initial concentration of hydrogen in the. containment is less than 1/2 percent by volume, due to initial hydrogen in the water and 4
from DO HO I assumed.
using a G (H2 ) of 0.45, a 4one 2 reactions If calculates the concentrabon percent by volume I
concentration will occur in a day. ]
If the G (H 3 ) is lower, say j 0.065, the 4 percent concentration will be reacheB in about 15 days. Although the actual G (H2 ) may be somewhat higher than ]
1 l
AS-78
_ _ . _ _ _ _ _ , _ _ _ . , . _ _ . - '- _ _ _ - - - - - - - - - ~
SNPS-1 0.065, it should be considerably under 0.45. These results shown in this Amendment, Fig. 5.17.C-1, by the solid curves. are I
TAe second case dre assumed to function is for a large break in which the minimum ECCS and 0.1 percent metal-water reaction occurs. Assuming a factor of 3 for margin and the other contributions as in the first should be case, the initial concentration G
no higher than 2 percent. Use of the higher value of of(H2)drogen Dy in about 10 hr.results in achieving a 4 percent by volume concentration For the lower value of G concentration is reached in a week. In both cases, it is )IF2 , this assumed i.e., no back-reactionthat was an equilibrium concentration lower than this does not occur, assumed. Physics calculations by General Electric was absorbed by thehave water. shown that 11 percent of all decay energy Since, performance, only a small consistent with the ,ECCS products leave the fuel, fraction of the radioactive fission radioactive sources are in it can the fuel,beitself. assumed that all the infinite irradiation was The decay power for primary containment were assumed. used and homogeneous mixing in the The effects of primary containment inerting and recombination are discussed in this Response 5.17 part 4.
Please refercurrent followin<; also to facilities:
similar Responsen to similar Comments on the a)
' *C/R
(.11 1, Station, Unit 1 (AEC Docket No. 50-319), Amendment 4 2., 3, and 4.
b)
Batch Nuclear Plant, Unit 1 (AEC Docket No. 50-321),
Amendment 3.
c)
Brunswick Steam Electric Plant, Units 1 and 2 (AEC Docket Noe. 50-324 and 50-325), supplement 4, C/R 6.2.3.
4.
have Atbeen this time, no system or equipment design criteria established relative to the possible or bases radiolysis extent of no equipmentasdesign determined has yetfrombeen the stated RED effort. Similarly, determined. The approach to determining the extent of radiolysis and a solution to the problem if indeed one exists, has been a three-fold one.
established a) It correct solution, through to secure a the applicable data, b) to arrive at a specifically defined problem, rational technical approach to a solution, if needed.
and c) to apply correct design
\
AS-79
SNPS-1 Mis technical approach is further elaborated below:-
a)
General AonroacA Parallel analytical and experimental offorts have already. been initiated by General Electric examinations are being carried out Company. The analytical-allowance to determine how much for radiolysis can be tolerated for arbitrary hydrogen concentration.
to the key parameters. limits, time to reach these limits, and sensitivity At the same time, an experimental effort is being.
conditions. carried out to determine whether a problem exists' at BWR then concentrate If a problem is found to exist, the R&D program will !
! the design of any solutions to the problem.on The finding theaspect third hydrogen of yield rat the < program is to study the feasibility of various solutions and .
select the optimum one for incorporation in the final design.
b) Analysis The objective of the analysis is to establish how the containment-responds to the addition of hydrogen from 'radiolysis assuming various generation rates. This hydrogen as well as the sensitivitywill
.example, to give the time buildup of key assumptions..
it has been found that the flausability limit can For be reached within either a few hours or a few days depending on the assumptions analysis will with alsorespect to hydrogen rate generation. . The replacement rates determine primary containment air volume required to hydrogen concentration. limit the atmosphere to a given Analyses performed up to now have been bounding- the effects on the primary con.for the purpose of tainment system. The primary containment hydrogen buildup has been calculated assuming upper and lower estimates of generation rates. It should be noted that a conservative
. hydrogen present includes over three timesestimate theof the initial quantity of water reaction, the expected metal-inventory. Thus, for the D02~80 2 reaction, and initial hydrogen metal-water reaction occurs, likely more smaller breaks in which the initial hydrogen present no essentially sero. is concentration reaches Thesethecalculations flammabilityare to indicate when limit. the It does The not follow that the hydrogen will necessarily be ignited.
ignition sources within the pressure suppression containment are minimal within it after since no valves or motors are required to operate
. a postulated design basis IDCA. Therefore, the potential for ignition of a mixture is very low.
However, as discussed in detail in Topical Report GE APED-565U O even if a mixture is ignited the pressure suppression containment is not jeopardized over a wide range of hydrogen concentrations and flame propagation velocities.
detonation were to Only in the event that true containment integrity. occur would there be any question of ,
There is a lack of experimental data as l i A5-80
i SNPS-1 ,
31 to what constitutes a detonatable mixture under the design basis IDCA conditions. It is generally agreed upon that a mixture of T ,~
4 percent to 6 percent hydrogen in air will not detonate, but that an M8 percent mixture will. N Again, because of the high humidityf and the effect of the sprays, these limits could be
, significantly higher in an actual containment. However, it is not the intent of the test program to delve into the questions related to phenomena of hydrogen-air mixture ignition or burning.
. c, ; Poten%al Solutions c
should the; radiolytic hydrogen be a problem for the BWR there are seveial potential solutions where feasibility will be examined.
Some of these ar~e discussed below.
- 1. Wnting venting of the primary containment air through the reactor nuilding standby ventilation treatment system may be feasible.
calculations would be performed showing the effect of key assumptions and meteorological conditions on the resulting offsite doses.
- 2. Recombiners l
Recombination of the hydrogen and oxygen is possible either by combustion or catalyst. The systemu have varying degrees of advantages, and these will need to be evaluated. A key consideration is introduction of a safety problem by the very
- ystem which should remove this concern.
- 3. Inertint This is not a solution because the radiolytic hydrogen is formed sto f.calometrically with oxygen. Thus, enough oxygen is always generated along with the hydrogen for flammability considerations.
With an inerted primary containment, it will be practical to keep the oxygen concentration only as low as 4 percent to 5 percent because of the ever increasing volumetric nitrogen changes needed i to go below this value. Examination of O -B -N, cixture tertiary '
.3 7 chart in this Amendment, Fig. 5.17.d-1, indicates that for any oxygen content above about 5 percent, a flammable mixture of hydrogen will occur for hydrogen concentrations above 5 percent.
Thus, even la an inerted containment, the lower flammability limits 4 percent to of an B -0 y -N2 2 mixture would be reached after about a o percent concentration of B g just as if air were present.
impractical It should also be noted that even if zero oxygen (an condition) were present to start with, radiolysis produce 7 one mole of 0 H 2 bas evolved 2 for two moles of 8 2 Thus, after enough to give a 10 percent concentration, there will automatically be associated with it about 5 percent l concentration or enough to make it flammable. 02 Therefore, AS-81
(
_ _ _ _ _ _ _ _ _ _ __ _ _ _ - _ _ _ _ _ )
l 1 )
f SNPS-1 inerting inherently fails to provide against any additional assurance reaching a flammable mixture. This, coupled with the reasons given in Topical Report GE-APED 5654, does inerting a feasible solution to this potential problem ornotformake the previous postulated problems.
Please refer also to similar Responses to similar comments on the following current facilities:
a)
Bell Station, Unit 1 (AEC Docket No, 50-319), Amendment 4, C/R 1, 2, 3, and 4.
b)
Hatch Nuclear Plant, Unit 1 (AEC Docket No. 50-321),
Amendment 3.
c) Brunswick Steam Electric Plant, Units 1 & 2 (AEC Docket Nos. 50-324 and 325) Supplement 4, C/R 6. 2.3.b.
l V
1
{
i 4
AS-82
SNPS-1 REFERENCES (1) " Considerations Pertaining- to Containment Inerting",
GE-APED-5654 ' Class 1, August, 1968.
(2) Jenks, G. H., " Effects of Reactor Operation on BFIR Coolant", October, 1965, ORNL-3848.
-(3) Zittel, W. B., Ma rch-April-1968, " Progress Review",
ORNL-TM-2230.
(4) " Limits of Flammability of Gases and vapors", US Bureau of Mines BULLETIN 503. l l
f ,
l t
l l
b l
)
A5-83
SNPS-1 6.0 gjGINEERED SAFETY FEATURES COMMENT 6.1: In order to enable us to evaluate the adequacy of the design of the control rod housing s upport, please: ,
- a. List the load combinations used for the design of the support structure and the applicable stress and deflection limits for each combination. Provide analyses which demonstrate the capability of the support structure to withstand the seismic loads and the loads due to impact of the housing on the support structure. Describe the analytical methods used to determine the response of the structure to the applied dynamic loads, and discuss the results of the analyses in terms of stresses and deflections for the critical components.
1
- b. Discuss the capability of the control I rod housing to resist buckling under l the design loads, and the potential ,
consequences in the event that it did !
buckle. l l
- c. Provide a failure mode analysis to show the performance capability of the support structure, assuming single failures such as, but not limited to, loss of nuts on a hanger rod, improper placement of a disc spring, and failure of a support bar, grid clamp, or grid to function properly. Describe any special steps taken in design and fabrication, to minimize the probability of such failures.
- d. Discuss the initial acceptance testing that will be performed to assure proper design of the support structure prior to power operation of the plant and the subsequent in-service surveillance requirements for the structure and its components, including tests and inspections, frequencies, acceptance criteria, and action required if these criteria are not met by the test or inspection.
AS-84
SNPS-1 RESPONSE: a. All structural components of the Control Rod Drive (CRD) housing supports are designed for a load combination of dead load, design basis accident jet force (which is equal to the reactor vessel design pressure times the area of the support
= housing) and impact force. For this once in a lifetime load condition, the design stresses are 150 percent of the AISC normal allowable stresses. The deflection is limited to a maximum of 3 in, which is about one half of one CRD notch movement.
]-
.q The vertical seismic load is determined by applying a static coefficient of 0.10 to the total dead load of the CRD housing and m
4 support components. This vertical load, due to seismic, is less than one percent of the total design load identified in 6.1a.
' The impact factor of 3 was determined by using a standard impact factor formula:
A 2n g 5=+f 1+E Impact was factored in the design of the structure by multiplying the applied vertical load consisting of accident, jet, and dead loads by the impact f actor of 3. The resulting force is then treatc. as a static load in designing formulas. i
- b. Please refer to SNPS PSAR Amendment 4, Appendix C, Section C-4, for the required information.
- c. The number and placement of disc springs is determined by the load and deflection requirenunts. The load requires one stack of ,
two disc springs while the deflection requires 24 stacks of two l disc springs. Therefore, the improper placement of one stack of I disc springs will affect only the impact factor applied support structure.
to the A reduced deflection of the disc spring stack will not increase the impact factor beyond the design impact because during the hot operating condition the factor of 3 maximum gap between the housing and the housing support structure '
is only 1/ 4 in rather than 1 in, as was assumed in arriving at the design impact factor of 3.
Loss of nuts on a hanger rod would require personnel errors of the same general type that can defeat any engineered safety system. Because of the procedural controls original which dictate the
' installation of the support structure and reassembly of the support structure following control rod drive maintenance, the nuts on the hangar rod will be retained in their correct position.
One of the procedural requirements is that the design gap between the housing flange and support structure be obtained. The only way to obtain the required gap is by adjustment of the two nuts
~
on the hangar rod. This is additional assurance that the nuts will be in place when the unit is put into operation.
A5-85
L SNPS-1 Failure of the support bar, grid clamp, or grid to function as a structure depends upon its ability to carry ' the design load .
i Since all l
the yield of the stresses in the structural camps at are below stresses of the material )
(150 percent of AISC allowable) , the component cannot fail structurally. i
- d. There will be no Housing Supports prior initial testing of the Control Rod Drive "
because:
to power operation of this. station 1.
The CRD structural housing supports engineering are designed according to acceptable practices.
- 2. I The CRD housing supports are conservatively designed by
! a. using a .,
l between the CRD housing flange and the supportdesign impact fac structure rather than a reduced impact factor based on 1/4 in. gap;
- b. using rather than the operating pressure of 1,000 psig fora computation of jet force.
reactor vessel d 3.
Quality control inspection is pro tided to assure that the CRD l
housing suoports are fabricated according to the design.
1 1
1 I
AS-86 i i
.SNPS-1 l
[ COMMENT 6.22 Please provide analyses to show tnat there would L be adequate suction, heads (NPSH) for all engineered safety ' feature pumps under all potential accident operating conditions.
i-RESPONSE: The ECCS pumps have not yet been purchased, and final design piping losses have yet to be determined. Although it. is premature to determine how much NPSE margin will be available for all potential accident operating conditions, an analysis has been
[ made, the results of which show, that for the design basis LOCA l conditions, adequate suction heads would be available for all ECCS pumps.
This analysis was based. on preliminary design layouts for the f a ECCS piping network, as well as preliminary vendor pump specifications. The analysis, as reported bezein, includes an outline of the ECCS pump performance requirements, as well as the j
assumptions and methods used for assessing.the available NPSH. )
l I. Required NPSH for ECCS P-All three ECCS pump systems have separate suction lines to each
- pump with the suppression pool as a continuous closed cycle source of core cooling water. The analysis reported herein is limited to the RER and core spray pump systems, since the HPCI 4
pump requirements (NPSB) are less stringent. The estimated pump {
characteristics for each system are as follows: 1
~
~
PumD -Reduired NPSH Rated Condition HPCI 21 ft 2,800 TDH a 4,250 gym a 130 F Core spray 32 ft 741 TDH a 4,725 gpa a 130 F RHR* 33 ft 443 TDH a 7,700 gpm a 130 F
- Based on LPCI rode with three RER pumps j These rated conditions corrW to conditions appropriate to the later stages of reactor vessel blowdown.
The detailed ;
criteria governirg these PSAR Amendment 4,Section VI.
specifications are presented in SNPS {
i Throughout the postulated design basis LOCA, the temperature of the suppression pool continues to increase steadily until a peak value is reached the Shoreham some Nuclear 10 hr Power after the start of the accident.
Station, For i listed above, the 130 F temperature, as initiation, since would nominalbe obtained pool at initial about 300 see will temperature afterbeaccident 80 F. below For any further increase of the pool temperature above the 130 F value, the NPSB required for each pump (at rated flow conditions) will be less, but exactly how much less is not known at this time. ,
A5-87
SNPS-1 ,
1 s !
The: flow and developed head for each pump systes will also vary as the postulated lOCA progresses beyond the reactor and. vessel i . blowdown phase. Fig. 6.2-1 illustrates the estimated performance- '
of the RHR and core spray _ systems during this ' period.
For. the RHR system, . as noted in Fig. 6.2-1, . the rated condition corresponds essentially to the ' end of blowdown (20 psi gage
- reactor. pressure). As a result, further reduction in reactor vessel pressure would cause, at most, a 1,000. gpm (total .for three pumps) increase in flow rate. Such a ~ slight increase is not expected to alter significantly the 33 ft required NPSB value-for each RHRS pump. The LPCI mode would be terminated manually by the operator, and the RHRS containment . spray / cooling subsystem-will be activated ' when the operator.has been assured that-the RHRS pumps are no longer required for reactor vessel flooding.
Fig.' 6.2-2 presents a preliminary estimate of the RHRS pump performance conditions when the spray / cooling subsystem is operational. At' this time (approximately 600 sec after start of the accident), as discussed in SNPS PSAR Amendment- 4, S ections VI-2. 2. 5. 2 and . V-2.4.2, one RHRS pump- (and associated. heat exchanger) would be adequate to ensur.e primary containment integrity. . As illustrated in Fig. 6.2-2, the piping resistance for the RHRS spray / cooling loop will be sufficient to . preclude excessive flow even with only one RHRS pump left on. As a result, the maximum required NPSR for this operational mode is expected to remain at approximately 33 ft.
Fig. 6.2-1 illustrates the expected performance of the core spray system during.the blowdown phase of the LOCA. The required NPSH of 32 ft pertains to a reactor. vessel pressure of about 121 psi gage. For higher vessel pressures, the flow would be less and the NPSB requirement would also be lower. As the blowdown is.
completed and vessel pressure approaches containment conditions, the core spray system pump flow would ae limited by. line losses to a maximum of about 6,000 gps. For this runout flow condition, the 36 ft.
required NPSH is estimated to be increased to approximately II. Available NPSB for ECCS Pumps A. Calculational Assumptions The NPSH for each of the ECCS pumps at any point in time i
- following a LOCA can be obtained from the following relationship: f Available NPSH = Suppression chamber air pa rtial pressure + static head of water (above elevation f of ptsp center line) - suction line losses d The calculational assumptions chosen to reach a conservative assessment of available NPSH for the ECCS pumps are related to the above expression in the following manner:
A5-88
p SNPS-1
- 1. Suppression chamber air partial pressure l For this f actor, the. assumptions chosen were aimed at minimizing the. air partial pressure within the suppression chamber for any time during the postulated LOCA. The assumptions are: i
- a. Minimization of noncondensables Initial drywell (average over entire volume) conditions.of 0 psi gage, 150.F, 30 percent relative hr.midity and initial suppression chamber (air) conditions of 0 psi gage., 130 F, and 100 percent relative humidity.
These assumptions can be contrasted with the expected nominal conditions in the.drywell of about 135 F, 1 to 1.1/2 psi gage, 35 percent relative humidity; and for the suppression chamber 70 to 80 F, 1 to 1 1/2 psi gage, and 100 percent relative humidity.
The . values of 130 F, together with 0 psi gage do not correspond to any realistic initial conditions for the suppression chamber and were chosen in the light of simple conservatism to demonstrate NPSH adequacy even for this extreme case.
- b. Actuation of containment spray throughout LOCA (after first 600 sec)
The ' assumption of containment sprays being used throughout the LOCA serves primarily to minimize the amount of noncondensables in the suppression chamber at any time following LOCA; since, when the sprays. are actuated, a lower drywell temperature results.
- c. A constant primary containment leakage rai.e, throughout the LOCA, of 0.5 percent of enclosed volume per day.
The leakage rate is the same maximum value used for LOCA dose calculations and corresponds to the maximum leakage rate to be allowed for the Shoreham primary , containment design. By considering the leakage rate to be constant and at its maximum value throughout the accident, the amount of noncondensables within the containment is minimized.
l
- d. Maximum service water temperature of 80 F This assumption serves to reduce the effective heat removal rate of the RHRS heat exchanger so that, as a result, higher pool j
. temperatures are obtained throughout the LOCA. This condition l results in minimizing the air partial pressure within the J suppression chamber.
- 2. static head of water (above elevation of pump center line) i Both the RERS and core spray system pumps were estimated, from preliminary design layouts, to have their pump suction line inlets at a maximum of 2.0 ft above the elevation of the suppression chamber floor. As a result, with a minimula initial water level of 18.0 f t, the static head of water available for each pump is at least 16.0 ft.
A5-89
SNPS-1
- 3. : Suction line losses The. suction line . losses ,were maximized by considering both the RHRS and core spray system pumps to be operating at their runout flow conditions throughout the LOCA. A flow rate of 6,200 gpm-
)
was used for the core spray system pump, and' 10,400 gpm was used for a . single RERS pump. A suction 'line size of .16 in, was used for the core spray system pump and 20 in. for the RHRS ' pump.
suction A line head loss of.6 ft (at runout flow) was obtained and used syista.
in NPSH calculations 'for both the 'RHRS and core spray B. Calculational Techniques
.The ~ details of how the available NPSH was calculated are as follows:
i First, the initial mass of air.'inside the primary containment was calculated, using the assumptions listed in Part A. The total-mass is then. given by (refer to Table 6.2-1 for equation nomenclature):'
Mt. 14.7 - Ps,a)Va + .(14.7 - Pvs)V as RTd
.RTs During the initial blowdown of the reactor vessel, essentially '!
all, of the air is transferred from the drywell to the suppression J chamber. After the blowdown is completed and as the drywell is cooled by actuation of sprays, air is bled from the suppression j chamber to the drywell through the vacuum breakers. The L- drywell/ suppression chamber vacuum breakers will serve to limit I the differential pressure between both chambers. With vacuus breakers rated at 0.5 psi (full open), the following relation is obtained Pd + 0. 5 = P, 1 Although the available NPSH calculations were based on the above !
relation, the NPSH would only be reduced by about 1 ft if the 1 l total pressures in both chambers were considered to be equal.
Now, it is needed only to specify the drywell and suppression 3
chamber temperature to calculate the total pressure inside the ;
primary containment. The suppression charber air temperature p will be approximately equal to the pool temperature. To determine the drywell temperature, however, we must consider the processes which are occurring. Water from the pool is being pumped into the reactor primary vessel by a reactor core spray cooling system pump. This water is being heated by the decay heat of the reactor core and is pouring out of the reactor AS-90
_ _ _ _ _ _ - _ _ - - - - , - - - - - - - - - - - ' ' ' - - ~ - ' - ' - - __ ___-_ - _ - - - - - - _ - - ---
SNPS-1 primary vessel via that postulated break into the drywell and then back into the suppression ~ chamber.
to the suppression chamber is necessarily The water which returns i
hotter than the pool mater. Because of the large flow of water involved, the low thermal capacity rate of change of the drywell atmosphere, and the slow time the drywell, the atmospheric temperatureofwill the temperature track the water of the water flo temperature quite closely.
drive the system.) Therefore,(That is, the water temperature will the temperature for the drywell atmosphere will be higher than the suppression pool temperature.
Due to the low containment leakage rate, the amount of leakage which takes Oban 1 percent)place during the first two days is negligible and will be ignored until that time.
(less Section V-2.4.2 of Amendment 4, SNPS PSAR, presents results of a parametric study of ECCS pump availability on the long-term containment case 4A pressure and temperature response.y The results for (one core spray system and one RBRS pump), as described in Section V-2.4.2.4, were used for this evaluation, since the required NPSH for the RBRS pump would be slightly higher than if two RHRS pumps were assumed operating. The temperature history for both the drywell and suppression chamber that was used for the NPSH analysis is presented in Figure 6.2-3. The portion of the curve af ter 1.4 x 105 see was estimated by assuming that the suppression pool would return to 130 F within 30 days. Applying equation 2 above and the ideal gas law yields:
MdwRTd + 0.5 + PVd = s s+P Vs Vd Vas also Mdv + Hsc " Mt These equations can be solved for M . The ideal gas law is then used to calculate l chamber. the air partial pressure in the suppression Since it is a closed system, the NPSH is merely the sum of air partial pressure plus the liquid head above the pump suction less the suction losses: i NPSH = (Ps -Py ,) + H - Suction Losses MscRT s 144 NPSH = + H - Suction Losses V
as _ D Fig. 6.2-3 indicates that af ter about 50,000 sec, the drywell and suppression The chamber temperatures are within 2 deg of one another.
calculations described suppression chamber to the dr here show that air bleeds from the this flow essentially ceases.ywell until this time, after which only phenomenon which will Therefore, from this point on, the leakage of air from the containment and affect the available NPSH is the
) temperature reduction.
A5-91
_ - -___ __ _ _ _ )
SNPS-1 Assuming that the suppression' chamber loses air at the same rate i
-as the drywell, namely, 0.5 percent per day, the following relation obtains:
M*se = W.9 SP x h sc The available' NPSH may. then be calculated as.'above as a ' function-l of time after initiation of the LOCA. The resulting values of available NPSH are shown in Fig. 6.2-4.
Note that after' 'considering 30 days of continual containment atmosphere outleakage (at a constant value of 0.5 percent per day) the available NPSH is shown to f all to a_ minimum of 38 f t.
The total suppression chamber pressure at the end of 30 days is calculated as 13.8 pria. This condition, however, would not be obtained in an actual case since the containment outleakage would I cease once atmospheric pressure was reached. When the latter restriction is considered, then the available NPSH would fall only ' to a~ minimum of 40 f t since atmospheric pressurc. would be-reached in 17 days.
C. Conclusions It should be apparent that if adequate NPSH can be maintained for all of the above events occurring simultaneously, the ability of the ECCS to maintain NPSH for all expected modes of - station operation. can be assured.
'Ihe NPSH be, available for either the core spray system or RHRS pumps will for the design basis loss-of-coolant accident, well above the expected required values for these units. The margin above minimum requirements would be expected to increase, rather than diminish, for potential accidents which are less severe than the design basis LOCA.
A5-92
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I SNPS-1 t
i TABLE 6.2-1 EQUATION NOMENCLATURE D Time after accident, days i H Elevation of water above pump inlet, ft Me Initial niass of air -inside the drywell and suppression chamber, lb moles Mdw Mass of air in the drywell, lb moles M sc Mass of a'ir in the suppression chamber, Ib moles Mje Mass of air in the suppression chamber af ter D days of leakage, Ib moles NPSH Net positive suction head at pump inlet, ft of water Pd Total pressure of drywell, psia Ps Total pressure of suppression chamber atmosphere, psia Pvd Partial pressure of vapor in drywell, psia Pvs Partial pressure of vapor in s~ ppression u chamber, psia R Gas constant Td Temperature of drywell, R l
Ts Temperature of suppression chamber (air and water), R l.
Vas Free volume of air in suppression chamber, f t8 l 1
l Vd Free volume of drywell, f t3
) i p Density of suppression pool, lb/ft3 i
i i
A5-93 l __________- 9
, SNPS-1 o
' COMMENT 6.3: We note- that the RHRS . pumps deliver flow at 4 pressures as high as 300 psi gage, (pg.'VI-2-27).
How does this-relate to the pressure differential in the . RHRS heat exchanger, wherein the service water pressure is to be always greater than the RBRS pressure in order to preclude the release of any radioactivity by this path?
RESPONSE: LThe RHRS service water similar in design to that licensed for system- TVA-3.
for SNPS-1 will be (See SNPS PSAR Amendment 4,.Section X-4.2.) '
The maximum calculated pressure at the inlet, of the primary side of the RHRS heat; exchanger occurs at the beginning of <
reactor noomal j shutdown operation on the RHRS pump suction with 50 psia reactor pressure' imposed (Mode-type "E" as defined in Table 6.4 of this Amendment in Response to Comment 6.4). .The total developed pump head at this flow67is estimated to be about'159 peig. - Accounting for .an estimated ft elevation difference, the total {
, to be 223 psi gage. pressure at the RHRS heat exchanger inlet is expected 4 The maximum' shutoff head of the RERS pumps as designed is i consistent with the 450 psi design pressure of the RHR heat k exchangers. {
Redundancy of operation is provided with independent two complete and RHRS heat exchanger loops where only one is required {
for safe station shutdown (this amendment, response to careent j 13.1). Radiation monitors are provided in the service water discharge lines of each RHRS heat exchanger.
Should a leak develop inadvertently the radiation monitor.in a heat exchanger, it will be detected by The faulty exchanger will be secured to prevent release of radioactivity and the heat load will be switched to the other exchanger.y l
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AS-94 l
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-SNPS-1 {
l COMMENT 6.4: What. are the . design parameters and the heat-
-removal- capability of the RHRS heat exchangers 4 for each of the several modes of . operation what and is the . rate of heat removal that is essential'in each ' mode of operation?
RESPONSE
within this Please refer to .the l Response to Comment 13.'1 Amendment, for references exchanger design basis, . description performance on . the ECCS-RHRS heat >
inspection and test requirements. analysis, and
) 1 Although long-term the RHRS heat exchangers 'are designed on the basis 'of a capability reactor shutdown cooling mode, they maintain the of removing the necessary residual (decay. and sensible) heat from the reactor core and primary containment suppression pool during all credible modes of RERS operation, j These operational modes, corresponding' design conditions and l respective heat loads are summarized in Table 6.4-1. I The RER heat exchanger performance that is - " essential", from safeguards standpoint, can a PSAR Amendment ~ 4, Figs. V-2-14, 15, 16. and 17. be ascertained from review of SNPS present the LOCA long These figures response with various en. term containment temperature and pressure q PSAR Amendment 4 Section gineered safeguards, as described in SNPS ' '
V-2.4.2. In all cases the heat exchanger performance is considered more ' than adequate.
For example, four RHR Fig. pumpsV-2-15 and presents the containment response if all both RERS . heat exchangers are considered operating. Once containment sprays are activated the heat' exchangers below limit the peak pressure to about 13 psi gage or 35 psi the containment design limits. Similarly, Fig. V-2-16 in SNPS PSAR Amendment 4 demonstrates that, with only one RHRS heat exchanger and one RHR pump, the peak pressure after containment 4
sprays are activated, is limited to about .20 psi gage: l well below the 48 psi gage limit. a value Thus, when the performance of two RER added, in heat exchangers is contrasted with only one, the margin 7 psi. terms of primary containment pressure limits, is only It can be concluded, therefore, that from an accident response criteria of how man standpoint, the heat exchanger is oversized, regardless RHR pumps Furthermore,y the design are considered to be operating.
condition for which the long-term (first 10 hr) basis LOCA integrated corresponds energyto the liI input to the primary containment is a maximum. After about 10 hr the decay heat rate would be less than the heat exchanger heat removal rate even with only 1 RRRS heat exchanger and one RBRS pump operating. It follows then, that the " essential" heat exchanger size (or heat removal rate) would be, from an accident response standpoint, less when the twc other modes of RHR systen
[.
l.
operation are considered i.e. bot standby and normal shutdown.
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I SNPS-1 i
1 TABLE 6.4-1 l Calculated Heat Removal Total Calculate Capability Per ' Heat Removal Mode Desion Conditions Exchanger Stu/Hr capability Btu /
l Containment .
l Cooling. . One _ heat exchanger post accident 49.5x10* 49.5x108 containment spray with heat re-jection, two pump operation C1 C2 same as above - with one-pump 71.7x10* 71.7x108 1 operation - peak pool temperature !
Reactor Isolation ;
D1 Steam condensing - 2 heat 74. 5x108 149x10*
exchangers 1-1/2f2BrLater D 2(A) 1 Heat exch. remaining in 107.2x108 condensing operation f 129.6x108 D 2(B) 1 Heat exchanger 22.4x108 switched to cool suppression k pool..
Normal Shutdown 'l E Af ter bloWown to main con- 118.2x10* 236.4x108 denser - 2 heat exchanger - 50 psia Reactor pressure F Continuation of normal shut 30.8x10* 61.6x108 (Design Mode) down from mode E - zero Psi gage Reactor pressure - 2 heat exchangers Reactor '
Temp. 125 F, 20 hr af ter scras AS-96 l
l
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SNPS-1
' COMMENT 6.5: Identify the equipment and components (e.g.,
, valves motors, instruments, cable, etc.) located in the primary containment which are required to be operable during and subsequent to a - loss-of-coolant accident and describe the tests that will j be or have been performed to demonstrate that .
ted these components would be operable in the combined high temperature, pressure, and humidity <
u/He
- Y E 9 accident.
RESPONSE: The following is a list of the . instrumentation and electrical equipment located within the primary containment which are required to mitigate the effects of a loss of coolant accident.
Main steam isolation valve air control solenoid valve Main steam safety / relief valve actuator solenoid valve Recirculating pump discharge valve motor (valve operator)
Recirculating pump suction valve motor (valve operator)
Recirculating pump discharge bypass valve motor The qualification testing of prototype pieces of this equipment is being undertaken in an autoclave pressurized with saturated steam at 62 psi gage. Typically, one complete operation cycle is performed at 5 min intervals for a period of 10 hr or until failure. Upon completion of the qualification testing a summary of the program will be submitted to the AEC. Electrical cable systems for this equipment will be designed so tnat the cable will be operable for the required period of usage in the design basis environment predicted c: iro location.
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SNPS-1 COMMENT 6.6: The three station ser-ice water pumps are each indicated to be %alf capacity" pumps, but only
, one pump can be driven by each diesel generator d (pg. X-4-2). This arrangement does not appear to a satisfy the criterion that adequate cooling water
. could be provided in the event of an accident 1 even if only one diesel generator were operable.
t Please discuss.
e y RESPONSE: Design of the Station Service Water System described n in Section X-4-1 and shown schematically in the flow diagram (Fig. X-4-1) in Amendment 4 SNPS PSAR was based on the single failure criteria, in which case no more than one of the three d diesel generators would fail to start. Since there is a Station h Service Water Pump connected to each diesel generator, a ' minimum e of two pumps will be available to carry the full plant load if required. The station service water requirement during a loss of a-c power emergency will be less than the rated output of a single station service water pump. Therefore, the second pump will act as a 100 percent spare during a loss of a-c power emergency.
t d
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~
SNPS 1 COMMENT 6.7%
h Please provide a single sc'bematic diagram which
.y -
showssystem, all the emergency core cooling systems, the r
RCIC their interconnections, and their o relationship to those major plant components to which they are connected, r
t vessel and core, the primary such as the reactor containment,
, condensate storage tank, and the
, lines. the main steam
RESPONSE
3 Please refer to Fig. VI-2-2ti, Amend. 4, SNPS m Section VI-2, for the requested- diagram. PSAR e
e n
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a ,
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AS-99
_ _ - _ _ - - _ _ . )
p
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0 SNPS-1 l
L COMMENT'6.8 Provide core an analysis to show;that minimum required l l cooling capability.would not be lost in the
[ revent of. a . failure of any . component in the -3 i
L emergency core cooling systems, including the rupture of any pipe in these systems. -1
RESPONSE
This facility and all other. current BWR; facilities i
conform ir depth to all aspects of the-functional and operational i design of the intent,of the proposed AEC Design Criteria. . Plear,e refer to SNPS PSAR, Amendment 4 .
- 1. PSAR. Section I-2.0 " Principal Design Criteria" and PSAR Section 'I-3.0 " Summary Design Description and Safety Analysis".
I l
- 2. PSAR Appendix G Design Criteria", " Station Comparative Evaluation with AEC Section 2.7 Features". (Criteria 37 through 65)" Group VII - Engineered Safety. )
In order to establish the conformance of the Shoreham f acility to 4 and requirements the of the AEC Design Criteria with regards to ECCS primary containment design, the following discussion is submitted...
1) criteria-conformance Identification i
The normal station control systems maintain station variables within narrow operating limits. These systems are thoroughly engineered and backed up by a significant system design and operation. Even if an improbable maloperation amount of experience in or equipment failure occurs, including a reactor coolant boundary break in up to and including the circamferential rupture of any pipe that boundary,-assuming an unobetructible discharge from both sides, which allows variables to exceed their operating limits, an extensive system of engineered safety features (ESP) limits the transient and the effects to levels well below those which are of public safety concern. Thesc engineered safety features (ESF) include the normal protection systems (reactor coolant system, station containment systems, station and reactor control systans, reactor protection system, other instrumentation ' and' ,
process systems, etc.), those which offer additional protection against a reactivity escurai.on (reactor standby liquid control
{
1 system, control rod supports), velocity limiters, and control rod housing those which act to reduce the consequences basis accidents (main steam line flow restrictors,of design primary containment atmospheric control system), and those which provide
. emergence core cooling and containment cooling in the event of a ;
loss of normal cooling (core spray cooling system (CSCS) core residual heat removal systes (.RHRS), core high pressure c,oolant injection- system (EPCIS), automatic depressurization system l.
(ADS), and the standby coolant supply systes). (Criterion 37) ;
Sufficient offsite and standby (redundant, independent and testable) auxiliary cources of electrical power are provided to attain prompt shutdown and continued maintenance of the station AS-100
SNPS-1 in a safe capacity of condition the power- under all credible sources circumstances . The
+ are adequate to accourplish all design ' basis ' accident conditions. required engineereG safety fea
. (Criterion 39)
The engineered safety features are -designed to provide high reliability in each ESF 'and toready testability. Specific provisions ' are made capabilities.
demonstrate operability and performance.
(Criterion 38)
' basis Components of the ESF which.are required to function after design accidents are designed to withstand the most severe forces ,
(
and credible environmental . effects' missi3es the events- from plant equipment failures which 'are: anticipated frominclu without impairment of their performance capability.
(Criteria 40, 42, 43) ,
l The performance designed such of tne emergency core cooling systems (ECCS) are phenomena are p that at rovided .toleast prevent twoclad different melt ECCS of different spectrum of postulated . design basis reactor primary _systes over the entire breaks.-
of all offsiteSucha-ccapability power. is available concurrently with the loss are . The ECCS individnal systems themselves designed to various levels of component redundancy such that no single active component failure in addition to the accident will negate the. required (Criteria 41, 44) emergency core cooling capability. _
To assure that the ECCS will function properly, -if specific provisions have needed, been made for testing the sequential
.operability (Criteria' 46,and47, functional
- 48) performance of each individual system. ,
Design provisions have also been made to enable physical visual inspection of the ECCS components. (Criterion 45) and The primary containment penetrations, is structure, designed including access openings and i to withstand the peak transient pressure design basis and temperatures which could occur due to the postulated loss-of-coolant accident. ' The conuinment design includes considerable allcance for energy addition included from metal-water or other chemical reactions beyond conditions that would occur with normal operation of systems (ECCS).
(Criterion 49) emergency core cooling Plates, structural members, forgings and pipe associated with the drywell have an initial NDT temperature of approximately 0 F when tested It in accordance is intended that with the appropriate code for the materials.
the drywell will not be pressurized or subjected to substantial Provisions are made for the stress removal at temperatures below 50 F.
of beat from within *be station containment system and to isolate the various process system lines as may be necessary to maintain the integrity of the station containment systems as long as necessary following the AS-101
_ m._.__._ _ __ _ _ ___ _ _ . _ _ _ _ _ _ _ . _ - _ _ _ _ _ _ _ _
SNPS-1 t '
,various postulated design basis accidents. The integrity. of the complete station containment is designed and maintained so that the- offsite . doses resulting from -postulated des'gn basis accidents will. be (Criteria 50, 51, 54) below -
the reference values stated in 10CFR100.
All -pipes or ducts which penetrate the primary containment and which connect {
0 provided with to attheleast reactor coolant system or to the drywell are two isolation j valves in (Criterion 53) series.
The station design includes. preoperational pressure and leak rate testing of the primary containment system, and a capability for leak testing at design pressure after the station has operation. consenced (Criteria 54, 55)
Provisions are also. made for demonstrating the functional-performance of the station containment system isolation valves and leak testing of selected penetrations. (criteria 56, 57) j l
The pressure suppression concept phenomena and the containment spray cooling system provide two different means to rapidly '{
j condense the ster.m portion of the flow from the postulated design basis loss-of-coolanc accident so that the peak transient pressure shall be lesa than the primary pressure. . containment design (Criterion 52) Demonstration of operability and ability to test the functional performance and inspect- the containment 60, 61) spray / cooling system is provided. (Criteria 58, 59, The reactor building standby ventilation system is designed such that- means are provided- for periodic testing of the system performance 64) including tracer injection and sampling. (Criterion This -system may be physically inspected and its operability demonstrated. (Criteria 62, 63, 65)
References to applicable sections of the SNPS PSAR Amendment 4 areSection G, given 2.7.
for the individual criteria of this group in Appendix
.2) Requirement-Conformance Analysis Examination of each AEC Design Criterion requirement and the Shoreham BWR conformance establishes the following with regards to the comment request. t A. Passive Component Failure (Such as a pipe or valve upstream of the core spray pumps)
No AEC Design Criteria require passive corpoaent failure (s) . A review of the applicable criteria establishes that:
AS- 102
SNPS-1 al' from Criteria 41:
l
- Engineered safety features such as emergency core cooling and containment heat removal' systems. shall- provide sufficient ,
perfc:unce capability to accommodate partial loss of installed capacity and ystill fulfill the required safety function. .As a minimum, each engineered safety . feature shall provide this
- required safety function assumino a failure of a single active component".
Active component . failures. aref designed for by providing a sulti-source, capability ECCS network.
b) from Criteria 44:
"At ferent.least two emergency core cooling systems, preferably of dif-design . principles, each with a capability for accomplishing adequate emergency core cooling, shall be provided.
Each emergency core cooling system and the core shall; be designed to prevent fuel and - clad damage that would ' interfere with the j~
emergency core cooling function and to limit the clad metal-water reaction to negligible amounts for all sizes of breaks in the reactor coolant pressure boundary including the double-ended rup-ture of the largest pipe. The performance of each emergency core cooling system shall be evaluated conservatively in each' area of - -
uncertainty. The systems shall not share active components and
_shall not share other features or components unless it can be demonstrated that (a) the capability of the shared feature or i component to perform its required function can be readily ascer-tained durina reactor operat:.on, (b) failure of the shared fea-ture or component does not initiate a loss-of-coolant accident and (c) capability of the shared feature or component to perform its required funct:.on is not immired by the effects of a loss-of
-coolant accident and is not lost during the entire period this function is recuired following the accident."
The LOCA does not impair the ECCS - Primary Containment Network since conformance to criteria 1, 2, 4, 38, 40, 41, 42, 43, 45, {
46, 47, 49, 51, 52, 53, 54, 55, 57, 58, 59 and 61, and the design i parameter of having Class I structures, systems, and components, (the primary. containment system and the ECCS inside a Class I building (reactor building) assure -complete pre-accident and post-accident LOCA, seismic, tornado, flood, and fire, protection for both core cooling continuity and containment integrity. ,
I
'Iherefore, no requirement is documented for the assumed failure l of a Primary Containment or ECCS passive component. Provisions I are provided for mitigating the consequences of non-Class I structure, systems equipment, etc., from having any adverse effect on the Class I equipment required for operation. I B. NPSH Requirement on ECCS pumping equipment (such as the core spray pumps)
AS-103
.1 SNPS-1 l In compliance with Criteria 38,.39, 43, 44, ~ 46, 47, 48, 49, and 53, particular attention should be given to. . .
a) Criterion'42 which states...
" Engineered safety features shall be designed so tha: the capability of each component and system to perform its required ;
function is not -impaired by the effects of a loss-of-coolant' !
accident". j That 'is, the pumping ability of equipment do_es perform under the postulated design basis LOCA conditions of pressure, temperature and their effects on NPSH. (See this Amendment, Response 6.2. )
b) Criterion 52 which states...
"Where active heat removal systems . are needed under accident- !
conditions to prevent exceedina containment desian pressure, at least two systems, preferably of different principles, each with full capacity, shall be provided".
This has been done so that the containment pressure and tempera-ture are reduced to maintain containment integrity and to guarantee continued post-accident (long term) performance of the ECCS. This circumvents any NPSE temperature-pressure effects.
Water level effects are of no design concern.
l A5-104
SNPS-1 COMMENT 6.9: Summarize the results of your analysis of the expected ECCS performance-(assuming loss of off-site power) by presenting the peak clad temperature, the percent of rods perforated, and the percent of fuel clad oxidized for the spectrum of liquid and steam breaks with various combinations of emergency core cooling subsystems operating as the result of postulated failures in each subsystem.
RESPONSE: Please refer to PSAR Amend. 4, Section VI-2.0 for a detailed, thorough analysis of the performance of the ECCS network for the break spectrum of credible, postulated reactor primary coolant system piping failures as the result of a single failure in a component in the ECCS network.
1 1
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I A5- 105 l
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i- SNPS-1
!- COMMENT 6.10:
We need additional information on the design bases for the standby liquid control system. Please provide the details on how the total reactivity capability of the evaluated, indicating system will be established and for imperfect mixing the in allowances to be made the core and other uncertainties. Also, provide additional details on how the design injection rate of liquid poison will be established and evaluated. .The PSAR (pg.
VI 2) implies that a " normal" reactor cooldown will due betoconsidered, but not a more rapid cooldown potential malfunctions or failures.
Please clarify and discuss this.- What uncertainties to be used toare reactivity capability attributed to the calculations establish the required total and injection rate for the system? Discuss the bases for the claimed accuracies of these calculations.
RESPONSE
as _ The design basis of the standby liquid control system, discussed provide shutdown in Section VI-6.0 capability of event in the Amendment 4 of the PSAR is to the normal drive system becomes control rod completely inoperative, while power operation. As shown in Table VI-6-1, at normal Amendment 4, of the' SNPS PSAR, 800 ppe in the boron concentration provided by the system is approximately the reactor vessel. Insertion of boron takes 1 hr, results. thus an input rate of about 13 ppm per min Fig. - VI-6-3 of the SNPS PSAR, Amendment 4 gives the curve of just critical conditions injection. followed as shutdown The power range decrease occurs due to boron from full power to zero power occurs from boron at approximately 230 ppe as shown in Fig.
VI-6-3, 18which about at the rate of 13 ppe per minute takes a period of 0.06, minutes this represents to achieve. Based on a power coefficient of an average reactivity change approximately 0.003 Ak/ minute. of The 4, method of insertion as discussed in the SNPS PSAR, Amendment is achieved by means of a sparger located near the vessel bottom. Flow up through the reactor core takes place reactor and is assumed to be accompanied by perfect mixing of the boron with reactor water. This assumption is utilized to simplify reactor physics calculations. To account for imperfect leakage, 25 percent extra mixing and solution is included as
- calculations of the required boren concentration are donemargin. based on standard GE nuclear. design methods. Appropriate modifications for poison-induced changes in the worth of control elements are nade. Calculations are done for beginning of life conditions when many control full power. rods are partially or fully inserted even at elements are For this case, the changes in the worth of control greater than at any other time and thus more worth is required in the poison control system.
A5- 106
l SNPS-1 s The accuracy in the above calculations is well within the margins . '
'.o , provided in the standby liquid control system design. These are:
1)
Y. t' of.:the the 0.05 Ak shutdown margin provided (as shown in Fig. VI 3 d-solution SNPS PSAR Amendment 4; and 2). the 25' percent additional above that . required which is included for mixing 8:
r . uncertainties and leahage.
s-
- The standby- liquid control system is not a protection system as D- , defined by Criterion 20 of the ABC proposed 70 . Design Criteria, n and therefore is not. designed to meet this criterion. The.syster is manually redundant actuated and is provided as .an independent and subcritical. . The capable. of making and holding the core system-t s ;
1 for fast reactor shutdown as has been noted in the' responses standby liquid conj the following facilities: to l 3 1, a) Pilgrim Nuclear Power Station, Unit "
Plant, (AEC Docket #50-293) Amendment #2, C/R 1.0.; b) Hatch Nuclear bait '1 (AEC. Docket #50-321) Appendix G, page G-36.; and c)
Brunswick Steam Electric Plant, Units 1 & 2, (AEC Docket Nos. 50-
' 324 and 50-325) Appendix G, Section 2.5. Refer to Appendis G
,. SNPS PSAR Amendment 4. .
of
!. 'f The system is an independent- backup _ system to the normal operational control system which conforms to 'the
- criterion 20. intent of i- to perform, If- the standby liquid control system is requested this means that many single component failurec. have
' lI occurred in the normal operation systems. Thus its system l
redundancy is not required. Camponent redundancy is part of the system than as adesign safetyas an operational availability convenience rather requirement.
}
For control additional systemdetails pleaseinvolving refer to the
. . . design of the standby liquid l- . .
- a. Batch Nuclear Mant, Unit 1 Amenhts 2 and 3, /R VI-14.1 and VI-14.2a (AEC Docket No. 50-321) throngh f.
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SNPS-1 1
COMMENT 6.11: To what extent will the standby liquid control system be testable before reactor operation and
- periodically during plant life to demonstrate operability of the system, including full flow of the i
neutron absorber solution through all piping .f and' equipment which normally contains solution in the standby condition? j RESPONSE: The standby liquid control. system is testable in parts. completely containing the poison While at manual power operation, the sections solution are provided with means to circulate solution the positive at full flow from the solution tank, through displacement )
the tank. pumps, and to return the solution to #
operation, When the reactor is shut down during refueling' system flushing is provided by means of a demineralized full flow into water test tank from which water can be pumped at-the reactor vessel through the poison sparger !
using the system pumps. )
The dominera11:ed water test can be conducted while at normal power test operation procedure. but this is not a normal operational availability l
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A5-108 I
SNPS-1
- ).13 i
- e i COMMENT 6.12: What provisions will be made in the design for the detection of and protection against the potential consequences of salt water leaking into the vapor
, suppression systes - ECCS water due to a leak in the RHRS heat exchanger either during normal (testing) operation or during an accident? Could y- ,
such a leak result in the tailure of any critical
, components due to chloride stress corrosion?
RESPONSE: Reference to SNPS PSAR Amendment 4, Section VI-2.3 and Figs. VI-2-6A and VI-2-6B will show that no RHRS service water (salt water) line is connected directly to any line which leads the vapor s.sppression system. System design and the relative- operating pressures in the RHRS heat exchangers and the RilRS service water system (please refer to this Amendment, Responses No. 6.3 and 13.1) are such that salt water in-leakage o'
does not occur. ,
r 4
A5-109
, SNPS-1 1
I 7.0 INSTRUMENTATION AND CONTROL k I
COMMENT 7.1: Please describe the i would emergency procedure which be necessary tofollowed shut down .in the the event-reactor it-tobecame a cold I i
condition sufficient from outside detail to show the that control' room in i possible this will be i equipment, by appropriate instrumentation manipulation of existing and including opening panels, jumpering controls, of wires, etc.
RESPONSE The typical shutdown procedure for a BWR under the conditions cited has been previously' described in detail in.... ;
- a. Cooper Nuclear Station, Unit 1.,
Amendment 3,Section VIII. (ABC Docket No. 50-298) t
- b. Cooper Nuclear Station, Unit 1, (AEC Docket No. 50-298) l Amendment 5,Section II. '!
- c. Pilgrim Nuclect Power Station, Unit 1 (AEC Docket No. 50-293) Amendment 5, C/R 14.0.
d.
Browns Ferry' Nuclear Power Station, Unit 3 (AEC Docket No. 50-296) Amendment 2.
- e. Bell Station, Unit 1 (AEC Docket No. 50-319)
Amendment 1, C/R 8.1.
f.
Batch Nuclear Plant, Unit 1 (AEC Docket No, 50-321)
Amendment 2, C/R VII-4.0.
g.
Brunswick Steam Electric Plant, Units 1 and 2 (AEC Docket Nos. 50-324 and 50-325) Supplement 2, C/R 7.12.
This same information is described below:
The station normal design operating permits the reactor to be brought from any condition (e.g., steady state, rated power) to the hot shutdown condition by manipulating equipment and controls outside the main control room in their intended manner. In addition, the station design does not preclude the ability to bring the station to the cold shutdown condition from outside the ,
- main control room. These actions, bringing the reactor to hot shutdown and be achieved in then establishing a variety of ways.
the cold shutdown condition, can Some of the more obvious methods which could be used are listed as follows:
1.
Methods to shut the reactor down (to hot shutdown) '
- a. Open the reactor protection system motor-generator sets' power supply breakers at the 480 v motor control centers.
AS-110
.SNPS-1 1 This causes 'all contirol rods _ to be inserted.
b.
V .. Open' the ' reactor protection system motor generator sets.' i output breakers.- This causes all rods to be inserted.
- h c.:
na Open the breakers 1to the scram pilot solenoid valves at the
.td reactor protection system buses. ' This rauses all control
.n rods to be inserted.
la d.
og De-energize ' the instrument air supply and bleed - pressure 3, from the scram valves' air supply. This causes all control i,
rods to-be inserted. '
e.
Exercise control thecan rods control rod scram be inserted testmethod.-
by this switch for each rod. All no : l f.
Actuate the turbine trip lever at the turbine front standard.
2.
Methods to achieve. the cold shutdown conditions a.-
After control rods are inserted, allow the reactor to cool-m-
by discharging steam to the main condenser
-and turbine bypass system. This process ca.via n bethe turbine controlled automatically by the pressure regulator. Automatic reactor feed-water control can maintain during cooldown and depressurization. reactor vessel water level
- b. -If the main condenser is isolated, the acICs can be initiated, monitored, and controlled manually' to remove decay !
heat and depressurize the reactor and maintain reactor vessel
- water level during cooldoun and depressurization. )
c.-
When reactor vessel pressure has decreased to about 50 psi gage the residual heat. removal system (RERS) can.be placed in the shutdown cooling mode of operation. Motor operated valves in the RERS can be operated by local keylock control
~
stations, and the' RHRS' main system pumps and service water pumps can be started by manually closing the circuit breakers for the motors. In this manner, the unit can be brought to the cold shutdoun condition.
d.
1 Reactor coolant temperature, reactor vessel pressure, and i reactor vessel water level can be monitored via indicators i in the reactor building. RERS circulating and service water i j flow can also be monitored via indicators in the reactor building.
/
I A5-111 l
SNPS-1 COMMENT 7.2:
Does the design of your reactor protection system and the instrumentation system which initiates and controls the engineered safety features conflict in any way with the requirements of IEEE Standard 2797 If so identify and provide justification for the areas of conflict.
RESPONSE
such as It is intended that the Reactor Protection Systems, reactor scram, primary Engineered Safeguard Systems, complycontainment isolation and the with the requirements IEEE 279. of A detailed evaluation of each supply is being made at this time with regardsystem within requirement listed in the GE-APED to each scope of specific IEEE 279 document. This work is expected toby publication bemid-1969.
summarized in a GE Topical Report scheduled for Certain Stone & Webster designed systems are not yet finalized.
Analysis of instrumentation for these systems, (Reactor Building Standby Ventilation System for example) will be completed with system finalization.in relation to IEEE 279, A5-112
_ _ _ _ _ _ _ _ _ _ - - - - - - - - - - - - - - - - - - - ~ _ _ _ _ _ - _ _ _ - - - - - - - - - - - - - - _ - - - - - - - - - - - - - - - - ' -
SNPS-1 COMMENT 7.3 In Fig. VII-6-1, actuating it is indicated that the
) building vent(closure monitorssignal) is logic for the reactor
( two-out-of-two. This arrangement does not appear to satisfy the single failure criterion of the IEEE Standard.
discuss. Please
RESPONSE
this facilit The logic for the Beactor Building vent monitors for taken twice (y1 of is not two out of two. The logic is a one-of-two 2 x 2) or more explicitly, .one of the two channels tripping in each of the two trip circuits. . . .
one upscale trip or one downscale trip.
The monitoring system thus criteria. satisfies the single component Please refer to SNPS PSAR Amendment 4, VII-6-1 for additional details with respect to the monitors.Section VII-6.0 and Figure
( -
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AS-113
9 SNPS-1 l COMMENT 7.4: To what extent are the water level instruments i used in the feedwater control system also used to control the HPCIS7 (pg. VII-5-2) 'Does this design satisfy Criterion 4.7 of IEEE Standard 2797 RESPONSE: No part of the Reactor Feedwater System Control System instrumentation is used in the HPCIS control. The HPCIS has )
independent instrumentation and control. The instrumentation system meets IEEE Standard 279 for redundance, channel separation, and testability. Refer to SNPS PSAR Amendment 4, Section VI-2.2.2 and Figures VI-2-2-A, VI-2-2-B, VI-2-3-A through C, and VII-5-1.
j l
In SNPS PSAR Amendment 4, Section VI-2.2.2 references terminology such as flow control, turbine control system, speed governor, control governor, etc., as described therewithin, explicitly !
refer to the HPCIS Steam Turbine and not to the Station Main )
Turbine Generator System as described in SNPS PSAR Amendment 4, Section XI-2.0.
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A5-114 !
1 L________-____-
SNPS-1 COMMElfr 7.5: Describe and evaluate the current design of the I instrumentation which is supposed to locate a break in the primary coolant recirculation- lines and to close and/or open the necessary valves to establish full LPCIS- flow, and the instrumentation which will prevent LPCIS flow from being diverted to the containment spray cooling system unless the core is flooded.
RESPONSE Please refer to SNPS PSAR Amendment 4, Section VI-
- 2. 2. 5.2,and Figs ' VI-2-6, VI-2-7-8, VI-2-8, and VI-2-9 for a thorough discussion of the LPCIS Injection System-Logic Control System.
The LPCIS Flow / Containment spray diversion interlock is described and shown in SNPS PSAR Amendment 4, Section VI-2.3.6.8 and Fiqures VI-2-6 and VI-2-78, respectively.
AS-115
COMMENT 7.6: The description of the l control rod position indicating system presented on page VII-3-3 of I
the PSAR does not seem to be consistent with that {
presented on page III-5-3 or that presented in other (e.g.,BWR plants which we have recently reviewed, Pilgrim and Cooper). Please resolve this {
apparent inconsistency. If the design has been j changed to that indicated on page provide a more detailed description VII-3-3, and an j
evaluation of this new system, including j consideration of the ACRS comments expressed on this subject in their letter to the Comrission on {
Diablo Canyon. j
{
RESPONSE: 1.0 Introduction )
The SNPS descriptions of the control rod position indication system in PSAR Amendment 4, Section III-5.2.4 (Page III-5-3), Section VII-3.2.1
- 11) (Page VII-3-3) and Appendix C, Section 2.3.5 (Page C l l
furnished with this with are in agreement one another and with the design to be facility. This design is in agreement with j
the following general comment.
facilities which have also responded to this i
a.
Bell Station, Unit 1 (AEC Docket No. 50-319) Amendment 1, C/R 6.24 and Amendment 2, C/R 6.27a, b, c, d
)
- b. i Batch 3, C/R Nuclear VII-2.0. Plant, Unit 1 (AEC Docket No. 50-321) Amendment I 4
{
With regard to the cited ACRS-concern comments referenced in the 1
Comment, the SNPS-1 design withdrawal, position indicationdoes not include ganged-rod via recorders, and external to the core nuclear instrumentation system. This station design involves single rod withdrawal action, multilevel control rod q position system and indicators a capability,an in-core nuclear instrumentation RBM system to preclude the violation of an inadvertent single operator error. )
j In order to clarify, again, the description of the control rod i position indication system in detail the following is included.
2.0 Control Rod Position Indication System - General The main control room reactor operator will have available three different displays of control rod position.
A. Control Rod Status Display B. Four Control Rod Display C. Process On-Line Computer Printout
)
These displays serve the following purposes:
(
AS-116 7
P
SNPS-1L A. Provide the reactor ~ operator and watch supervisor with a e
I-continuously available, easily digestible presentation- of each control rod's status. .
B.-
eachProvide controlcontinuously rod difficultyavailable, easily discernible warning of and its source.
C. Present numerical control rod position for each rod as required.
(When are needed due control- rod to computer motion is desired or log readings failure.)
D. Log all control rod positions on a routine basis.
. The control. rod ~ status display is located on the apper vertical section of control room. the reactor control board in the center of the main (See this Amendment, Fig. 7.6-1) . It provides - the following continuously available information for each control rod separately: control rod fully withdrawn, control rod fully '
inserted, control rod selected for . motion, scram accumulator status, scram valve position, control- rod drift alarm. Se display is presented as shown in this Amendment Figs'. 7.6-2 and 7.6-3.- Interspersed among the control rod status modules are l
high and low trip indication lights for each of the local. power range monitor (LPRM) system neutros detectors. Se control rod .
status modules and LPMI alarms are all arranged as they spatially ]
i relate to one another in an upper plan view of the core.
the reactor _ ' operator has available in an easily readable form 1hus 1
l' a
) master display which he can quickly scan to note control rod position (in or' out), the control rod selected for motion, and ,
the -nature and spatial location of any abnormal' control rod or neutron monitor- condition. During normal operation since more i than half of the control rods are fully in or fully out, the remainder of the control rods would not present a status light.
Therefore, the display is normally lit by both red (fully out) i and green (fully in) lamps and one white lamp for the selected control rod.
Less position light and thanwould there half ofbethe nocontrol alarms.rods would show no I, To provide more precise indication of control rod position when a control rod is being moved, the lower vertical portion of the reactor bench board supports the four control rod display (see this Amendment, Fig 7. dr-3) . The information presented on this display includes the LPRM values for each of the detector arrays surrounding the control rod selected (see Fig 7.6-5). Since each detector array contains four sensors in a vertical column and q
there can be a maximum of four detector arrays surrounding a rod, sixteen meters are installed.
l four control rod position modules.Between the LPRM indicators are These four modules will j
j display control rod position in two decimal digits and rod selected status (white light, off or on) for the four control )
rods located within the LPRM detector arrays being displayed. i 1
The control rod positions range from 00 to 48, representing the f ully in position with 00 ;
l increment, e.g., 00-02, equals six and 48, fully out; each even physical inches of control rod !
AS-117 l
i
moviment. Tha centrol rods presented on display t the four control rod those on' panel, the control rod status panel, the rod selectcorrespond and the core.. pushbutton The reactor correlate the control rods on the four' rod display to theiroperator cancore location. then quick The ' process on-line each control rod and prints out in acomputer receives position indication from pre-arranged sequence all control rodanytime printout positions. _
it is The reactor operator may order a computer (two digits) and pushing the printout button. desired by dialing the corzect of The printout time aboutdepicts
_ printout two minutes the is limited by the speed of typewriter. The control rod positions an in array-corresponding this Amendmentto the Fig . 7.other 6-4) displays
. and actual core location (see The printout is always in the same order, ifbythere signify eitheris aa blank missing or 99.or garbled input, the printout will The three displays Signals for the control rod status display areare essentially independent o the' control rod position information system hard wired fror buffer outfits,'so RPISC will not affectthat a signal failure of other cabinet (RPISC) this display. Likewise, theparts of computer can the fail-and the control rod status and four control rod displays will continue to function normally.- In this case, the reactor operator select the' would rodsuse the four control rod display and sequentially position for his log.so Because as to obtain individual of the digital difficulty control rod for the operator to reactor visually search out a digital position indicator on a large display (if such a display were used) shift his eyes back to hissologsheet, and on, then. relocate the next indicator on the panel, the four increase employed. in speed and reliability den this backup method sust becon and ' forth ' whileThisscanning is because the process of shifting the eyes back a large array of small indicators requires select pushbuttons, more total time than the operation of punching rod i.e., visual circuit response time. scan time is longer than the The displays will be used in the following ranner. Indication of the total core condition will be observed by a rapid visual scan of the control rod status display. If there is an abnormality on any rod it the reactor operator, at which notice he willwill be immediately alar scan the control rod- status This information panelisfor the location and nature of the difficulty.
~ operatir. continuously available to the reactor require continuously Since there isand no immediately foreseeable circumstance which would control rod digital position, only the rod selected for motion available knowledge of and one time by the four control rod display.a maximum It is expectedofthat three of its this will result in less cor.Cusion for the reactor operator.
t.S- 118
SNPS-1 ,
It is a. general principle of human engineering that data displays should I be designed so that information required for use by the man be made available in accordance with the use time requirements, i.e., indications which are necessary for immediate action should be designed for rapid assimilation; those bits of data which are not required for use in forming a rapid judgment i
may display be stored or otherwise delegated to the periphery of the data system. Of course, the assumption is made that l
extraneous or unnecessary data is not displayed in the system.
A corollary of this principle would state that where the human decision is based upon a predetermined logic, the display data should be eliminated from the man-machine interface by substitution of a logic circuit as the decision maker. The result is less tedium for the human, more time to give attention to matters of greater importance, improved system efficiency and reliability. The operator is thereby not saturated with an incomprehensible array of information, only a small portion of which is digestible in a given time period.
These principles have been successfully employed in the control rod position disprays by:
A. Condensation of the digital control rod position indicators used on previous control rod displays to the four control rod display with access by pushbutton selection.
B. Presentation of all the information required to make an
( l immediate operation decision on a continuous individual control rod display.
- c. Substitution of a digital computer to perform the centrol rod position log writing operation.
ne system design, besides resulting in a more efficient man-machine relation, also increases system reliability (less wiring by about 15 wires per rod and simpler circuitry) and, ultimately improves reactor safety.
3.0 control Rod Position Indication System - Detailed Desian A. Initiation Signals The control rod position indicating system is an operational aid to the reactor operator, designed and constructed to achieve reliable presentation of drive position information as discussed in amendment, section 2 above. The Rod Position Information System (RPIS) processes control rod position data from each control rod and routes it to: (1) the control rod status display, (2) the 4 control rod display, and (3) the process on-line computer. Since this system performs only an " Opera ting informatit. ." function and is not used for any " Protective" functions, there exist no criteria of independence for this system. Consequently, the RPIS does not or separation cont.ain
' the A5- 119
SNPS-1 degree of component or power supply redundancy, separation or '
independence required of protection systems.
Fach individual control rod position is sensed by actuation of magnetically operated position indicator switches contained in a center tube of each drive mechanism. This postion indicator probe is described in PSAR, Appendix C, Section 2.3.4.
ne drive position, full-in, full-out and overtravel position signals are carried by 29 conductors from the position indicator ,
plug on the drive to one of five containment penetrations. i outside the penetration, certain conductors are combined to minimize the effect of possible position switch failures. . A total of 19 conductors per drive are taken to the Rod Position ,
Indicating System cabinet in the station main control room. i These connections and the interrelation of the RPIS cabinet with j other components is shown diagrammatically in this Amendment Fig.
7.6-6.
The Rod Position Indicating System is a highly reliable solid state integrated circuit system. The system is comprised of )
individual circuit cards for each control rod drive, for each ]
Four Rod Display readout, and the computer input. %e RPIS I cabinet supplies the following signals:
- 1. Four Control Rod Display rod position.
- 2. Four Control Rod Display rod select light.
- 3. Control Rod Status Display, full-in and full-out lights.
- 4. Computer Control Rod position information. ,
S. RPIS inoperative signal to computer. 1 The " Rod-Out" red lamps for each control rod on the control rod status display are energized by a reed switch on each control rod l as the rod becomes fully withdrawn. This reed switch illumistes the two lamps directly using a +6 volt d-c power supply voltage.
The " Rod-In" green lamps for each control rod on the rod status display are energized from RPIS solid-state logic using the same
+6 volt de supply. The reed switch from each control rod representing the fully inserted positior is used within the RPIS to produce two "All rods in" refueling interlock signals as well as the rod status lamp output.
Control rod position data from each control rod is scanned periodically bv the RPIS and is multiplexed in serial form on to common data lines within the RPIS. The data on these comacn data lines is then transferred to particular buffer registers by a gating control signal which operates in synchronization with scanning control signal for each control red.
AS-120
1 SNPS-1 The 4 Control Rod Display is driven froe particular buffer i registers which store the current position of the selected 1 control rod and its predetermined neighbor control rods.
Similarly, the process on-line corputer input is derived from a particular buffer register which stores the current position of the selected rod (in the operator-following mode) or the I particular scanned rod (in the scan mode). The scanning of control rod position by the RPIS is asynchronous with the computer-controlled scanning of the buffer register during operation in the scan mode. i Within the RP7t c3 a let are two 6 v d- c power supplies which light the dist .sy: '; r itetas 1, 2, and 3 listed previously. Each light in the :m il Rod Display is separately fused. Rod "in" and "out" lig / a the Rod Status Display are fused in groups of l
six. Power > Jtained from the 120 v, 60-1d vital bus. The internal inte se ed circuit cards are also driven by multisource j supplies from the 120' v a-c vital bus. (See SNPS PSAR Amendment )
4, Section VIII-3, and Fig. VIII-3-1.) i The control Rod Status Display receives accumulator and scras information from the Accumulator Monitor Panel. The " scram" i signal is generated by position switches on each scram valve.
The " accumulator" signal is generated by either a float-type level switch in the accumulator which will be actuated if water leaks past the barrier in the accumulator and collects in the
[ a bottm of the accumulator, or a pressure sensor which will detect i
loss of accumulator gas pressure. These signals are discussed more fully in SNPS PSAR Amendment 4, Appendix C, Sections 3.3.2 and 3.3.4. The power supplies for the "accmulator" and " scram" lights on the control Rod Status Display are 24 y a-c transformers located in the Accumulator Monitor Panel and supplied from the 120 v, 60 cps-1d vital bus. The " scram" lights and "accutaulator" lights are each fused in four groups (eight ,
fuses). '
The drive selected and drift light on the Control Rod Status Display receive their signals from relays in the Manual Control i Relay Panel. These signals are generated from relay contacts in this panel. Four 24 y a-c power supplies are also located in this panel,one on each ccumulator bank of control, drives to power the select and drift lights. The 120 v, 60 cps, 1d vital bus power feeds ther transformers. The transformer outputs are individually fused.
l B. Status and Alarm _ Annunciators l The Control Rod Drif t indicating light for each control rod on j the rod status display is energized by a latching relay driven from the RPIS solid-state logic.
' The RPIS solid-state logic receives inputs fr(m the odd-numbered 1 e
reed switches on each control rod which represent mid positions AS-121
)
I
SNPS-1 between successive- control rod notches. Furthermore, alb. even-t numbered reed switches are also brought. into the RPIS to represent the . rod notch positions. . These position inputs are {
used in conjunction with discrete reactor. manual control irputs to determine whether any control rod is drifting out of its established even numbered control rod notch position.
Only one control rod may be selected for movement at any given time. All- other unselected control rods are monitored continuously for drif t, and will provide an output from the. RPIS colid-state logic if any control rod leaves 5 ts notch position or whenever the mid-position between successive notch positions is reached.
For the selected control rod, the drift monitor operates in 'an identical manner except when the control rod is being deliberately moved from one notch position to . another by the reactor operator. For this particular. situation, a control signal the from the selected rod is used with a control signal from reactor manual control timer to inhibit the de:ift output for the selected control rod during its travel between notch pocitions. This inhibit signal is removed when the selected rod '
setcles into its notch position at the conclusion of the control rod movement cycle.
A schematic diagram of the control rod drift circuitry for each control rod is shown in this Amendment Fig. 7.6-7.
The scram and accumulator indicating light circuitry is discussed (
in this response,Section 3.A above.
One amber light for each (marked-ACCtBO is on the reactor control panel and lights whenever either or both.of the following occurs:
- 1. Water leaks into the gas volume of the scram accumulator.
This is detected by a float switch on the gas side of the accumulator.
- 2. Pressure in the accumulator decreases to a prescribed value.
This is detected by a pressure sensor on the gas side of the !
accumulator. i When the amber 1' ~ht for the individual control rod lights, the operator can as;.craine which of the above problems exists by actuating a switch on the accumulator test panel in the reactor building. Separate lights are provided to identify either by lou pressure or water leakage condition.
Control rod overtravel position indication is provided by a separate reed switch on each control rod located 2 in. beyond the fully withdrawn position. This position is attainable only if .
the centrol rod blade uncouples from the drive mechanism.
Warning is provided by news of a horn and single alarm light om l
A5-122 l
SNPS-1 an- annunciator panel located on the vertical section of the j
reactor control board in the ststion main control room.
C. Power Supplies As ditcussed in' Section 2. A abcVe the Four Control Rod Display and the Control Rod Status Display receive inputs from three individual systems, i.e. CoM Control Rod Manual Contr 1 System Rod Position Information System, and Monitoring System. Accumulator /Scras Each of these subsystems contain one or more power The supplies for groups of individual lights on the displays.
design criteria for the control rod position indicating system, including power supplies, is to achieve operational reliability. To insure operational availability and not for safety considerations, refer to SNPS PSAR Amendment 4, VIII-3,and Fig. VIII-3-1 Section ;
where the Control Rod Position Indication a-c bus. TheSystem is shown connected to the 120 v, a-c,18, vital multisources and normal power operations. are both available during emergency l
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._____ _ - - _ - _a
_ _ - - . . - , - _ _ _ _ _ - - - - - ~ ~ _
COMMENT 7,7s Describe be ' th3 cran 2 Cf phyDic31 provided which will identification to protection system and engineeredreadily distinguish equipment, safet feature similar itemscomponents, cables, wiring, yetc.
sa fety. not- related to protection from or l
RESPONSE: {
All Class nameplates I equipment items functiou. or lubels numbers which clearly identify wiAl have distincti 7e the All. wire ends will have permanent markers equipment include an identifyingand terminalcode identifying numbers letter or forn essential equipment and r
the wirewi i ng will cable trays, condui umber. The details havesome that not been finalized; however,t or cables for essential for system form of color terminations and at sufficient coding at system this time it is anticipated facilitate ready identification. intervals along would be applied at the route to i
AS-124
un.vu- l CosetENT 7.8: - Please describe and evaluate g- to be provided to:
tho instrumentation a.
Control the flow control valves in the RHRS referred to on page VI 13 of the PSAR.
- b. Monitor and control the temperature and leval in the Standby Liquid '
Control Syster.
- c. Monitor the integrity of the core Spray. headers..
- d. Monitor reactor vessel shroud water level.
- e. Control the control room ventilation system during an accident.
- f. Automatically detect the need for aakeup water in the fnel atorage pool-and automatically initiate the 1 operation of the makeup water system.
RESPONSE: a.
The valves cited will be capable of being manually j throttled during plant ope ation to adjust flow, if necessary, to the respective RER subsystems to maintain proper flows under all l modes of RER operation.
These ' valves serve to operationally " trim
- the RER system to. the i proper design flows, and are not for the purpose of automatic feedback men control automatic usually associated with flow controlling valves.
operation is required the signal to the valves will, override any manual signal and position the valves to their preset ECCS configuration.
b.
Please refer to SNPS PSAR Amendment 4, Section VI-6.2, and Fig. ,VI-6-1 and VI-6-2 for the requested information.
- c. The integrity of each i
of the Reactor Core Spray Cooling System locP injection lines (headers)is monitored by differential pressure instrumentation [(dPIS/43a) or (dPIS/43b)) channel shown in SNPS PSAR Fig., VI-2-4, and VI-2-5, and is decribed- in SNPS PSAR, Amendment 4, Section VI-2.2.4.
i
- d. The instrumentation
!. designed to monitor the reactor vessel shroud
- 1. water level is shown in SNPS PSAR Amendment 4, Fig. VII flood, Vor an example, in order to interlock RHRS-LPCIS mode for 2/3 refer to level indicating transmitter switches instrumentation (LITS /73-A) + (LITS /73-B) .
e.
b The control hendment room ventilation system is described in SNPS PSAR 4, Section X-3.2.
The controls will be designed for a AS-125
SNPS-1 high degree of ' reliability, but will requirements not necessarily meet of IEEE the enoineered safeguard system.279, Theas this ventilation system is not an details will be developed with systeminstrumentation design. and control
- f. The makeup' controls swetion I-2 are a revision. as described in SNPS PSAR Amendment 88, There Automatic makeup is not required.
operatorwilltobe annunciators small changes ininwater the main levelcontrol in the room to alert the i i
fuel pool, so that c',rrective action may be taken by the operator.
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SNPS-1 COMMENT 7.9 You state that purging of the primary containment will be performed until the oxygen content is reduced to less than five percent. Describe and evaluate the instrumentation to be provided to monitor and control the inerting system.
RESPONSE: The need for the incorporation of the primary containment atmospheric control system in order to maintain an atmosphere of low oxygen content within the primary containment is not necessary as described in GE Topical Reports . . . ..
- 1. GE APED 5454, 1968, March, class I titled: " Metal-Water Reactions-Effects on Core Cooling and Containment" by P. W. Ianni
'2. GE APED 5654, August, 1968, Class I titled:
" Considerations Pertaining to Containment Integrity" l
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SNPS-1 4
i COMMENT.7.10: Provide the criteria which, were used'to . establish p
j L the radiation monitoring instrument locations, j
? ranges, types, and. sensitivities shown in Table VII-6-1.
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RESPONSE
I. Process Radiation Monitorina System
[. The- design: criteria used to establish ' the Process Radiation l Monitoring System instrument locations and. equipment . type, l ranges, and sensitivities are given in SNPS PSAR Amendment 4, SectionsVII-6.2.2 thru VII-6.2.6. Functional Block ' Diagrams of the subsystems are given in SNPS PSAR Amendment 4, Figure VII-6-1 i and the details on equipment ranges and sensitivities are . listed j in SNPS PSAR Amendment 4, Table VII-6-1.
- A. detailed analysis of the system instrument channels under both
! normal and accident conditions has been reported in Pilgrim E
Nuclear Power Station, Unit 1, (AEC Docket No. 50-293,) Amendment
- 5, C/R 17.0, and is directly applicable - (within a factor of 3) to this facility reference, but with the following exception.
( . The normal offgas radiation indication from this station will - bei L . 520 eps measured,at T=10 hr. or 74 cps measured at~ T=3 days.
L ' These values are based on a release rate of 83,000 pCi/sec at 10 i
hours or 18,000 WCi/sec at- 3 days which is equivalent to an offgas emission rate of 500,000 pCi/sec at 30 minutes. Range of
, - the offgas monitor is 0.1 to 108 cys.- The monitor is capable of l detecting release rates three orders of magnitude. below the
!. minimum expected offgas release rate through 3 orders of j magnitude above the maximum expected offgas release rate.
II. Area Radiation Monitoring System .
The area radiation monitoring system described in PSAR Section VII-6.3 is not required nor, in general, designed for the design basis accidents considered in PSAR Section-XIV.
The area radiation monitors will be located in areas where personnel may be required to work for extended periods of time during normal operation and will alarm when radioactivity exceeds a preset level. Pc-table radiation monitors will ' be used when doing work during maintenance operations in these areas. Area monitors are provided in the fuel storage and handling areas in accordance with 10CFR70, paragraph 70.24 (a) (1) .
i A5-128 L_ _ _ _ _ _ - _ _ - _ - _ _ - _ ___ ___ - - - - - - -
1 SNPS-1 Depending upon the specific requirements, area radiation monitors can be selected which have any one of the following ranges:
- 1. '10-a to 10a ar/hr
- 2. . 10-1 to 103 ar/hr
- 3. 10V to 10s ar/hr
- 4. 10a to'108 mr/hr The range selected for any particular area monitor is based on the radiation levels existing, with design allowances for credible in the area during normal operation operating malfunctions or incidents.
The following supplemental capabilities (not necessarily design basis) are considered .ir. the design of the specific system:
- 1. To warn of excessive gamma radiation levels in areas where nuclear fuel is stored or handlei.
- 5. T . personnel with a record and continuous indication in the station main control room of gamma radiation; levels at selected locations within the various station i
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structures.
- 3. To contribute supervisory information to the station main g I control room so that correct decisions may be made with respect to deployment of. personnel in the event of an accident. A
- 4. To assist in the detection of i movement of radioactive material unauthorized or inadvertent I Radwaste area. in the station including the
- 5. To supplement other systems including Process Monitoring System, Leak Detection, etc. , in detecting abnormal Radiation migrations of radioactive material in or from the process streams.
- 6. To provide local alarms at key points where a substantial change in radiation levels might be of immediate importance to personnel frequenting the area.
- 7. To furnish information for making surveys which are accepted practice in the industry.
The instrument presented in SNPS sensitivities for the area radiation monitors are PSAR Amendment 4, Table VII-6-1. Specific locations for these monitors have not been developed as yet. The intent the is to locate approximately 30 area radiation monitors in following types of locations: Reactor Control Rm.,
Radwaste control Room Reactor Building and Turbine Building. Operating Floors,
(
g fuel storage and handling areas, various valve operating stations, laboratories, and decontamination areas.
AS-129 l
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__,__m--- --- - - - - - - - ---'---" ' ' ' - - - - - - - ' - ' - ' - - - ' __ __ _ _ __-__ __ _ . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
SNPS-1 III. Station Site Envirc.ss Monitoring System niis system is being developed. It will include recording gamma-radiation monitors and air partientate samplers as ir..licated in SNPS PSAR Amendment 4, Table VII-6-1. It will also include dosimeters for measurement of integrated doses.
l 15-130
i SNPS-1 1 CONMENT 7.11 For each of the scram functions listed, indicate the number' of instruments used to monitor each
{
parameter and the arrangement of the trip - logic.
Bow . are the 10 percent closed signals from each
'i of the' eight steam line isolation valves combined-to give a reactor. scram signal?
RESPONSE: Please refer to the following SNPS PSAR . Amendment 4, Reactor Protection System (RPS) doctamentation:
- a. Section VII-7.1.2.1 (RPS - System Logic) ;
- c. Section VII-7.1.2.6 (RPS - System Inputs)
- e. Fig. VII-7-1-A, -B, and -C (RPS - Scrar - Tripping i
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Block Diagram) '
The above references completely- identify the number . of instrinnents used to monitor the applicable scram parameter and indicate the channel arrangement of tripping logic. Specific j i
sensors and their subchannel arrangements are shown on the PEID and FCD for their respective systems.
Please refer to SNPS PSAR Amendment 4, documentation:
- f. 1 Fig.. VII-4-5-A and -B (Nuclear Instrumentation System
,, neutron Flus Inputs) i
, .g. Fig. VII-5-1 (Reactor Water Level and Containment and !
Reactor Pressure Inputs)
- h. Fig. VII-6-1 (Main Steam Line Radiation Inputs)
- 1. Fig. VII-6-1 (Main Steam Line Valve Closure Inputs) j.
'ig. III-5-5 (Control Rod Drive System - Discharge Volume Inputs)
- k. Fig. VII-7-1C. (Turbine Stop Valve closure Inputs)
- 1. Response 7.12, this Amendment (Turbine Control Valve Closure Inputs) l Re 10 percent closed signals from each of the eight (8) main steam line isolation valves are combined to give a reactor seras l
signal in the following manner:
The position of each main steam line isolation valve is monitored by a double pole single throw switch. The two position sensing channels assiciated with a single switch provide valve position signals to different trip systems of the reactor protection system. To facilitate the description of the logic arrangement, the position sensing channels for each valve are identified as follows:
The four main steam lines (A, B, C, D) each contain an inboard valve (F022) and an outboard valve (F028).
valve isolates that particular steam line. Closure of either A5-131
SNPS-1 Each valve provides two isolated switch contacts which form eight
-(8)-neactor Protection System Trip Channels.
The setpoint for operation of these- valve switch contacts is 10
_ percent
' word closure rather than full closure; as a result, use of the
" closure" in' the valve closure, as well as following full valve description closure. includes partial The logic for these switch contacts for.a normally energized trip channel is as follows: ( . ' Denotes Logical "And")
Trip Channel A = F0227. . F028A Trip Channel C = F022B . F028B Trip Channel E = F022C . F028C Trip Channel G = F022D . F028D Trip Channel B = F022A . F028A Trip Channel D = F022C . F028C Trip Channel F = F022B . F028B Trip Channel H = PO22D . F028D These trip channels are combined into four System Trip Logic's (A1, A2, B1, B2) as.follows:
Reactor Protection Logical "Or") (+ Denotes
' Trip Logic A1 = Trip channel A + Trip Channel C Trip Logic A2 = Trip Channel E + Trip channel G Trip Logic B1 = Trip Channel .B + Trip Channel D Trip Logic B2 = Trip channel F + Trip channel H Consequently, main steam lineTrip Logic Al will be tripped when one valve of A is closed concurrently with one valve of main steam line B; Trip Logic A2 requires closure of one valve in main ne C and one valve in main steam line D; steam requires li' Trip Logic B1 closure of one valve in main in main steam line C, and Trip Logic B2 requires closure steam line A and one valve of one
. valve in main steam line B and one valve in main steam line D.
Reactor Scram, due requires that the Reactor Mode Switch be into Main Steam Line Isolation V the "Run" position and theconditions these reactor vessel pressure be greater than 600 psi gage.
exist, If l Logic Hence, A1 or A2 is tripped concurrently with Trip Logic B1 orthen T'! . th reactor designated lines are scram will result closed in thewhen following one combinations:
or more valves in the AS-132 l
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, SNPS-1 i ht (1) i 10 (2) Steam Line A . Steam Line B . Steam g'h e (3) Steam Line A . Steam Line B . Steam!
i ip (4) Steam Line A . Steam Line C . Steam Steam Line B . Steam Line C . Steam!
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A5-133
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SNPS-1 COMMENT 7.12:
What, specifically, is the " rapid" rate of closure l of the turbine control valves which will. scrar l- the reactor (pg. VII-7-5) ? What is the basis for I this value? Pow will this scram function be implemented, i.e. primary sensing element, redundancy, logic, etc.?
q RESPONSE: The. " rapid" rate valves is 0.20 seconds. of closure of the turbine control l l
Please refer to SNPS PSAR Amendment 4, Section VII-7.1 and Fig.
VII-7.1-A through c for identification of sensors, logic-channel arrangement, etc.
With the reactor and turbine-generator at rated power, fast closure of the turbine control valves can cause the reactor
.closure primaryscram, system pressure to rise. The turbine control valve fast which initiates a scram earlier than either the nuclear instrumentation system or the reactor primary system high pressure, is provided to assure i a satisfactory margin to the.
reactor core thermal-hydraulic safety i station operational transient. limit for this abnormal The scram additional counteracts any '
reactivity due to pressure change by inserting negative reactivity with the control rods. Although the reactor )
primary system high pressure scram, in conjunction with the '
primary system relief valves, is adequate to overpressurizing the reactor primary system, the turbinepreclude control valve fast closure scram provides additional margin reactor primary to the j system pressure safety limit which is protected further by the primary system safety valves.
The turbine control valve fast closure scram setting is selected to provide trip logic was timely indication of control valve fast closure. The i
chosen to ensure protection with other scram initiators. consistent reactor scram i l
s entire scram function leads the following transient
- a. Turbine-generator acceleration protection devices trip to !
i initiate turbine-control valve fast (about 0.20 second) closure.
- b. Turbine control valve fast closure is sensed by the reactor protection system, which initiates a scram.
b.
The turbine bypass valves are opened simultaneously with turbine control valve closure.
- d. Reactor vessel pressure rises to the primary system valve setpoints, causing them to open for a short period. relief steam passed by the relief valves is discharged The suppression pool. into the ;
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A5-134
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- e. The turbine bypass system 1 pressure after the primary system relief valves close. controls reactor primary systeiq For each of four turbine control valves, there is a solenoid operated fluid from fast acting pilot valve that dumps the hydraulic control valve fast closure."the control valve actuator, thus initiating " control i l
each pilot valve A position switch will be associated with Fig. VH-7-1C. ) for use in the reactor protection system (see initiate a Fast closure of any three control valves will reactor scram. J will be tested 80 percent during normal operation with load reduced to aboutEach con!
as follows:
- 1. Close one control valve slowly until plug is approximately -
1 in ,
valve. off seat, then actuate fast acting uolrnoid operated pilot i
e (other control valves will retain control) 1
- 2. Trip corresponding stop valve.
3.
Return sequence. to normal control and continue to test other valves in
'Ihe detailed logic for the reactor protection system will developed with system design. be
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COMMENT 7.13: What is the trip logic for initiates each signal which closure of " group 1" isolation valves?
RESPONSE
The trip logic for each signal which initiates closure of the " group 1" isolation valves is one out of two, twice.
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1 AS-135 L____________.. __ _ _ . _
SNPS-1 COMMENT 7.14:
To what extent will loss of the station process en-line computer restrict plant operations and what are the bases for these restrictions?
RESPONSE
Please refer to SNPS PSAR Amendment 4, Section VII-9 for details of the process on-line computer.
The process on-line computer will be an operational aid and information processor, not an engineered safety device. The manual monitoring and calculational technique methods will be adequate during periods of normal station operation when the.
power peaking does not approach limiting values. (Refer to SNPS PSAR Amendment 4, Section III-3.3.4.) The expectation attaining enhanced higher power by the use of andensity operation on a rout.ine basis of is dependent upon si + equipment.
on-line process computer, but not improved speed and accuracy with which the reactorThe enhancement results fr obtain power operator can distributions and performance index during operation. Thus, the uncertainties which are inherent values rapidly and factored will be into manual method of core monitoring and calculations reduced in process on-line computer. both time and uncertainty by the use of the If the operating peaking factors tend to become larger at some time in the reactor core life (due to less distribution as a result information, thus less optimum control rodrefined power patterns of the computer nuclear instrumentation system will being out of service), the in-core detect and control the approach of this condition, and the station reactor operator will have to reduce power to a suitable level to maintain peak thermal conditions below limits. No safety problem will exist in the {
situation described above.
Adequate backup equipment components) (normal operational systems and and/or procedures will also ensure that station safety will not be impaired or compromised during a computer outage.
This includes all the computer capabilities areas in regards to normal station operation (e.g., as data logging and annunciators).
Comments of the the following same nature facilities: have been responded to in regards to
- a. Hatch Nuclear Plant, Unit 1, Amendment 2, C/R VII-6-1 and VII-6-2 (AEC Docket No. 50-321) .
{
b.
Brunswick Steam Electric Plant, Units 1 and 2, (AEC Docket Mos 50-324 and 50-325) Supplement 3, C/R 7.3 I
AS-136
ENPS-1 '
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COMMENT 7.15:
1p You have Lindicated that. each RBM' channel will
'9 automatically be . calibrated to read the tame as.
a reference APRM prior 'to rod. withdrawal. Please 1 describe. the system that determines which .of _ the APRMs 'are' to serve; as the references. What happens when the referenced APRMs are bypassed or inoperative?' Perform an analyEisf.ito show. that
'the local protective function of the RBM cannot be compromised'by this. calibration, including the-condition where one RHM channel is . bypassed . and the- second- RBM. channelf receives 'a valid calibration signal from its reference APRM which is sensing a flux lower than that in the vicinity of the selected rod. . i RESPONSE: .}
Please refer to SNPS PSA?t> Amendment 4,. Sections'VII-4 {
and VII-7.3 and Fig.' .VII-4-5-A and B, an Fig. VII-7-2-A and B for:
more detailed description of this system. i
'u Upon selection of a control rod for withdrawal, each RBM assigned to monitor that control rod is automatically calibrated to - read the same as the reference APRM.
The two RBM channels, although onif serving as operational aids, )
are assigned one channel to each trip system. .The reference APRM signal is permanently. wired to the RBM and the arrangement is )
su' that ~if; the reference APRM is bypassed another APRM in the same An inoperative trip system automatically. furnitihes the reference signal.
reference APRM, as well as any other inoperative j APRM, not be will cause a rod block alarm on its own and . thus it will 4
> possible to' withdraw a. control rod until the inoperative condition on the APRM has been corrected or that particular APRM channel is procedurally bypassed.
The response of the RBM to the worst case single contrd rod withdrawal error is such that, if the local relative power change remains within the preestablished amount for which control rod withdrawal is allow *:d, the RBM will meet its desig when calibrated in terms of the core average power.n requirements This automatic calibration is such that, when a control rod is selected,(althcugh block the RBM not gain adjustment function is initiated and a_ rod implemented.
immediately alarmed) is simnitane 1 sly calibration period. This control rod block remains in place during-(1 to 3 seconds). The gain of the RBMthe
. is varied starting at a minimum value of 1 te its maximum 8 continually comparing the output value obtained value with of the reference APRM value. If the RBM i.hternal averaging amplifier
! output of value is greater than the APRM reading at the initial gain 1, the gain is automatically set at this minimum value. This case arises if the local relative flux average is higher than the ,
core average flur and for this situation, the RBM Ds made to read l o the more conservative, i.e. , higher value. If the RBM output is i initially less than the average powe,r, the range of gains is I AE-137
Y
, SNPS-1 scanned and. the. outputs. compared until the proper value is obtained. 'If for some reason the gain' cannot be adjusted to the proper .value within two tries, the RBM is automatically made inoperativeland withdrawal of the selected control rod is not permitted.
one .RBM rod Since' only block signal is necessary to prohibit,
- . control- rod withdrawal', it is' permissible to bypass either of the two RBM's and still meet the design objectives of the _ Rod Block Monitoring System.
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I AS-138
SNPS-1 COMMENT 7.16: . lease provide 'a description oof the proposed
- design for the Recirculation Flow Monitors.
should include the number of monitors toThis
! be j provided and their relationship to the RBM and APRM systems. Also, include an evaluation to show that the power-to-flow rod withdrawal interlock rod block and control features the single failure criterion. these systems meet of can the recirculation block?
flow comparators initiate a rod If so, provide a description i
and an evaluation loop of this function for both one and two-operation.
RESPONdE: Please refer to PSAR Amendment 4, Figure IV-2-10 for design of the Reactor Coolant Recirculation System Flow Monitors.
Refer also to PSAR Amendment 4, Figure VII-7-2A and B for Scram and Rod block action of the RBM and APRM.
The instruments which utilize based on reactor coolant control rod block trip settings Average recirculation system flow are the MonitoringPower System Range Monitoring System (APRM) and the Rod Block (RBM). The details of the instruments used to provide method and control rod block settings are as follows: the reference flow signals for the a.
on A differential pressure signal is derived from a flow sensor each of the two
, loops. This differentialreactor coolant recirculation system flow pressure si current signal through the use of a (gnal is converted to a (AP) to (I) converters are arranged soAP)that to (I) converter. The the differential pressure signal from each loop is furnished to a square root extractor associated with each of the two logic channels. Refer to this Amendment fig. 7.16-1).
The conditioned signals from each of the two loops are added together in a flow summing unit which coolantprovides an output recirculation system signal flow.proportional to the total reactor the These flow signals are fed to place. appropriate flow converters where further conditioning takes ,
i The conditione'd flow converter output signals are used as the reference signals for the control rod block trip units of both the APRM and RBM.
{
The f ailure of the flow convertor signals or the flow-measurement system 1
l and APRMitself does not impede the necessary operation of the RBM Systems. The following is cited: ]
- a. Analyses presented in the SNPS PSAR Amendment 4, Sections VII and XIV show that the reactor protection system is adequate to prevent excessive fuel failure in the event of a single reactor operator error or single equipment malfunction. The Rod Block Monitor (RBM) system is provided as an operational aid to insure I operation within prescribed control rod patterns. Rod patterns outside malf unctions. these are thus single operator errors or single equipment Therefore, both RBM malf unction due to reactor AS-139
l-i SNPS-1 j
~ coolant recirculation. system flow . monitoring malfunction plus operator or additional equipment malfunction is not credible.
- b. Based on a complete evaluation of the reactor dynamic performance and considering during normal power operation, expected maneuvers, .'
various mechanical malfunctions or a single operator.
be provided error it was concluded that sufficient protection will by the Average Power which is provided with a fixed upscale Range Monitoring (APRM) System approximately 120 percent of reactor level SCRAM trip at equipment will not be rated power. Thus, the furnished with variable flow-scram trip feature. the. capability for a c.- There will be no provisions recirculation flow. for a SCRAM trip due to low
- d. There are, however, features in the reactor coolant recirculation system flow converter circuitry which provide for control rod block and alarm in the event one of the two flow converters shows a low recirculation flow indicator.
- e. The recirculation flow converters f v nish a rod block under conditions of an upscale er instrument inoperative if condition or the ' comparison a predetermined between the two flow converter outputs exceeds amount.
Each flow converter receives a from each of the two flow loop and loop flows.
the indication is proportional to the sumsignal In the event of single loop operation, the rated) associated with the loop inthe flow converter would indicate amount of flow (portion of operation.
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AS-140 l
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i SNPS-1 COMMENT 7.17: Figure X-2-3A indicates that three signals will 8
automatically . isolate the Reactor Cleanup and Demineralized System while on page X-2-12 it is stated that isolation is accomplished by manual operation. .Please resolve this inconsistency.
RESPONSE: Refer to SNPS PSAR Alaendment 4, Section X-2.3.3 and Fig. X-2-3A and X-2-3C.
Isolation of the reactor cleanup derrineralizer system is automatic but also it can be initiated manually from the station main. control room. The system is isolated (via system isolation valve MOFOO1, MOFOO4 and MOF042) automatically only upon receipt of the following signals:
- 1. Reactor vessel low water level
- 2. Standby liquid control system actuation (pumps running)
- 3. High inlet temperature to the filter demineralizers (high heat exchanger-discharge temp) e i
/
A5-141
SNps-t
- 8. O COMMENT 8.1:
ELECTRICAL POWER SifSTEN What special provisions will be mad of the speed andemergency diesel e.in the desigt generators to en What source reliability of of continued operation' .
these units? starting power will be used;fot.
requirements of How will the reliability conflicting dieselpressure, oil gene rator or protective devices suchof e as lot temperatures be resolved?.high coolant or bearing
RESPONSE
accelerating (10) seconds. to rated speed and voltage y to load readThe starting and d the speed and reliability of engine stThe engines providin willten ssed air and be systems. g for each arting will be enhanced by Each andstarting system consisting of airengine, t reservoir (s) air engine. compressor .
is capable of start circuit,.
starting the two fuel systems between one s n iseach enhanced engineThe by re and its associated day tank.providi and ystem the is other uses station battery. supplied a by a fuel pump driven off t ;
requirements to Either d-c fuel motor system driven can pasp operating off the thedaystorage tank. tank will flow by gravity toperate the fuel the o maintain the Fuelengine oil fromat fu level in Engine protective devices with generator differential will not trip the engithe operation. exception of ov ne during emergency and high coolant or bearingEngine protective devices provide includin pressure the engine temperature will be action. which permits audible and visual alarm in the connected main to contro The the room and at engine operators to take appropriate percent of rated speed, and theoverspeed trip, both generator which will be set at 120 of which action, are the detect onl very unusual conditions differential requi i relays, emergency operation. y devices which r ng immediate trip the engine during l
1 A5- 142
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SNPS-1 COMMENT 8.2 What assurance Shoreham is there generator or thethat the loss of. the tn j~ '
generating. unit in your system the willloss not of the largest !
external grid to the ' perturb power would not be extent that offsite 9 engineered safety available to operate the the total system will features? What percentage of
" capacity - represent? the Shoreham generating 9 RESPONSE:
The . loss' of
' the external ~ grid in any manner . thatthe Shoreham generator will not p
-d losstion Se of both normalL and reserve offsitecould VIII-2-1, power.cause simultaneous for the As stated in PSAR the design basis
]' transmission supplies will be to provide a system 138Kv and 69Kv Y reasonably.
69Kv conceivable transmission multiple failure systems,. nor single in the such that138Kv external no r
3 results in failure onsite, .which' !)
.', - will cause a simultaneous interruption of supplies the powerth , )
both the NSS and RSS transformers. to'.
{
Stability j I that all studies the- that have transmission been completed in the past indicate fulfilled. system design objectives will be 4 i
detail the Studies now in progress will confirm and document in satisfied. necessary assurance Upon completion of that the all. objectives will. be study, i
t August
- 1969, a . summary of the results will be now scheduled for f ,
submitted to the The percent of the LILCO system capacity in 1975.Shoreh y 22 {
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A5-143 i l
SNPS-1 COMMENT 8.3:
Under what conditions may the emergency buses be interconnected? Will the interconnection of these buses be manual or autorratic?
RESPONSE: Please refer to PSAR Fig. VIII 1. The 14160v emergency buses 101, 102- and 103, which supply the emergency shutdown loads, will not have bus- tie breakers .between them.
Each.' bus will have its own supply breaker from both the NSS and RSS transformers and each . bus section will be located- in a.
separate room of a class I building so that trouble at one bus would not involve the When operating on standby others either mechanically or electrically.
diesel - generator power, under the remote possibility of loss of all offsite power, the NSS and RSS transformer supply breakers will be open and the buscs will be electrically separated. ,
)
The 480 v. buses 104, 105, and 106 which supply the low voltage auxiliaries required for emergency shutdown, will be supplied 4 from stepdown transformers connected to 4160v buses 101, 102, and.
103 respectively. These low voltage buses will be provided bus tie breakers that will be normally in the open position. with_ The i^
-480v bus tie breakers may be closed manually, under operator control, if necessary for maintenance of a stepdowwn transformer or 4160v transformer feeder. The 480v buses 104 and 105 which supply. duplicate- auxiliaries will be completely. independent.-
480v bus 106 which supplies vital equipment that is not duplicated trip the will be provided with automatic transfer equipment to normal supply and transfer to an alternate supply from either 480v bus 104 or 105 on loss of voltage.
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SNPS-1 k
1 COMMENT 8.4: How long -can a diesel ' generator operate' ~ vithout-
'g service cooling water flow: aow does this time j . period relate to the time for the service water pumps to be sequenced into operation when the only source 1 of electric power is the diesel j generator (s)?
l RESPONSE: The diesel engines considered for this' project make use of a closed loop cooling water system. The engines can operate for approximately 3 min: at. full load without service-water flow. When an emergency start signal. is received, under design basis loss of coolant- accident conditions, the diesel generators start and. accelerate to rated speed and all emergency core cooling system load is automat'cally programmed within 40 sec .(Refer to SNPS PSAR, Amendment 4, Table VIII-4-2) . This allows ample time for the service water pumps to be started after
.the ECCS load has been applied.
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i A5- 145
-SNPS-1 COMMENT 8.5:
Please describe and evaluate the controls i actuate switches and which offsite power supply which provides powerthe tocirc engineered (normal) safety features upon loss of the main include generator the time supply.
The switches. required- toevaluation actuate should these RESPONSES 1.
. The auxiliaryPlease refer .to Section VIII-3-2 and Fig power system for. .
VIII designed so that the' normal' will station. service Shoreham station will be (NSS) transformer circuit breaker thus making it independent o This pr.rmits shutdown the MSS- transformer .to be or oil rator.
transmission connection isas system available.
long as the transformer is operation will be highly The 138kv transmission that the NSS transformer will be available to saf lreliable and it e y shutdown the i trips originating in the reactor orracor turbine systems.
orgenest all The 138kv generator leads including connection and the NSS transformer high speed differential relays.which will initiate of the ce y fast auxiliaries for trouble within this important area transfer 111 of the 4160v buses will be transfered from the N following the reserve station service (RSS) transformer n six withi r to .
cycles, MSS transformer or the 138kv generatoran operation of the diffe transfer. connection. This fast with storedis possible energy because operating mechanisms the 4160v breakers will be equip transfer before that complete the running _ motors have the enough to cause transient torque or inrush problems. dropped out of step A second form of transfer initiated by low voltage will random transformer. transfer of the 416Cv buses from permit the probable causes for the simultaneousThis type of transfer w loss of the- Shoreham the aforementioned NSS transformer and 138kv e in ge which transfer initiate be transfer, are extremely limited. or connection will fast approximately 20 percent of normal whichdelayed until residual volt motors permits the voltage may be out of phase.to be reenergized even essential thoughupplyresidua large Non-essential loads, so The that inrush current will not producemotor loads, will be tripp excessive voltage dip.
on the motor characteristics and thetime for residual voltage t epend loads and inertia of the connected at least 20 cycles (60 cps time base)in addition, low voltage trans kv system backup relays which might to permit operation of 138 voltage.
It is therefore anticipated that clear low the cause of low voltage transfer n
of essential auxiliaries will be completed within a time range o AS- 146
SNPS-1 approximately 30 to 120 voltage to decay. cycles depending on time for residual The standby accident- diesellow or from generators will be started bya signal' from
. speed and voltage in approximately 10 sec. bus voltage and will accelerate to If voltage of an essential above bus all then has not been restored by one the schemes described supply to the 480v load center, will4160v breakers at the essential bu diesel generator. connected to the bus. be tripped open' and the l
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A5-147 :
SNPS-1 9.0 RADIOACTIVE itASTE SYSTEMS COMMErr 9.1:
. One plant'sof your wasteprincipal design criteria is that the
- 20. One of thewill be in acccrdance with 10 CFR 20.10 6.b (1) , requirements of is that the applicant10makes CFR 20, reasonable effort to minimize the radioactivitya contained in effluents to Please explain how the use of unrestricted areas.
6emineralizers regenerative type in the radwaste and spent fuel pool rather cooling systems for the Shoreham plant, is consistent with these objectivesspite the inthan nonre radioactive liquids that of regenerating these demineralizers.will result from
RESPONSE
i IX-1.3, Amendment 4 of SNPS PSAR.The response to. Secti this on commen AS- 148 i
SNPS-1 COMMENT 9.2: Please clarify whether the radiation monitor on C ,f . j the ' discharge pipe from the radioactive liquid waste system will automatically terminate discharge from the system upon detecting a high concentration of radioactivity.
RESPONSE: The response to this comment will be found in Section IX-2.2, Amendment 4 of SNPS PSAR.
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t SNPS-1 COMMENT 9.3:
The PSAR indicates that the liquid radwaste sample tanks and. the floor drain sample tank will be located outside the radwaste building.
What are the maximum quantities of radioactive wastes that could these be numbers?
in these tanks and what is the basis for Analyze the potential radiological tanks. consequences of the rupture of any
RESPONSE
The response to this comment will be found in Section IX-2.2, Amendment 4, of SNPS PSAR.
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SNPS-1 .
COMMENT 9.4: Which. portions of the liquid radwaste
' Class system are I
and which. are class II? What are the maximum quantities 'of radioactive wastes that can be contained within the. Class ~ II portions of'. the system and what is the basis for these numbers?
Analyze the. potential radiological consequences of failures in the. Class- II portions of the system.
RESPONSE: The response to this . comment will be found in Section IX-2.2, Amendment 4, of SNPS PSAR.
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AS-151
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SNPS-1 1
COMMENT 9. 5 : Please provide the assumptions and calculations, {
and their bases, which were used to evaluate the i
radioactive liquid and gaseous release rate j estimates for normal reactor operation. 1 RESPONSE: Based on current Radwaste System design, the chief sources of liquid waste being discharged are the Chemical and Waste collector Tanks. On the basis of estimated flowsWaste and j
activities, both of these sources are assumed to have a maximum ;
l daily discharge activity of 6 x 105 DCi/ day, making a total of 1.2 x 10* DCi/ day delivered to the radwaste filter through which the total effluent must pass.
Further, assuming a conservative value of 50 percent efficiency for the radwaste filter, the . maximum daily discharge activity would be about 6 x 10s pCi/ day based. on a fuel leak rate equivalent to a reactor water activity concentration of 2.4 pCi/ml and .a noble gas release rate of 100,000 pci/sec at 30 min decay.
Since the maximum Shoreham noble gas release rate is estimated to be 500,000 pci/sec at 30 min decay, the maximum daily activity of these radionuclides which are fission products have been multiplied by a factor of 5, resulting in a total maximum activity discharge of 9.9 x 10s pCi/ day, exclusive of tritium.
Discharge concentrations have been calculated on the basis of an assumed total discharge flow of 480,000 gpa (450,000 gpa circulating water and 30,000 gpa service water).
values of the above parameters for the radiologically significant isotopes have been tabulated in Table IX-2-1, Amendment 4, of the SNPS PSAR, and are compared to the corresponding limits imposed by 10CFR20.
Gaseous release rates are discussed in detail in Section IX-3, Table IX-3-1, Amendment 4, SNPS PSAR.
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SNPS-1 l
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l COMMENT 9.6: Provide the following information l t estimated concerning tritium releases from this facility i
i during reactor operation:
- a. The assumptions, calculational j methods, and bases for the tritium '
release rate figures quoted in Table IX-2-1,
- i. b.
The maximum tritium release rates that '
could occur during normal reactor. .
operation, and how these values were determined.
c.
! Describe the means that will be used to monitor the tritium levels in the radwaste systems prior to release to the environment.
RES PONSE:- l (a) The following sources of tritium are considered: l (1) Ternary I burnable poisons fission - fuel leakage (2) Activation of-(1) Ternary Fission - Fuel Leakage It is assumed that 0.01 percent of all fissions are ternary and produce I tritons (Reference A - Pg. 21) .
that tritons produced in a leaky fuel element It is further assumed completely to the are released elements can be determinedreactor coolant and that the number of leaky release rating by 1000 pCi/sec.
by dividing the 30-min noble gas (2) Activation of Burnable Poisons It is assumed that poisons does not leak or tritium produced by activation of burnable B - Pg. 20).
diffuse through claciing (Reference CALCULATIONS
_ Percent Puel Defect
. = 500,000 uci/see x 1005 = 1.82%
1000 pCi/sec per rod 27,440 rods Total Release from Fuel
= 1.82% Defects l
AS- 153
e SNPS-1 Tritium Produced by Ternary Fiesion N='
2550 Mwt x 10' w/Mw x 3.1 x 1010 Fissions x 10-*
T w - sec~ ' Fission-
=
AT =
7.905 x 105 atoms /sec
_. 6 9 3'
- 12. 2 6 yr x 3.15 4 x .107 sec/yr = 1.792 x-10-8 sec-1 A= AN =
_1. 792 x 10-* sec-1 x 7. 90 5 x 1015 3.7 x 10* pCi/ dis - sec . atoms /sec =383 Ci/das
~
~
=
383 UC1/sec x 8.64 x 10* sec/ day = .3.31 x 101 pCi/ day Tritium Released to Discharae Pipe -
= (Assuming equilibrium) 3.31 x 10' uCi/ day x 0.0182
= 6 x 10s uCi/ day
_ Service and circulating Water Flow 450,000 GPM (Cond. Circ.)
+ 30,000 GPM (Service) = 480,000GPM 480,000 GPM x 3785 ml/ gal x 1440 min / day = 2 6 .
x losa al/ day Tritium Concentration in Discharoe Pipe __
=
6 x 10s uC_i =
2.6 x 10ma ml 2.3 x 10-7 pCi/ml It should be doubled, as shown in Table IX-2-1, . Amendment 4noted tha tritium 3x 10-3 release would be many orders of magnitude SNPS PShR, the UCi/m1 allowed by 10CFR20. below the l
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AS- 154
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L SNPS-1 REFERENCES A.
to TID-24635 " Sources of Tritium and its Behavior upon Release the Environment", An AEC Critical Review Publication, Dec.
1968.
B. Docket 50-300, Exhibit B-3,."3rd Supplement to PSAR, Easton Nuclear Station, Question No. 7, " Tritium Controls".
b.' Refer to Response 9.6a references.
- c. Refer. to Response 9.64..
Since the credible. tritium releases from a BWR (regardless of power level) are. negligible, continuous or periodic operational monitoring is not necessary.
Occasional air or liquid grab samples from various station a
process -systems streams may be collected and . analyzed offsite at a commercial radiochemical lab for tritius concentrations.
Liquid scintillation multichannel B-ray instrument equipment, e.g.
can
, Packard Instrument Company - Tri-Carb, used by these labs detect tritium concentration levels as low as 1 x 105*
pCi/ml).
This instrument (2 x 10-8 pCiAal)is by more sentitive a factor of 1than the MPC levels of x 108 10CFR20 O
AS- 155
r- -
SNPS-1 COMMENT 9.7:
The' PSAR states a wind ~ tunnel test ' program will be instituted to determine whether .the atmospheric dilution L will: correspond toof gaseous releases from the plant vent.
!. than a ground level an elevated release rather release, eve: tough the l plant vent is approximately the same theight - as
-e the surrounding. buildings.
l :. . results You have reported the of _this_ test program Please document these results, as to ~
us orally.
well important details of as the L-n the . environment surface roughness,this program, such l, proximity- of Long Island- Sound, topography, etc. , were modeled in the wind tunnel.
RESPONSE
the characteristics ~ ofWind tunnel studies have been performed fo
< operating conditions. station exhausts during all potential of some for the Shoreham tests, the nearby Nuclear Power topography. Station building com ,
. The test program was aimed at evaluating the effluents for two conditions, namely: characteristics of A.'
t i Potential design basis accident conditions with the offsite reactor power building isolated and no available.
B.
Normal, full power, station operations Previously, of tests it was orally reported to the AEC-DRL that the result at all wind speeds forfor Condition A was that some entrainment was octaine conditions. the neutral stratification atmospheric elevated release This points conclusion is appropriate for effluent from the exhaust .(at 50 ft/sec) a situation where about a 1,000 cfm 1 the reactor building.
is released Table 9.7-1at 10 ft above the. center of I data for this case. and Fig. 9.7-1 provide the Entrainment, as utilized herein, whereby .the pertains to the condition effluent an to deviate isolated stack.
significantly from a dispersal obtainedp with "significant" the plume and deviation The criteria used to determine occurs were both a visual observation when a of the ground occurs. measurement of where, if any, plume contact with the PSAR.* Downwash Such a method was previously referenced in
. entrainment whereby the plume is considered does not aspenetrate a condition of ' severe ,
4 cavity the building region adjacent to the building. streamline but rather ventilates {d
- Davies, P.O., A.L.,
Behavior of Effluent Emitted and Moore, D.J., " Experiments from on the t: Level of Tall 513-533, Reactor Buildings," Int. J. Air Wat.PollStacks at or Near the Roo 1964. {
., Vol 8, pp 1
AS-156 1
SNPS-1 ba ic Further wind tunnel tests were conducted in an attempt to at Preclude, without use of, f auxiliary equipment, any entrainment at er wind speeds of 4 m/sec or less and for any wind direction. Se
,3 results were inconclusive. This approach was discontinued when
,g it became apparent that, in order to claim and prove any benefits w from the data, a considerable expansion of the research study
,, would be required. Consequently, accident dose calculations were 33 based on dispersal with full downwash for all wind speeds and all 3, wind directions. Furthermore, the AEC-DRL recommendations for
,f modifying the volume source dispersal equation to account for a
- d 300 meter site boundary were followed.
For the wind tunnel test program on normal station operations, g Condition B, excellent plume dispersal was obtained for low wind 1 speeds and all wind directions. These test conditions correspond 1 to daily station operations with releases from the radwaste off-gas holdup system mixed, before release to the environs, with normal ventilation air flows from the turbine, reactor and radwaste buildings. The results of these tests will be utilized g in deriving a conservative limit for releases from the radwaste offgas holdup system in order to comply with the reference annual doses given in 10CFR20.
He test data and conditions simulated for normal station operations are presented in the Exhibit C for this reply.
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UNPS-1 COMMENT.9.8: You indicated '(pg. IX-3-4) .that' you introduce the plan to outlet dischargeof the turbine buildingturbine-condenser vent fans offgas to ~ the which Which of. the monitors described intop Fig.of the reactor build VII 1 will failure be used to monitor this effluent? Could the in radioactive turbine offgas being dischargedof building area? into the the tu that two monitoring points areFig. provided offgas IX-3-1 indicates system each in the How do these monitors to systems 'theconsisting of two monitors.
relate
RESPONSE
of-release monitors described on page IX-3 station ' The normal. ventilation exhausts structures will be combined into from the separate ventilation exhaust duct as shown in Fig. 9.8-1one common station radiation .
The associated air exhausts and the offgas system are identified in Discharges diluted with fram the. the radwaste offgas holdup system turbine will s figure..
- first be then mixed with the air flowsbuilding ventilation exhaust air and building prior to release to the from the radwaste and reactor emirons.
As indicated in Figure operated building dampers will be installed at the outlet the ofX-3-3 of Amendm the associventilation system fans. turbine the spare,ated motor operated 100% capacity, fan will bedamper actuatedwill close automatical The the motor same operated time, damper associated with the automatically.
spar open e fan would, at ventilation air to continue automatically thereby enabling to flow without interruption Should the the second fan fail, its thereby stopping associated damper will also close air flow altogether. This type of an installation precludes radioactive off ga directly into the turbine building area. s from being discharged Response to the wind tunnel AEC Comment 9.7 (this Amendment) test program for normal provides data from conditions. The test station l cfm indicate that no'downwashdata for ventilation releases ofoperating 330,000 1 station occurs in the vicinity 22 mph). structures for wind speeds of up to 10 meters /sec of the series wasThe 330,000 cfm volumetric flow used for this (about ventilation exhaust flows. based on a preliminary estimatetest of the total st
- ventilation exhaust rate issystems not expected hastonot been completed, the differ total ventil value. significantly from a on this The ventilation air inlets located at areas where potential intake offor the station structures will be minimized. Contamination contaminated air is expected to present a significantof air inlet ventilation flows is not station operating problem ,
t A5-158
SNPS-1 considering also the site meteorology.both the above-referenced wind tunnel tes PSAR Amendment 4,
~ indicates that on an annual basisTahie 11-9-10StiPS of ft" elevation 10.9 (for all directions) are ex,pected tothe wind speeds at a 355 meters /sec-
. higher frequency w(about . 24 mph) for 92 percent. of be less than-the time. A corresponds to the ould be expected for the effluent discharge point. for the 240 f t elevation which ventilation exhaust duct. station l
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AS-159 t______________
SNPS-1 COMMENT 9.9 Bow 3-8) will the radwaste building vent' system (
pg. X-be monitored for radioactivity? If not monitored, your design. provide a justification toitsupportis RESPONSE: The monitored prior radwaste building system will be to discharge to the atmosphere. ventilation air The will be dual channel.
Response 9.8, Fig. 9. 8- 1. Please refer to this Amendment, monit 9 .
AS- 160
_ _ _ _ _ _ _ _ _ - _ _ - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - ~
SNPS-1 Table 9 7-1 MINDW. DISMS IN bETERS FROM REACTOR BUILDING FOR PLUf2 CONTACT WITH GROUND V3 = 15.2 tes 1#
- 10. _ f. - -150 5 g g 2 n N ne 24 00 250 100 4 None 120
- 240 60 6 ,
120 300 30 150 45 g
- 105 275 60 45 10
- 90 150 , ,
60 60
- 15 *
+ *
- V. : Free stream vind speed, seters per see V, :
Stack exit velocity, meters per sec 9 : Wind approach angle, degrees
- Complete grour.d.
contact downstream of reactor buildinc i
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SNPS-1 10.0 AUXILIARY SYSTEMS COMMENT ' 10.1:
Please provide an analysis to show that no single pipe rupture, could incapacitate any of the station's auxiliary cooling water systems which are essential for- safe plant shutdown or to mitigate the consequences of accidents service (RHR -
water system, station service water system, and reactor building cooling system). water RESPONSE: a. RHR X-4-2, Amendment 4, Service Water, per the description in Section in SNPS PSAR: two
, including piping. and valves are provided.independent sys tems ,
redundancy, single failure will not prevent Thus, through full a safe shutdown. plant
- b. Station Service and displayed schematically in Fig. X-4-1 inWater System as described in Se Amendment 4, SNPS PSAR ': a header supply with redundant piping and valving to the Reactor Building and to the Diesel-Generator Coolers is provided so that for 100 a safe percent plant of required capacity will always be available shutdown.
- c. Reactor Building Closed Loop Cooling Water System: the two loop systems Amendment 4, described in Section X-4-3 and shown in Fig. X-4-2, SNPS PSAR are so designed that duplicate units are cooled from separate _ loop header systems. Pipe ruptures or failures in the plant. any one loop will not prevent the safe shutdown of AS- 162
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g 11.0 POWER CONVERSION SYSTEM COMMENT 11.1: What signals open and close the main turbine bypass valves? What other special requirements or inhibiting signals, e.g., " lift limiting devices" (pg. XIV-2-12), are involved in the control logic for these valves? r
RESPONSE
1, which is attached to this General Electric Co. Control Block Diagram, Fig. 11.1-i response, describes the control devices required to operate the Turbine Bypass Valves. The " Lift Limiting Device" is shown as the " Combined Max Flow Limit. " The device is used in conjunction with the " Combined Max Flow Set" to limit the combined bypass valve flow to a preset value between 95 and 125 percent of steam flow referenced to the turbine design flow.
The turbine bypass valves are operated by the fc11owing control devices:
- a. By one of two redundant pressure regulators; one regulator is a backup to the controlling regulator having identical characteristics. Their duties are interchangeable,
- b. By the remote manual opening jack; control and indication will be located in the main control room.
I 4
- c. By the turbine control system. Demand signal from turbine control system is proportional to the differential between the turbine control valve opening required by the pressure regulator and the actual turbine control valve position.
- d. By loss of hydraulic or electric power,
- e. By an independent vacuum trip device.
f.
marked "Combd. Max By setting a total steam flow limit at the control station Flow Set." This limit consists of two circuits controlled by a single potentiometer. One signal is used to form a low value gate with a pressure control signal to establish maximum flow signal to the control valves. The second signal establishes a limit on the bypass flow as a function of I control valve flow. The algebraic summation of the C.V. flow and maximum flow bias signals form another low value gate to modify the bypass valve flow signal.
l A5-163 i
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4 SNPS-1 COMMENT 11.2: The four main steam lines and two feedwater lines run parallel and in close proximity up to and somewhat beyond where they penetrate the primary containment. What design provisions will assure that the failure of one of these lines, including the jet forces, thrust loads, pipe whip and the.
release of high temperature steam and/or water that may be associated with such a failure, could not induce failures in adjacent lines, isolation valves or other critical components?
RESPONSE: The design will incorporate anchors or stops located to limit movements .of the pipe. These stops will be designed to withstand the jet forces and thrust loads and prevent pipe whip l . associated with the clean break of any pipe and thus preclude L progressive failures from this type of malfunction.
In addition to the piping, the only other critical components in the area of concern are the feedwater and steam isolation valves, containment-penetrations, and containment liner- plate. Penetrations are covered in section V-2.4.7 and the containment liner is ' covered in Sections V- 2. 4. 6.1 and V- 2.4. 6. 2, SNPS PSAR, . Amendment No. 4, respectively. Rupture of a steam line or feedwater line in the -area in question would be autcmiatically detected and isolation valves would closed. In the event that hign temperature steam or water induces a failure in an adjacent line or isolation valve, either inside or outside the containment, the use of redundant isolation valves will ensure flow stoppage and the integrity of the containment will be maintained.
o D
A S- 164
- m. .
SNPS-1 12.0 STRUCTURES AND SHIELDING COMMENT 12.1: Your design criteria for Class I equipment and equipment supports combinations there state that for certain loading will failure," (pg. XII 56 6) . How, be "no functional will you implement this specifically, criterion in terms of absorption capability? limitations on stress, defor RESPONSE: The design implementation criteria and the planned of the design criteria design
' limitations on stress, deformation, with reference - to capabilities of the and/or energy absorption Amendment 4, Appendix Class I equipment are defined in SNPS PSAR description D Loading Criteria. Definitions or are cited in Appendix D and Section XII-2of "no functional failure" Amendment, Response 1.2. as well as in this This requested information has supplied to the AEC on ... been previously documented and a) Cooper Nuclear Station, Unit 1 Amendment 6, C/R I, II, III, and V(AEC Docket No. 50-298) b)
! Pilgrim Nuclear Power Station, Unit i
. No. 50-293) I (AEC Docket Amendment 6, C/R 6.0, Amendment 8, C/R 1.0 and 2 c)
Bell Station,1,Unit Amendment C/R 18.3 (ABC Docket No. 50-319) '
d)
Duane Arnold Energy Center, Unit !
No. 50-331) 1 (AEC Docket '
PSAR Section, Appendix C.
e)
Brunswick Steam Electric Plant, Unit 1 and 2, Docket Nos. 50-324, 50-325) (AEC l
4, C/R 4.7 PSAR Section V and Supplement
. 4 I
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A5-165 1
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SNPS-1 COMMENT.12. 2 : Please identify any Class I systems or which components are housed in, supported by, or are adjacent to a class II structure, and indicate why the integrity of any such Class I items could not be jeopardized by failure of the associated Class II structure.
RESPONSE: The Response to this Comment will be found in Section XII-2.1,. Amendment 4, SNPS PSAR.
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AS-166 {
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! COMMENT'12.3:
Identify any structures having both class I and Class II elements and indicate the design criteria that will be used for these structures.
RESPONSE: The Response to this ' Comment will be found in Section . I III-2.1,. Amendment 4, SNPS PSAR.
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SNPS-1 j
COMMENT 12.4t Please estimate the differential dynamic displacements due te seismic effects between the
' reactor building, the primary contairusent and the m&jor structures and components primary containment.- within the I I RESPONSEt The Response to this Comment will be found in SecticnE i XII-2.1 and V-2.3, Amendment 4, GNPS PSAR. Detailed design has I not progressed sufficiently to present numerical vaines at- this l time. They will be submitted in a later Amendment. "
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AS-168 t
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SNPS-1 ic COMMENT 12.5: We need- additional information on the design for
' h2 differential thermal expansion of the primary containment, the NSSS at.d other major structures hf and components within the primary contcinzent, the reactor building, and fuel storage pool.
i Indicate whether steady state or transient thermal gradients are critical and the order of- a
."l magnitude of corresponding stresses.
]
iO' RESPONSE: Detailed thermal analysis has not progressed sufficiently to evaluate the magnitude of the stresses caused by these loads. This information will be submitted in a later amendment.
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SNPS-1 COMMENT T2.6:
Discuss the possible magnitudes and consequences due to static settlements of adjacent buildings of differential and dynamic design will accommodate these. loads and how the
RESPONSE
progressed Seismic analysis of sufficiently to the station structure has net dynamic loads on the foundation. evaluate the r..agnitude of the these Detailed submitted in a later amendment. information will be -l discussed in section II-5.7, Amendment 4, Static foundation loads are SNPS PSAR.
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SNPS-1 COMMENT.12.73.
Will the bottom of the located below ground containment structure be water level?
. water proofing or other protection willIf so, be what used between the the ground and the containment?
Evaluate cracking liner. and of ground water thereby re This should include the potential on liner stability and liner corrosion. effect RESPONSE: The V-2.3, Amendment 4, SNPS PSAR. Response to this Comment will be f l
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- i. 1 SNPS-1 COMMEter 12.8: Please supply the proposed . design following information on the k of the primary conta'nment including the floor dividing 1
structure, drywell and the suppression chamber: the 1
o
- a. Scaled load plots for moment, shear, .I deflection, longitudinal force, and hoop tension for.
loadings: dead load, pressurethe following j loads, design earthquake, wind, and liner 1
{
and accident).
and concrete thermal loads (normal (
- b. The normal operating and ' transient accident thermal gradients through i the drywell.
- c. A description of how torsional (e. g. , l seismic) the design.loads will be considered in
{
d.
A list and justification of the values of modulus of elasticity (Ec I "Dd Poisson's ratio (p ) for the cracked and uncracked reinforced concrete-structure at different elevations, and an explanation of their use in the design of the primary containment )-
structure, including consideration of thermally induced liner / concrete interactions and the effects of concrete shrinkage.
RESPONSE: a.
containment designTheisinformation
. complete. will be not be available until the amendment. It will be submitted in a later l
- b. Thermal gradients be sutanitted with a later Amendment.have not as yet been prepared. They will The Response Amendment to Comments 4, SNPS PSAR. e and d will be found in Section V- 2 . 3 ,
AS-172
SNPS-1
. COMMENT 12.9: Describe the analytical procedures used to arrive
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L ' 'at the loading forces, moments, and - shears on the base slab,- considering non-axisymmetric loading and deformations of the mat. To what extent have transient thermal gradients been considered?
RESPONSE: The Response to this Comment will be found in Section V-2.3, Amendment 4, SNPS PSAR.
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I AS-173
SNPS-1 COMMENT 12.10:
With respect to seismic design of the pritaary secondary and.
containment structures, please.
' describe:
- a. 'the manner in which damping will be considered- in the structural design.
Indicate the damping values to be employed for rocking motion and horizontal translation.
- b. The limitations on deformations that will .be employed. in the design for the design basis (larger) earthquake and the operating- basis- (smaller) earthquake.
RESPONSE: The Response to Comments 12.10 a and b will be found in Section V-2. 3, Amendment 4, SNPS PSAR.
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I AS-174
SNPS-1 COMMENT 12.11: With respect .to the primary containment liner design, describe:
- a. . The types and combinations of loadings considered with regard to liner buckling, and the safety factors ' to be used.
-b. The geometrical pattern, type and spacing of liner . anchors; and, the analysis procedures, boundary condi-tions, and results with . respect to buckling of the liner under the loads cited above.
- c. The possibility of elastic and inelastic buckling. Provide sample calculations showing the influence of all pertinent parameters, such as:
Variation of plate thickness due to commercial tolerances; variations in the yield point of the liner steel; Erection inaccuracies (local bulges, offsets at seams, wrong anchor location, etc.);
shrinkage of concrete; variation of Young's modulus and Poisson's ratio for cracked and uncracked concrete, and as a f unction of stress level in concrete (elastic and plastic);
Potential loadings due to ground water .l infiltration, earthquakes, '
temperature gradients, and subatmospheric pressures witnin the '
containment.
- d. The stress and strain limits used for the liner, the bases for these limits, and the extent to which these limits relate to liner leakage, and liner cracking.
- e. The type, character, and magnitude of cyclic loads for which the containment liner will be designed,
) including pressure / thermal '
load AS-175
_ _ _ _ - - - - - - - .--__-.-_----J
SNPS-1 variations and the number of and basis for the cycles generated by earthquakes,
- f. The manner in which the t emperature rise and expansion of the liner under DBA conditions and the resulting effect on the concrete is considered in the design analyses.
- g. How variations in thickness and yield point of the liner due to commercial tolerances are taken into consider-ation in the design of the concrete backing the liner.
- h. The extent to which protective coatings and/or insulation will be applied to the liner.
- i. The design approach that will be used where loadings must be transferred through the liner such as at . brackets or equipment mounts. Provide typical design details and computations.
- j. How the shears, especially those due to thermal expansion and earthquake, will be accommodated by the base slab liner, including the arangement for transfer of shear loads through the .
liner at the bottom of structures and I equipment supported on the liner.
- k. An analysis of the capability of the liner / seal arrangement at the base- ,
to-cylinder juncture, to absorb the strains it will experience under l design basis accident and earthquake i 4
conditions. Include code references '
to which this, and other special liner areas, may be designed. l Discuss the potential effects of local concrete cracking in this area on liner anchors.
- 1. The procedures for analysis of liner stresses around openings. Frovide the met hod of liner design to accommodate these stresses and the l
related stress limits. Justify the i
proposed thickening of the liner at I
penetrations. Discuss the liner anchors at this location.
AS-17e l
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I SNPS-1 RESPONSE: -The Response to Comments a through g will be found in
- Section V-2.3, Amendment 4, SNPS PSAR.
- h. The Response to this Comment will be found.in Section V- 2. 5, Amendment 4,'SNPS PSAR.
The Response to Comments i through I will be found in Section V-
- 2. 3, Amendment 4, SNPS PSAR.
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l SNPS-1 COMMENT 12.12: With \
describe: respect to the -liner l anchorage desigt.$
I a.
The liner anchorage system to be used including sizes of anchors, and weld sizes. spacing {
- b. The j analytical techniques to be procedures ani i
used in line: ,"
- anchorage design, calculations. including sample ,
c.
How the elastic liner and' inelastic buckling of !{
design of the will be considered in the !
possibility- of anchors. Discuss the unbalanced acting on one provide or several anchorsloadsq a study showing that and no sequential failure of the massive buckling of the liner,anchors o,i RESPONSE: The mass failure of anchors could occur.
Section V-2.3, Amendment 4, Response to Comments a, b, and c SNPS PSAR. ound in 5 I
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AS- 178
SNPS l COMMENT 12.13:. With regard to primary containment penetration I 1
design, please describe !
- a. The design . criteria that will be applied to ensure that, under postulated design basis accident conditions, potential torsional,
(
L axial, bending, and/or shear loads that may ' be applied to the.
penetration by piping will not cause a breach of the containment. Include
.the design criteria intended to prevent pipe- rupture between the i penetration and containment isolation -
valves. What design . codes will be j
.1 applied in the design?
- b. The extent to which the penetrations and the applicable surrounding liner regions may be subjected to vibratory loadings from egoipment attached to :
the piping systems. Indicate how these loads will be l treated in design. {
- c. criteria for concrete' thermal i protection at . penetrations, including '
those for the temperature rise to be permitted in the concrete 'under j operating conditions and the time j dependent effect that' loss of thermal {
protection (both coolers) would have
- on the containment's structural and leaktightness characteristics. Indi-cate the thermal gradients that will i be used for design purposes. (see also item 5.2)
- d. The capability of the penetration designs to absorb liner strains without severe distress at the openings.
- e. For all penetrations, the criteria that will be used for the bending of reinforcing bars which have to clear the opening including maximum slopes and minimum bending radii to avoid. l local crushing of concrete.
- f. How normal, shear, bending, and torsional stresses in concrete will l be considered in the design of penetrations !
between approximately l AS-179 1
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SNPS-1 nine ' inches and four feet in i diameter.
- g. The length required to anchor the bars in cracked concrete, and the use of ACI Code 318 or any other code for concrete under biaxial tension, and cracked in two directions. Will any special provisions be made to anchor bars at openings?
RESPONSE:. a. The Response to this Comment will be found in Section V-2.3, Amendment 4, SNPS PSAR. i
- b. The Response to this Comment will be found in Section V-2.3, j Amendment 4, SNPS PSAR. !
- c. The Response to this Comment will be found in Section V-2.3, Amendment 4, SNPS PSAR.
- d. The - Response to this Comment will be found in Section V-2.3, Amendment 4, SNPS PSAR.
- e. Circumferential ano meridional reinforcing will not be bent to clear openings. Reinforcing will be placed as descriLea in i Section V-2.3.3.8.1.
- f. The Response to this Cotament will be found in Section V-2.3, !
Amendment 4, SNPS PSAR.
- g. The Response to this Comment will be found in Section V-2.3, Amendment 4, SNPS PSAR. !
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l AS-180
SNPS-1
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COMMErr 12.14: For;the larce primary containment penetrations, please describe a.- The primary, secondary, and thermal loads that will be considered in the design of the openings.
- b. The stress analysis procedure that will be used in design.
- c. The method that will be followed for the design (working stress design method,. ultimate strength design ,
method, or both). If ultimate I strength is used,- the f actored load combinations -should be given together with the corresponding capacity reduction factors.
- d. How the existence of biarial tension .,
in concrete will be. treated in the l design and how the normal and shear j stresses due to axial load, two- l directional bending, two-directional j shear, and torsion will be combined. l I
Also, state the proposed criteria, design methods, and materials for'the design of the thickened part of the-concrete wall (ring girder) around the large openings.
- e. The method to be used in designing the l thickened part of the concrete wall j structure, around large openings, l regarding the compatibility of )
deformations with the adjacent J portions of the structure. Include j the manner of considering temperature I gradients and shrinkage. How will torsional stresses be considered to be taken by cracked concrete?
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- f. Additional information on reinforcing i patterns that will be used around large openings (i.e., rebar size and spacing) .
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- g. Sah.ple computations, listing all the criteria and analyzing the effect of all pertinent factors such as cracking, etc. What safety factor will be provided in the design of large openings?
AS-181 f
SNPS-1 RESPONSE: 'a. The Response Section V-2.3, Amendment 4, SNPS to PSAR.
this Comment will be found in b.
The Response Amendment to this 4, SNPS PSAR. Comment will be found in.Settion V- 2. 3,
- c. The AmendmentResponse to this . comment will be found in :Jection V-2.3, 4, SNPS PSAR.
d.
The Response Amendment to this 4, SNPS PSAR. Comment will be found in Eection V- 2. 3,
- e. The AmendmentResponse to this 4, SNPS PSAR. Comment will be found in Section V-2.3, f.
will be submitted in a later Amendment. Drawings of the reinforcing
). !
Design has information not progressed at this time.
sufficiently to preseIt -this Amendment. It will be submitted in a later '
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l AS-182
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(. . COMMENT 12.;5: I Show on method, by ~what what statistical basis and frequencyand Cadweld splices and butt weld splices will be tested. ,
l specified for acceptance or rejection. Include standards t{
RESPONSE: The
! V-2.5, Amendment 4, SNPS PSAR. Response to this Comment will
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I' AS-183
CNPS-1 l
l COMMENT'12.16 Indicate the extent of user verification testing (
of certified liner innterial properties, such as .
NDTT, plate thickness, ductility, weldability, etc., and and ductility.certified reinforcing rod yield point
RESPONSE
The Response to this omanent will be found in Section V-2.3, Amendment 4, SNPS PSAR.
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AS-184
- - - _ _ _ _ _ _ _ _ _ _ _ _ - A
7.
SNPS-1 COMMENT 12.17: Indicate'which ASME or API code sections will be i .. applicable to the liner and donne ' erection tolerances. .
RESPONSE: The Response to .this Comment will be found in Section V-2.5, Amendment 4, SNPS PSAR.
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AS-185
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l SNPS-1 l I' j COMMENT 12.18: Describe the procedures and quality measures tor field welding of control {
the ASTM A 537 Grade B heat treated liner plate. i qualifications, welding proceduresInclude welder l j
, post-weld heat treatment, methods of inspection, and 1 construction records to be kept. Indicate the procedures j
and criteria ' that will be placed on seam welds to control weld porosity and to assure i adequate ductility. Describe the seam weld !
i radiography program; also, provide an evaluation of this proposed program to provide assurance l that flaws capable of developing into leakage paths under design basis accident conditions will not, in fact, exist. Describe the procedures and quality control measures for welding liner studs and leak test channels.
RbSPONSE: The V-2.5, AmendmentResponse to this Comment will be found in Section 4, SNPS PSAR.
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0 AS-186
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SNPS-1 l
COMMENT 12.19: Describe the inr,Mumentation program l
, for i structural testing of the primary containment structure, including:
- a. Identification of concrete and liner areas to be instrumented.
- b. Purpose, type, expected accuracy, and redundancy of instrumentation.
- c. The range of strains and deformations-expected.
- d. The protective measures that will be taken to insure instrument performance during structural' testing, considering the interval between instrument installation and its use.
RESPONSE: The Response to this Comment .will be found in Section V-2.5, Amendment 4, SNPS PSAR.
AS-187 o
SNPS-1
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COMMENT 12.20:
How will the floor, which divides the drywell and suppression to capability chamber, be tested to demonstrate its pressure for which it is designed? withstand the 2
RESPONSE
V-2.5, Amendment 4, SNPS PSAR.The Responee' toSection this Com i
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+ I AS-188
SNPS-1 F. .
COMMENT 12.21: Provide
\
stresses e for table thr.t compares. the the 115 percent computed and 110 percent stresses accident due to the loss-of-coolan alone, e
basis accident conditions.and to the earthquake plus i
Include the following:
- a. Stresses in the concrete reinforcing steel, including and thermal stresses openings;, especially at large temperature methods of combining stresses ~ duegradients:
normal, to tangential, bending, torsional loads assumptions and ;
cracking; on stresses in stirrups; etc.
k b.
Influence of shrinkage.
c.
Influence deformations of liner elastic and plastic 3
d.
Liner stresses before cracking concrete occurs of and af ter cracking.
- e. Influence of gradients. transient thermal \
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RESPONSE
information Design at thishas not progressed sufficiently to time. pr k' a later Amendment. esent.this The information will be submitted in i
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i A5-189
l SNPS-1 COMMENT 12.22: Provide an analysis of crack size, spacing, and pattern expected during containment structural
. testing.
RESPONSE: The Response to this Comment will be found in Section V-2.5, Amendment 4, SNPS PSAR.
AS-190
SNPS-1 COMMENT 12.23: Describe the surveillance capabilities provided by
, the design with reference to both periodic inspection of the primary containment liner measurement of deformations during periodic structural and/or leakage testing of the- primary containment.
RESPONSE: The Response to this Comment will be found in Section V-2.5, Amendment 4, SNPS PSAR. .
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! AS-191
t SNPS-1 ColeGNr 12.24: What are the loads on the soil at the base of the containment structure due to dynamic stresses frcan the design basis (larger) earthquake?
Discuss the potential soil-structure interaction under these conditions.
RESPONSE: Seismic analysis has not pro sufficiently to present this information at this time.gressed The information will be subsitted in a later Amendment.
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A5-192 ;
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COMMEdT 12.25: l
( Will the soil under all class ' I structures be' 4' excavated to Ele M ion-12 -and . backfilled?
the backfillwill procedures will be subjected, and whatSpe compaetior.
be used to control measure- tLa RESPONSE: The Response 1 XII-2.1, Amendment 4, luud PSAR. this Comment will be found in Section !
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AS-193
______._a.----" - - - - - " - - ' ' ' ^ ___ . _ _ _ . _ _ _ - - - - - - - - - - - - - - - - - -_ _ _ _ _ _ _ --- b
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. I SNPS-1 CO39tErff 12.26: What is the basis for the limitation of 15 structural and/or leakage tests of the primary containment ,
at design pressure (pg. V-2-2, 9/26/68 Rev.)?
RESPONSE: The correction to the statement referenced in this Comment will be found in Section V-2.1, Am mdment 4, SNPS PSAR.
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A5-194
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i SNPS-1 COMMENT 12.27:
. In2.3.3.4 describing the design of the liner in Section' of. the PSAR it {
' Loads induced in the liner isresulting noted that Eg, the {
distortions of the from the i i- concret.e hypothetical earthquake" will structure due to I the design loads. be inclMed with on the procedure Furtherto be information is desired used te compute Eg, particularly such as joints at points of strain concentration or cracks concrete cracking is assumed to inoccur).
the concrete (if
RESPONSE
V-2.3, Amendment 4, SNPS PST.R.The Response to thisSection Comment wil 1 i
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AS-195
_ ___ -- - ----- --- - ---- ---- - ~
SNPS-1 coMMEttr 12.28: The statement is made in the PSAR containment construction that "the proposed for station will be adequate to accommodate all this l operating design basis accident, and earthquake i
loads assigned to it without using diagonal reinforcing steel Further elaboration is to resist these loads."
reque sted regarding the manner or mechanism by which the above loads will be transmitted to the foundation.
regard, diagonal steel is In this shown at the joint between the mat; what containment criteria will be cone used and to the foundation determine cut-off the in Figure V-2-7 (9/26/68 Rev.) point for this steel, which is indicated base of wall only". of the PSAR cs *at RESPOMSE:
V-2.3, Amendment 4, SNPS PSAR.The Response to this Comment Section will be found in e
A5-196
SNPS-1 COMMENT 12.29: On page V-2-20 and following there is a discussion of the load combinations that are to be employed in the design of the containment structure. For applicable stresses it is noted that the load capacity of th(e te.nsion members will be based on the guaranteed minimum yield strength. Please clarify how closely you will follow ACI code 318 for loadings other than tension, i.e. flexure, shear, etc.
RESPONSE: The Response to this Comment will be found in Section V-2.3, Amendment 4, SNPS PSAR.
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i AS-197
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SNPS-1 COMMENT 12.30: on page XII-2-12 of the PSAR it is noted that the vertical shock spectrus will be applied simultaneously with the horizontal forces.
However, no statement is given to indicate that the stresses' arising from vertical and horizontal earthquake loadings will be added directly and linearly, as appropriate, to the stresses arising from operating . conditions, dead load, etc.
Clarification of this. point is requested.
RESPONSE: The Response to this Comment will be found in Section XII-2.1, Amendment 4, SNPS PSAR.
AS-198
SNPS-1
.I. ' COMMENT 12.31: The method of. seismic analysis presented in Appendix are A is not completely described. Details requested concerning the method combination of. modal ~ responses to obtain .of design values, and the. specific methods of handling dam foundation rotation. and tipping, ping and presentation of . typical results of modal deflections, shears, ofand calculations important parts of the plant. moments of RESPONSE: The method of damping, and foundation combining. modal responses, handling rotation and tipping will be found in Section V-2.3, Amendment 4, SNPS PSAR.. The analysis has not progressed sufficiently . to present typical results of modal deflections, shears, and moments at structures. various parts of the-Amendment.
The. information will be submitted in a later I
I e i AS-199
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SNPS-1 COMMENT 12.32: Describe-the methods and design criteria that will' be used for the seismic design of the reactor vessel .. support skirt and its supporting concrete base, the reactor shield wall, the reactor top stabilizing ring, the recirculation pump 4 supports, and the down comer pipe supports. l t
RESPONSE: The seismic design criteria for the SNPS-1 'will be found -in Section XII-2, Amendment 4, SN'S PSAR, and Appendix D.
The general method of dynamic analysis of the reactor vessel
{ concreto support structure, reactor shield wall, stabilizing
) ring, recirculation pump supports, and downcomer pipe supports is i
described in Section V-2.3, Amendment 4, SNPS PSAR.
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SNPS ' COMMENT 12.33:
( ) Describe analysis the of methods class that will-be used for seismic components in general.- I mechanical equipment and analytical models Describe specifically the emergency reactor to cooling be used for the normal and piping systems, including- location ~
conditions. What are the values of lumped masses and support and the bases for the damping dynamic analysisfactors which will be used in the for mechanical equipment and components and their supports or mountings?
RESPONSE
The Response to this cossnent will be Amendment 4, SNPS PSAR. Amendment, Response 12.32, and Se found in. this l
AS- 201
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1 COMMENT 12.34: With reference to your proposed design criteria j for protection against tornadoes - (II-11-3) please '
define "safa plant shutdown" as applied to. these criteria and provide a listing of the structures and equipment whose integrity must be preserved to meet this definition of safe plant shutdown.
Please confirm our understanding that the same -
tornado design conditione indicated (pg._ V-3-4) for the secondary. containment will'be used for all applicable structures and equipment.
Relative to these design conditions, what-atmospheric pressure drop-and rate of drop is the plant designed for? What is the basis for considering that a' 300 mph wind. acts on a )
component (pg. V-3-4), when your design criteria . {
call for 300 mph horizontal peripheral tangential i velocity and a ' 60 mph transnational velocity, 'j which, when combined would result in' a.-360 aph wind speed relative to the groad? What stresses and/or strains will be allowed in structures and equipment essential for safe shutdown under those loading combinations which include tornadic Joads? At what tornado wind velocity higher than the design value indicated (300 mph) do you estimate that the plant could still be safely i shut down?
RESPONSE: " Station Safe and Orderly Shutdown" is' defined as the transition of the reactor core from normal, steady-state, full power operation (2,550 Nwt, 1,000 psi gage, 575 F) to the cold, saberitical, depressurized quiescent state in manner consistent with the station principal design criteria.
Refer to SNPS PSAR Amendment 4, Section XII-2.1 for a list of structures, systems or components involved in the station safe and orderly shutdown operation.
Fbr Response to Cos:ments on pressure drop and rate, wind speed design criteria, and allowable stresses and/or strains, please refer to SNPS PSAR Amendment 4, Sections V-3.2 and XII-2.1.
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A5-202
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. SNPS-1 COMMENT 12.35: We are concerned that spent fuel elements in the {
} storage pool could be daniaged by tornadoes and l release significant radioactivity offsite unless !
design provisions are made to protect them. '
Please indicate your design bases and describe those design aspects included for protection of !
spent fuel from tornadoes. In particular, consider protection of the fuel elements from debris which may be dropped onto the elements, and measures which could be taken to prevent fuel overheating if the water is lost as a consequence of missile damage.
RESPONSE: The Response to this Comment will be found in Section V-3.2, Amendment 4, SNPS PSAR.
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A5-203
SNPS-1
' COMMENT 12.36: Since the load factor for tornadic conditicns is 1.0, will' the load reduction factor used for '
concrete be 0.75 rather than 0.85? If not, what ' '
is the basis for using the higher value?
RESPONSES. The Response to this Comment will be found in Section III-2.1, Amendment 4, SNPS PSAR.
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' ' COMMENT 12.37: Please . provide a loadings .will discussion of how torna&
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'* be translated . into' direct, torsional and shear loadings on structures. This would. include
-tornado. funnel, theitsinfluence of the size. of the effect of transnational potential -non-uniform speed and the pressure distributions. horizontal-RESPollSE: The V-3.2,' Amendment ' 4, SNPS PSAR. Response to' this Comment will be found in S e
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AS-205
L SNPS-1 13.0- STATION OPERATIONS COMMENT 13.12 Since both RHRS heat' exchangers are apparently required percent of in order for the ECCS to achieve 100 !
desigu. cooling capability. -what.
limitations reactor do you propose to invoke on continued exchangers becomes inoperable?-operation in the event on
RESPONSE
Sections and Please refer to.the following SNPS PSAR Amendment 4, Figs. in ' order 'i Exchanger design basis, detailedto review' the ECCS-RHRS Beat analysis, and test and inspection requirements: description, performance a) Section I ~'.'.
- Reactor Core Cooling - Principal Design Criteria b)
Section I-2.3 - Station containment'- Principal Design l: c) Criteria Section VI-2.0 I-3.4 - Station Auxiliary Systems d) Section -
e) Section V-2.4 ECCS f)
Section'WIII-4.0 - Engineered Safeguards Analysis g) Section X-2.5- - Standby Diesel-generator Systems
- Reactor RHRS Shutdown Cooling Sub-System -
h) Section XIV-3.4
- Loss-of-coolant Drywell) accident (Inside the i)
Section Appendix G - Criteria-Conformance 38, 41, 42, 44, 45, 49, 52, 58, 59, 60, and 61.
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Figures VI-2-6 and VI-2-7A through C. i The Amer.3 ment, Response 6.4,above referenced documentation and furth heat exchangers is necessary establishes trat only pLne of the two for any operation. safety-related station use of bothAdditional economic benefits can be derived by- the the - heat exchanger loops at the same time e.g.,
reactor hr or shutdown cooldown rate for refueling operation may be 20 30 hr depending water temperature, etc. ,on orthe theheat exchangers length of time available, service permitted on the reactor is hot standby prior to returning to power operation i or shutdown.
In regardof details to operation with a defective RHRS-heat exchanger, the
" Limiting Specification-Opera ting condition of Operation" are " Technical-Permit-FSAR" concern items and details and conditions of equipment the documented at that time. availability will be satisfy the PSAR-stage The referencescited above describe and requirements. " demonstration of capability" (
AS-206
t SNPS-1 COMMENT 13.2:
'i Please provide the following information relative to emergency planning for the facility:
a.
Duties, responsibility and authorities of all individuals and organizations expected to be utilized emergencies. in b.
Extent to which organizations - these individuals and have been consulted relative to their- capability and willingness an emergency.
to assist in the event of
- c. Preliminary outline of protective measures and actions to be incorporated in the emergency plans, especially in regard to the public beach on the site, the nearby summer camp, and the government installation and residence within the site.
RESPONSE: The XIII-4.4, Amendment 4, of SNPS PSAR. Response to this Comment will be i
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AS-207
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COMMENT 13.3: What qualifications will you require for the personnel who will fill the positions of reactor ,
engineer, instrument and controls engineer, and l the radiation protection and chemical engineer indicated on Figure XII-2-1? Fig. XII-2-1?
RESPONSE: The Response to this comment will be found in Section XIII-3.1, Amendment 4, of SNPS PSAR.
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l A5-208
SNPS-1
, COMMENT 15 .. Relative to the proposed nuclear training program
, outline (pg. XIII-3-2 and Figure XIII-3-1),
please provides
- a. Additional information on the " basic nuclear course";
- b. The extent of radiation safety training to. be given to operating staff other than technicians;
- c. The schedule for the training program and the relationship to the plant construction schedule.-
RESPONSE: The response to this comment will be found in Section XIII-3.2 and Fig. XIII-3-1, Amendment 4, of SNPS PSAR. .
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AS-209
SNPS-1 COMMENT 13.5: Provide a summary of the proposed pre-operational
. test program which indicates the scope of the ,
testing to be performed for each system which is related to safety, including the objective of the testing, the design criteria which will be verified, and the acceptance criteria.
I RESPONSE: The operational test procedures and plans for this facility will be presented in detail- with the Final Safety !
Analysis Report which will be submitted in support of the request for an operating license. However, in order to facilitate future references in this area of concern, the following information is provided.
There are three stages of testing that are performed during ,
station design and construction, which are provided to assure '
that the systems and equipment installed in the station will ,
perform as intended by design. The three stages of testing can !
be generally classified as follows:
- 1. Functional Tests
- 2. Preoperational Tests
- 3. Startup Tests
- 1. Functional Tents Functional tests are those tests performed on systema or equipment whose performance has not been previously demonstrated in a- manner such that the design has been completely verified l either functionally or phenomenologically. For this station, the ,
necessary functional testing on all safety systems has been I demonstrated and their performance has been or will be reported.
As typical examples, the following are cited:
- a. Reactor Core Spray Cooling Topical Report GE APED j (spray cooling) 5458, ' Effectiveness '
of Core Standby Cooling System for General Electric Boiling Water Reactors" -
March 1968 '
- b. Reactor Core Residual Heat (Same as 1)
Removal System (Flood cooling)
- c. Control Rod Velocity Limiter Topical Report GE APED 5446, " Control Rod "elocity Limiter" -
Aarch 1967
- d. In-Core Nuclear Instrumentation Topical Report GE APED 5706, "In-Core Neutron .
Monitoring for General A5-210
_.-____________._.--______A
SNPS i:
t Electric Boiling Water
, Reactors" - November 1968
- e. Main Steam Line~ Isolation Valves GE APED Topical Report to be issued April, 1969
- 2. Preoperational Tests Near the time of completion of the physical construction of the
, station, the assembly of equipment, and completion of construction testing, the preoperational test program begins.
This period is designated as Phase I of the test program during j which detailed testing of components, subsystems, sytems and combine.1 systems is performed.
The purpose of the preoperational test program is three-fold:
confirm that construction is complete to the extent that' the.
equipment and systems can be put into use during completion of other construction; adjust and calibrate the equipment to. the extent possible in the "oold" station; assure that .all process wid safety equipment is operational, and in compliance with license requirements,- to- the extent possible and necessary to proceed into initial fuel loading and the startup program.
An extensive preoperational test program is planr.ed for the reactor ' unit being started. This program will start approximately six months before initial fuel loading. The actual duration of each test is short relative to the entire six month programs the longest test, the control rod drive system checkout, takes approximately one month of testing.
Key systems are sequenced for completion and testing early enough to provide auxiliary services for testing and operation of other systems or for construction activities; e.g. , the use of the makeup system for chemical cleaning. This results in an early requirement for electrical systems, demineralized water makeup, and cooling water systems.
A2ter nuclear fuel is loaded in the reactor, all interconnected auxiliary systems are treated as potentially radioactive.. In time, many of these systems are treated as potentially radioactive to impose restrictions and time limitations on maintenance work.
During normal station operations, some components and systems cannot be observed for proper performance. To avoid these limitations during the preoperational tests, all of the nuclear steam supply system and its auxiliary systems are normally tested before fuel loading.
(- As an example of the documentation just cited, please refer to
!. Oyster Creek Nuclear Power Plant Unit No. 1, Amendment 11,
', g Comment IX-3, (AEC Docket No. 50-218). This document contains a l A5-211 1
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SNPS-1 detailed outline and tabulation of typical preoperational tests which were recommended for use.at the operating stage.
- 3. Startup Tests The startup-test period begins with the loading _ of . fuel- and continues through the completion of the warranty.run. During this period, the stati<.n is taken to its designed, operating , . }
full power condition ' in a s'afe; controlled,- gradual fashion.
l Extensive testing is performed at selected operating conditions ;
to dennonstrate safe, efficient performance of station components.
Three documents will. be supplied by GE-APED at the operating license time specifically for implementation- of the startup testing of GE-APED supplied equipment. Similar documents will be j
. prepared for safety systems supplied by S&W. '
a)' The_ Startup Test Specification will define the minimum test I program for' safe and efficient startup. The Specification will l authorize and require the performance of the described tests. It- f will be. used to limit and . define the freedom for change of startup test activities. They will be reviewed and approved by l l the responsib3e design engineering personnel. Every test must be !
performed to the extent specified. '
b) The Startup Test Instructions will contain the results of i analyses made to facilitate the startup testing activities i required by the Startup Test Specification. The Startup Test (
Instructions usually include (1) methods, data, and calculational l aids for use in the main control room to determine core _I performance parameters- restricted by license or warranty. The parameters determined are, for example, fuel assembly flow, power, quality, MChFR, etc., (2) Reactor Period vs.
Reactivity Table, (3) Calculated core power distributions and control rod patterns, (4) control rod worths for use in determining shutdown reactivity margins, (5) methods and data for calibrating in-core nuclear instruments, and (6) other information useful for startup test activities and not already published. I c) The Startup Test Procedures will present the recommended test i methods and describe the steps for performing the tests required I by the Startup Test Specification. Procedures will be reviewed by LILCO prior to approval by the Station Superintendent as 3 described in Section XIII-4, Amendment 4, of SNPS PSAR. The {
Startup Test Procedures will also contain criteria for judging I the test results, where applicable, and planned data and l
l calculation sheets which provide for site analysis of the data.
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SNPS-1 COMMENT 13.6:
)-
What provisions are you making to insure operatc!
communication j telephone lineswith are offsite points in the ever operation, an lost during either norma natural accident disaster such as a or in the event of earthquake? severe storm or a: .
RESPONSE: The XIII-4.4, Amendment 4, of SNPS PSARresponse to this comm I
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AS-213 1 l
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SNPS-1 14.0 ACCIDENT ANALYSIS COMMENT 14.1:
During refueling, the normal ventilation system is
.in operation and radiation monitors located near the refueling area will detect- excessive activity from an accident. A signal from these monitors will close the ventilation dampers, start .the mixing fans and initiate operation of the standby gas treatment system.
required for this What is the estimated time sequence of events to take place ? . By comparing this time sequence with a conservatively short estimate required for fission products to of the time to the be transported environment following a fuel failure, estimate the fraction of fission products released . prior to isolation of the normal ventilation system. Provide _all necessary information to justify these estimates, including ventilation system flow rates, description of the air mixing equipment, and its effectiveness, location of monitors, etc.
RESPONSE
The res~ponse to this comment will be found in section V-3.3 of Amendment 4, SNPS PSAR.
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e AS-214 J
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TURBINE h CONDENSER'~-
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gC V AIR EJECTOR-s STATION
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' h CONDENSATE WATER DEMNERALIZER VENTILAT
-- 5 h 21-:-: .
EXHAUST DUCT O [n = ~ < = yp--
HEATER DRAIN DEMNERALIZER CLEANUP FILTER-DEMNERAll2ER
/ 5 0 /~7 af = ~
/
N/ '"
y Or
/ 6" O(1,7 g 6-O i
Process WATER A B-E-F-GA (EXTERNAL LOOP) = 3 sec A B 3 GA(INTERNAL LOOP) = 24 sec STEAM: A B C D = 8 sec
, gas: Vy: MINIMUM - @ Hrs Y-Z- ABOUT 3 sec.
Detection 01: Main Steam Line Monitors (4) '
A-B-C-C'= 6 sec D2: Air Ejector Off-Gas fionitors (2)
V-V'= 100 to 1000 sec
D3: Ventilation Exhaust Duct GasMonitors(2) FIG. l.3d-l Z-Z' 100 see RADI0 ACTIVE WASTE CONTROL SYSTEMS -
REACTOR GAS AND LIQUID PATHS - SCHEMATIC
- l. a f-10 s
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