ML20235F083
| ML20235F083 | |
| Person / Time | |
|---|---|
| Site: | 05000000, 05000447 |
| Issue date: | 10/11/1974 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML20234E460 | List:
|
| References | |
| FOIA-87-40 NUDOCS 8707130205 | |
| Download: ML20235F083 (388) | |
Text
_ .__ .--_ - _ _ _ - _ _ - _ . _ _ _ . _ _ _ - - _ .
I October 11, 1974 IUIi f. E 0Ylii SAFETY EVALUATION i
0F THE GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT d
DOCKET NO. SIN 50-447 l
, UNITED STATES ATOMIC ENERGY COMMISSION DIRECTORATE OF LICENSING WASHINGTON, D.C.
ezw x.
3 s=
(
m l
8707130205 G70623 PDR FOIA THOMAS 87-40 PDR
T BLE OF CONTENTS 1.0 Introduction and General Description of the Plant 1-1 1.1 Introduction 1-1 1.2 General Plant Description 1-7 1.3 Shared Systems 1-12 1.4 Comparision With Similar Facilities 1-12 1.5 Identification of Agents and Contractors 1-15 1.6 Summary'of Principal Review Matters 1-16 1.7 Facility Modifications as a Result of Regulatory Staff Review 1-20 1.8 Requirements for Future Technical Information 1-21 2.0 Site Characteristics 2-1 2.1 Geography and Demography 2-1 2.2 Nearby Industrial, Transportation and Military Facilities 2-1 2.3 Meteorology 2-1 2.4 Hydrology 2-5 2.5 Geology and Hydrology 2-6 3.0 Design of Strdctures, Systems, and Components 3-1 1 3.1 Conformance with AEC General Design Criteria 3-1 3.2 Classification of Structures, Systems, and Components 3-1 3.2.1 Seismic Classification 3-1 3.2.2 System Quality Group Classification 3-4 3.3 Wind and Tornado Loadings 3-6 3.4 Water Level (Flood) Design 3-9 3.5 Misc 11e Protection Criteria 3-10 3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping 3-13 3.7 Seismic Design 3-17 3.8 Design of Seism'ic Category I Structures 3-21 3.9 Mechanical Systems and Components 3-29 3.10 Seismic Qualification of Category I Instrumentation and Electrical Equipment . 3-36 4.0 Reactor 4-1 4.1 General 4-1 4.2 Mechanical Design 4-3 4.3 Nuclear Design 4-11 4.4 Thermal and Hydraulic Design 4-30 5.0 Reactor Coolant Systems 5-1 5.1 Summary Description 5-1 5.2 Integrity of Reactor Coolant Pressure Boundary 5-1 j 5.3 Thermal Hydraulic Systems Design 5-14 3 5.4 Component and Subsystem Design 5-15 1 i
l l
l l 6.0 Engineered Safety Features 6-1 6.1 General 6-1 6.2 Containment Systems 6-2 1
6.3 Emergency Core Cooling System (ECCS) 6-44 6.4 Habitability Systems 6-57 6.5 ECCS Pump Suction Strainers 6-59 7.0 Instrumentation And Controls 7-1 i
l 8'. 0 Electric Power 8-1 l
9.0 Auxiliary Systems 9-1 9.1 Fuel Storage and Handling 9-2 9.2 Water Systems 9-8 9.3 Process Auxiliaries 9-8 >
9.4 Air Conditioning, Heating, Cooling and Ventilating Systems 9-10 9.5 Fire Protection 9-18 10.0 Power Conversion System 10-1 11.0 Radioactive Waste Management 11-1 11.1 Summary Description 11-1 11.2 Liquid Waste Treatment System 11-2 11.3 Gaseous Waste Treatment System 11-3 11.4 Solid Waste Management Systems 11-20 11.5 Process and Effluent Radiological Monitoring Systems 11-23 12.0 Radiation Protection 12-1 12.1 Shielding 12-1 12.2 Ventilation 12-5 12.3 Health Physics 12-6
L
! o i
I 15.0 ' Accident Analysis 15-1
.5.1 Abnormal' Operational Transients 1 15-2 15.2 Design Basis. Accidents 15-5 15.3 Loss-of-Coolant Accident (Radiological Consequences) 15 ' 15.4 Fuel llandling Accident 15-9 15.5 Control Rod Drop 15-9.
I 17.0 Quality Assurance 17-1
" 17.1 General' 17-1 17.2 Organization 17-1 17.3 Quality Assurance Program 17-4 17.4 Conclusion- 17 ~('
l Appendix A Chronology I
Appendix B Effluent. Treatment Systems Branch Position No. 1 Appendix C Hendrie to Hinds Letter Dated April 19, 1974 .j 1
Appendix D 1hchnical Report on The GE 8x8 Fuel Assembly (2/5/74)
Appendix E Review and Evaluation of GETAB for BWR's, dated September, 1974 l
l l
i i
i
}
~.
s p -
1 l- ABBREVIATIONS-l a-c '
. Alternating current ACI American Concrete Institute-ACRS Advisory Committee on Reacto'r Safeguards
~
ADS Automatic Depressurizat'on i System
.AEC United States Atomic Energy Commission ANS. American Nuclear Society .
ANSI American National Standards Institute
,ASCE a ~American Society of. Civil Engineers
~ i-ASME American Society.of Mechanical Engineers l
ASTM American Society for Testing and Materials +
k Btu /hr British thermal units per hour BWR Boiling Water Reactor i
{
cal /gm Calories per gram l cfs Cubic feet per second CHF Critical Heat Flux C1/yr Curies per year l
CP Construction Fermit DBA' Design Basis Accident i
d-c Direct current i 1
DC. Diesel Generator AT Differential Temperature DOT Department of Transportation
tP .
i i
ABBREVIATIONS (Cont'd)
ECCS Emergency Core Cooling System ESF Engineered Safety Feature
- O F Degrees Fahrenheit -
FPS Fire Protection System
- FSAR a Final Safety Analysis Report ft Feet 1
, GDC AEC General Design Criteria GE General Electric Company gpm. Gallons per minute Gd 0 Gadolinium Oxide 23 HEPA High Efficiency Particulate Air !
- hr Hour HPCS High Pressure Core Spray .
)
- ID Inside Diameter IEEE Institute of Electrical and Electronics Engineers in inch kV Kilovolt k,ff Effective multiplication factor i
______m_______.
I ABBREVIATIONS (Cont'd) kilowatt i
kW kW/ft kilowatt per foot kW/1 kilowatt per liter ib Pound lb/hr Pounds per hour LHGR Linear Heat Generation Rate '
LOCA Loss-of-Coolant Accident LPCI Low Pressure Cuolant Injection LPCS a Low Pressure Core Spray 1
, LPZ Low Population Zone m Meter m Square meters MM Modified Mercalli (earthquake intensity)
MCHFR Minimum Critical Heat Flux Ratio MCPR Minimum Critical Power Ratio
~
mhos/cm Reciprocal ohms per centimeter mph Miles per hour mrem Millirem urem/yr Millirem per year MSL Mean Sea level datum MSLIV Main Steam Line Isolation Valve MWe
- Megawatts electrical MWt Megawatts thermal
l P
f)
\.
I l'
l.
l t 1
. 1 ABBREVIATIONS (Cont'd) !
NDT Nil Ductility: Transition {
NFPA National Fire Protection Association NOAA National Oceanic and Atmospheric Administration NPSH Net Positive Suction Head NSSS Nuclear Steam Supply System
OD Outside Diameter PCA d Primary Coolant Activity PDA Preliminary Design Approval g PMF Probable Maximum Flood <
PMH Proba.ble Maximum Hurricane PMS Probable Maximum Surge 1
PHP Probable Maximum Precipitation PSAR Preliminary Safety Analysis Report !
~
psi Pounds per square inch i psid Pounds per square inch differential psig Pounds per square inch gauge (above atmospheric)
QA Quality Assurance QC Quality Control q
RBCCWS Reactor Building Closed Cooling Water System RCPB Reactor Coolant Pressure Boundary l
______________A
, ABBREVIATIONS (Cont'd)
Rem- Roentgen equivalent man RER Residual Heal Removal
- rpm Revolutions per minute .
RPCS Rod Pattern Control System RWCS Reactor Water Cleanup System
. scfm Standard Cubic Feet Per Minute 4
. SER- Saf'ety Evaluation Report sec/m3 Seconds per cubic meter
! l . SGTS ' Standby Gas Treatment System d
' i SWS Service. Water System .
TLD Thermal Luminescent Dosi, meter UO 2 Uranium Dioxide pCi/see Microcuries per second X/Q Relative Concentration i
i i
l
t <
s a
, l o 1-1
\
1.0 INTRODUCTION
AND GENERAL DESCRIPTION OF THE PLANT 1.1 ' Introduction The initial AEC policy statement on standardization of nuclear *
- i i power plants was issued on April 28, 1972. It provided the impetus l
for both industry and the AEC to initiate active planning in their respective areas in order to realize the benefits of standardization while maintaining protection for the health and safety of the public 1
and for the environment. In a subsequent statement issued on March 5,
, 1973, the AEC announced its intent to implement a standardization dpolicy for nuel, ear power plants. The s,t,andardization policy presents
, three procedural options for standardization applications. Option 1
- is a " reference system" concept that involves the review of an entire
( facility design or major fractions of a facility design outside of the context of a license application. The standard design would be referenced in subsequent license applications. Option 2 is a "duplic ate I
plant" concept in which a 1 hited number of duplicate plants are to be I
constructed within a limited the span. Option 3 is a " License to J .
Manufacture" concept in which a number of identical plants would be i
manufactured at one location and moved to a different location for operation.
On April 30, 1973, General Electric Company (GE or applicant) filed the General Electric Standard Safety Analysis Report (GESSAR) e in response to option 1 of the Commission's policy statement. On 1 l
} Jtily 30,1973, the GESSAR application was docketed.
i The review of GESSAR is being carried out by the staff using a j
( l l similar procedural sequence to that used for custom plant reviews. l The initial phases, i.e., preliminary review, question rounds, etc.,
b 9 e 3 ,
e 1-2 are analogous to the normal construction permit stages of review; I
however, the conclusion of the review will not result in the
'] __ __ . granting of a construction permit. Instead, . Preliminary Design - - !
Approval (PDA) will be issued following satisfactory completion of 1
the staff and ACRS reviews. Specific mechanics of the staff review, change procedures, supplementary SER's, post-PDA review, Final Design Approval (FDA), duration of the FDA, and FDA changes are covered in the AEC standardization program plan issued in June 1974.
GESSAR contains safety information for a BWR-6/ Mark III nuclear power plant, including the nuclear steam supply system (NSSS), the
- i. ,
, engineered safety feature systehis, the containment and auxiliary buildings, the control room, radioactive yaste system and related systems and structures. This complex is [eferred to as the nuclear island. See
( !
{ Figure 1.1 for the scope of the nuclear island. The rated core thermal power of the plant is 3579 MWt (1220 MWe net). The ECCS designs basis 4 d
power is 3758 MWt (105% of rated power). Because it is not related to l l
i any particular site, it contains site envelope parameters to which its '
i design is applicable. These site envelope parameters have been chosen 1
to permit utilization of the standard design in much of the United States.
'l j Not included in the GESSAR nuclear island scope are the turbine-1 I generator and auxiliaries, the turbine building, portions of the main 1
steam system (beyond the isolation valves), the main condenser, the l l
1 circulating water system and intake structure, condensate storage facilities, offsite electrical power, the ultimate heat sink, raw and 0
e
. o. , .
o -
1-3
(:
-potable water systems, parts of the service and instrument air systems outside the nuclear island, the auxiliary steam system and, of course, the site.
.,* Since GESSAR does not cover the entire facility, it is necessary l
to specifically and extensively describe the safety-related interfaces
.l -
between the nuclear island and the balance of plant. The interface information addresses the pertinent safety-related design
. requirements including the dimensional, structural and operating
] environment inputs to transient.and accident analysis, and performance l
requirements n.ecessary to assure the compatibility of the nuclear island to its mating portion of the plant. In some cases, ranges of.
1 interface parameters are more desirable to provide a sufficiently broad
( envelope and thereby reduce the pos,sibility of re-review. This may
- mean that " worst case" combinations will need to be considered, although judgment is needed since these can sometimes lead to unrealistic con-clusions. Care must be exercised in selecting the particular parameters that are specified as interfaces. In general, an interface parameter should be established by the standard design applicant to whom that
.' parameter is important for proper operation of equipment with1n his scope of design or whose scope of equipment design determines the value
{ of that parameter. However, GE is not required to specify the detailed design parameters for a system that is provided by the BOP designer, I
j even though that system is the means by which an interface parameter is l controlled. The detailed design of that system is the responsibility of the BOP designer. Only the interface parameter value needs to bc l specified by GE with the burden of providing a suitable system to
'I
- a
- y ... , , , .et e 9 s 8 =. e*v****** '- *'*** *
- tv * * ---e- , . . < ,
_ s -
S 4 . ,
. )
.. g a 1
, 1-4
(.
achieve the value falling on the BOP designer (or utility applicant).
The interface parameters specified by GE, however, should include
. requirements imposed on systems, components and structures not addre'ssed ~
in GESSAR as well as requirements imposed by those latter items en systems, components, and structures addressed in GESSAR.
This Safety Evaluation Report (SER) summarizes the results of the technical evaluation of GESSAR as performed by the Commission's Regulatory staff and delineates the scope of the technical matters ,
considered in evaluating the radiological safety aspects of the
.I aproposed. facility necessary to provide the standard design a
, Preliminary Design Authorization (PDA) which will make it an acceptable.
" reference system" under Option 1 of our " Standardization Policy."
( -
In the course of the safety review of the material submitted, we held.a number of meetings with representatives of the General Electric
, Company and their consultants to discuss the plant design and performance l
. under postulated accident conditions. During our review, we requested the applicant to provide additional information that we needed for
.I our evaluation. This additional information was provided in amendments
,, to the application. As a result of our review, a number of changes were made in the facility design. These changes are described in the i
a applicant's amendments and are discussed in appropriate sections of this report. Section 1.7 provides a listing of the principal design
] i changes which were made. A chronology of the principal actions relating to the processing of the application is attached as Appendix A to this Safety Evaluation Report (SER). .
(
-- y .
. I 1-5 * .
F; ; .e ;~
- ;t,.9 . ."'* , . N .}* , , ; . . <! ,
- .s .. '
. . . . j I
- i
(. .
1
,
- s b
l i
y Offgas System .
)
< 1 Radwaste Building
.i Auxiliary Building . .
. a
.. .... x.. ,
Diesel Generator y Control Building l (' Building # y Reactor Building i y Diesel-Generator '
y "_ Buildi.ng Fuel Building . .!
. l t
J1
-~ '
F.igure 1.1 Nuclear Island ,. ,
i,
.=
4
( -
6 .
SP g e
l I f i
o ^
l-6
( '
The review and evaluat,1on of the proposed design of the facility reported herein is only the first stage of a continuing review by the Atomic Energy Commission's Regulatory staff of the design, construction, -
and operating features of the GESSAR facility. Utilities will reference GESSAR and match it with BOP designs and obtain construction permits.
Construction activities will be accomplished by applicants under the surveillance of the Commission's Regulatory staff. We intend to review
, the final design of GESSAR when it becomes available to determine that the Commission's safety requirements have been satisfied prior to i
, aprovidingaFinalDesignAuthorizationJFDA).
Then a utility would
. *; again be able to reference GESSAR to meet our operating license
} -
requirements. After an operating license is issued, the facility
( may then be operated only in accordance with the terms of the operating license and the Commission's regulations under the
, j continued surveillance of the Commission's Regulatory staff.
l 1.2 General Plant Description
}
l The nuclear steam supply system (NSSS) for GESSAR utilizes the
)l BWR-6 class of boiling water reactor. Each NSSS will have 20 jet l
I pumps supplied by two recirculating water lines, four main steam 11nes, and two feedwater lines. Fuel rods for the reactor will contain slightly i
, enriched uranium-dioxide (UO 2
) in sintered ceramic pellets. Some of l the fuel rods will have ceramic fuel pellets that contain gadolinium-
.l oxide (Gd230 ) in a mixture 'with the uranium-dioxide. The gadolinium j will serve as a " burnable poison" designed for power pattern and k
4
4 s ,
w e- .-
1-7 l
f(
-reactivity control. The fuel pellets will be enclosed in Zircaloy-2 l
i cladding tubes which will be evacuated, backfilled with helium, and i
. . k sealed by welding Zircaloy end plugs in each end. .A fuel channel will
, ' . lf ,
1
., enclose' a bundle of 63 fuel rods in an 8 x 8 array (one ' fuel rod' position l will contain a water filled rod). Water flowing through the core will.
serve as'both a neutron moderator and as a coolant. Movement of. water '
i and a two phase water-steam mixture through.the core will be accomplished by the driving force from the 20 jet pumps (10 per recirculation.line) i
",' and two recirculation pumps and,from convective forces. Steam from the
.i d boiling process, in the reactor core will. be demoisturized and dried, .
then vented'through the four main steamlines to the turbine-generator !
system (which is outside the GESSAR scope) where its energy will be
(' converted into electricity. The co,ndenser and condenser cooling system I
are not a part of the GESSAR scope.
- An off-gas treatment' system' consisting of a recombiner, condenser, j i
moisture separator, and deep-bed charcoal filters will provide for: l j retention of noble gases for decay to acceptable concentration levels
'l l prior to release from the plant's vent.
. r
,. The reactor coolant pressure boundary will include the reactor
{
vessel, the two recirculation lines, and main steamlines, feedwater
'! lines, and branch lines to their outermost isolation valves. !
1 Enclosing the reactor system will be a reinforced concrete i A,
- cylindrical structure (called the "drywell"). Enclosing this "drywell" 3 structure is the steel " containment" structure. The drywell's t
9 .
__o
s -
I ,
- 1-8 I function is to force most of the steam, released in a postulated accidental pressure boundary break, through the suppression pool
~
located at the bottom of the containment, thus, condensing the steam. .
(vapor suppression) and limiting the pressure buildup below the 1
containment's maximum design pressure of 15 psig. Piping restraints have been designed and will be installed within the containment to limit the movement of piping during its postulated post-rupture move-ment (pipe whip) so that safety related components are appropriately protected., A hydrogen control' system is included which will limit
,the concentration of hydrogen evolved during a postulated loss-of-coolant accident which, over a'relatively short time period, could build up to unacceptable levels,in the drywell by mixing it with the air in the larger containment volume. Long term hydrogen buildup
{ will,be controlled by oxygen-hydrogen recombiners located within fi the auxiliary building. Isolation of the primary containment will I occur automatically whenever there exists a potential for the release i
i of radioactivity due to high activity levels in containment. For instance, the primary containment and the nuclear steam supply system I will be isolated and shut off, respectively, for the unusual conditions j of low water level in the reactor vessel, high radiation level in the main steamline, main steamline high flow or low pressure, drywell high
.j pressure, and other conditions described in GESSAR,
.+
I k
. 3
9, '
.*, )
o '
1-9
(
The reactor protection system (RPS) will provide the means to protect against conditions that may cause fuel failures or a breach-
- , ing of the nuclear system process barrier, thereby limiting uncontrolled
^
l
,. releases of radioactivity to levels within 10 CFR Pa t 100 requirements.
The RPS will initiate a reactor scram following an abnormal operational I
transient or pressure pulse, or following a gross failure of fuel or-of the nuclear system process barrier. The RPS will be a reliable system designed to meet the standards specified in IEEE-279. !
i I
. Normal reactivity control or rapid scram (shutdown) of the '
i
" reactor will be. achieved by the bottom-entry cruciform-shaped control l rods (neutron absorbers) that will be moved vertically in the spaces between fuel assembly channels by a hydraulic mechanism; water is the l
(* -
l hydraulic fluid. For rapid insertion, nitrogen under pressure in i
an accumulator provides the driving force. Each control rod will be f 4 i independent of the other rods and have its own hydraulic control I
system. A standby liquid control system will also be available f for use in injecting a* boron solution into the reactor for emergency, long-term reactivity control.
l
.; Engineered safety features will provide the capability to 1solate containment, shut down the reactor, restrict radioactivity t
{ releases to acceptable levels, provide for heat removal for long-term
.j
'j . core cooling, and condense steam with1n the containment. Details t 1
- on these engineered safety features are presented elsewhere in this -
, . ],
- Safety Evaluation.
t- I k
2 . . ~ . . . . - . .... ., . . , . - - - - - . . . .
9 -
- 'o e l ,
1-10 i 4
. i The containment will house the reactor and its pressure-suppression type primary containment system. The auxiliary building i will house the engineered safety features and their auxiliaries and * ' '
, division 1 and 2 switchgear.
Operation of the standby gas treatment system (SGTS) will produce a negative internal pressure, after building isolation, ;
such that the atmospheres within the building enclosing the containment and within the auxiliary building will be filtered and discharged via the SGTS.
a Other safe,ty related structures su,c..h as the fuel building, the
, control building, the radwaste building, and the diesel generator
~
buildings are also described in this SER.
(, Our review of GESSAR was conducted like a standard construction -
permit review with the additional review of the interface area. We !
used the Commission's Regulations and Regulatory Guides as the t
bases for our review.
i ' In order to make it easier for GE to freeze i their design, we have taken the following positions in applying
.i k ,. 4 Regulatory Guides to GESSAR: First, we established March 1, 1974, l
as the cutoff date for GE to demonstrate compliance with published i regulatory guidance, i.e., published Regulatory Guides. Second, for 4
guides published after that date, we will discuss the intent of the
- guide with GE and agree as to the extent to which that guidance.should I
- be incorporated into post PDA revisions to GESSAR. Finally, for items i
of major safety importance, we will require immediate compliance. We '
t :
feel that there will not be many of these and they will be limited I k
9 -
1-11 I.
to concerns similar to tho e associated with pipe breaks outside containment and post-LOCA hydrogen control.
1.3 Shared Systems *
- i GESSAR is a design for a single unit. Shared systems have not l
been discussed or reviewed for GESSAR. Structures, systems and components I
within the nuclear island that are important to safety will not be shared. Some BOP items that are related to safety systems may be shared.
Included in this area are such things as offsite power and the ultimate
,, heat sink. The detailed designs will be reviewed in the context of
.4a utility's application for a construction permit. ,
- 1. 4, Comparison with Similar Facilities ,
Many features of CESSAR are new GE designs, however, many aspects
( of the plant are similar to those we have evaluated and previously l approved for other nuclear power plants. To the extent feasible and appropriate, we have made use of our previous evaluations during l
our review of those features that are similar to GESSAR. The results I
i of our review of GESSAR, as listed in this SER will be somewhat different l
.I l than previous SER's. The GESSAR SER will be different in that it is intended to stand on its own with a minimum use of references to f previous designs and reviews. To the extent possible, we have j presented, in this SER, our technical bases for each safety conclusion
'i reached. -
.i To assist in better understanding the relationship of the GESSAR
- (BWR/6-Mark III) design to other GE designs, GE has. presented a comparison of principal design features with Perry, Grand Gulf and t
I r
,-e -
'1-12
{ La Salle in GESSAR Tables- 1.3.1, 1.3.2, 1.3.3,-1.3.4, 1.3.5, 1.3.6, and 1.3.7.. A listing of principal parameters and features is presented in Table 1.1 of this SER. Our SER's for these other applications * -
are available for public inspection in'the PDR at 1717 H' Street, N.W., Washington, D.C. 20545, 1.5 Identification of Agents and Contractors '
As stated previously, General Electric Company is responsible for the " nuclear island" scope of supply. Future utility applicants
_; referencing GESSAR will' retain their own architect engineers, con-structors, turbine-generator vendor, and consultants as needed. We
. . . . .. 1 .
will need to conclude for each future application that the utility applicant,.along with his contractors, is technically competent to manage, design, construct and operate a specific reactor prior to C' -
issuance of a CP.
1.6 Summary of Principal Review Matters
~
Our technical review and evaluation of the GESSAR information submitted by the applicant included and excluded the principal matters discussed below.
a.
We evaluated the site design envelope parameters including the wind loadings, design bases tornado, design bases flood elevation, the design bases earthquake, the snow loading, and 4 maximum precipitation. Items in our site evaluation that will be covered in future applications for plants referencing GESSAR, !
3 1 but not in GESSAR are the exclusion area determination and control, the population distribution, use of adjacent lands b
r I
k
1-13
$ Table 1-1 Principal Parameters and Design Features of GESSAR, Perry, Grand Gulf and LaSalle Parameter or Feature
[ GESSAR Perry Grand Gulf LaSalle Rated Power Level (MWt) '3579 3579 3833 3323 Desi'gn Poier (MWt)
.. (ECCS design bases) 3758 3758 4025 3489
. Net Electrical Output (MWe) 1220 1220 1290 1086 No. Fuel Assemblies 732 732 784 764 Fuel Rod Array 8x8(63 rods) 8x8(63 rods) 8x8(63 rods) 7x7(49 rods)
No. of Control Rods 177 177 1 93 185 Maximum Design Linear Power (kw/ft) 13.4 13.4 13.4 18.5 Reactor Vessel, ID (in.) 238 238 251 251 Vessel Height, inside (ft.in.) 70-10 70-10 73-0 72-11 Vessel wall thickness (in.) 5.7 5.7 6.14 6.14 CladThickness(in.) 1/8 1/8 1/8 1/8 No. recirculation loops 2 2 2 2 Recirc. pump flow. rate (gpm) 35,400 35,400 44,900 47,250 No. jet pumps 20' 20 24 20 No. Ste"am Lines 0
4 4 4- 4 Ste.am Line ID (in.) 26 26 28 26 6 '
Core Flow (16/hr) x 10 105 105 113.5 108.5 Steam Flow (16/hr) x 106 15.396 15.396 16.488 14.95
(- -
ECCS 5
No. LPCS pumps 1 1 1 Flow (gpm) 6000 6000 7000 1
)
6350 at 122 psid l No. HPCS pumps
, j l 1 1 1 1 Flow (gpm) 1465 1465 1550 1550 at 1130 psid l
) No. RHR pumps 3 3 '3 j 3 (LPCI mode)
.1 Flow Rate at.
- l 20 psid per pump 7100 7100 7450 7450
, No. ADS Systems 1 1 1 l 1
j Containment Design Concrete Concrete Concrete Concrete, Drywell Cylinder Cylinder Cylinder Cone Section Desi.gn Pressure (psig) 30 25 30 45 Containment freestanding freestanding freestanding -
steel steel steel
( 'j cylinder cylinder cylinder Design Pressure (psig) ^15 15 15 7
l
1
- 9 f
- 1-14
~I and waters, the effects, of the presence of nearby military * '
' facilities, effects of industrialior transportation accidents close to the plant, the consequences of an aircraft crashi the * -
, ' evaluation of the meteorological measurements program, the
.s -
effects of toxic chemical or gaseous releases on plant operation, t
and the effects of explosions near the facility.
Our site' evaluation was performed to determine whether the considered site characteristics were appropriately included in the design. We also considered whether the design of the facility in'luding c all engineered gIfety features, could be
,, e::pected to meet the Commission's siting criteria (10 CFR Part 100) when located on the niajority of potential reactor sites C within the U.S. .
- b. W'e evaluated the design and expected performance of the nuclear island's structures, systefus, and components important to safety to determine whether they are in accord with the Commission's i,
General Design Criteria (GDC), the Commission's Quality
~
Assurance Criteria, and other applicable guides, codes and standards, and whether any departures from criteria, codes i
- f. j and standards have been identified and justified.
i
- c. We evaluated the expected response of the facility to various
?; . . _
, j anticipated operating transients and to a broad spectrum of postulated accidents and determined that the potential consequences (
of a few highly unlikely postulated accidents (design basis j
( 'l .
0
% .j r
____-__.._-_____m_ _ _ _
3 i '
1-15 4
,(:
accidents) would exceed those of all other accidents considered.
We performed conservative analyses ~ of' these design. basis accidents l j and determined that the calculated potential offsite doses that- - -
might result in the very unlikely event of their occurrence would ;
l
-l
'be well within the Commission's guidelines for site acceptability, ,
as given in 10 CFR Part 100, for typical sites.
j.
I
- d. We did not evaluate the plans for the conduct of plant operations, the organizational structure, the technical qualifications of the operating and. technical support personnel, the measures taken !
d in t.he,unlik,ely event of an accident,.that might affect the general
'5 public since these items are outside GE's scope of' supply in GESSAR
, and will be addressed in future specific utility applications referencing GESSAR.
-e. We evaluated the design of the systems provided for control of l 3 l the radiolog~ical effluents from the plant and determined that l l
1 these systems, in conjunction with an acceptable BOP design, 4 l
1 reasonable meteorology and site boundaries, will be able to l
control the release of radioactive wastes within the limits of
)1
,1 the Commission's regulations in 10 CFR Part 20 and that the
. plant will be operated in such a manner as to reduce radioactive releases to levels that are as low as practicable in accordance with the Commission's regulations in 10 CFR Part 50.
1 t
- g. .
I t
l
^
l 1 1
- I-y
, i o e
., l 1-16
( -
- f. We did not evaluate the financial qualifications of the applicant to determine whether the financial position of the applicant
.--~~ is adequate to design and const'ruct the facility, since GE is.
~
,s not funding construction activities.
The GESSAR appl 1 cation contained new and significantly modified features that are different than previous BWR designs that have been evaluated'by the staff. These items are noted below,
- a. GE has proposed new designs in five instrument and control
, areas .- These changes are a.new control position detection a systeg; a new method of increasing.the negative reactivity during a scram; the use,of ganged control rods; a revised rod
- pattern control system; and finally a solid state, 2-out-of-4
(' protection system. ,
I b. Based on further boiling transition tests, GE has developed a new figure of merit for expressing the BWR thermal margin.
The result is a new GE Thermal Analysis Basis or GETAB.
- c. We worked with GE in obtaining resolution to some previously !
)
outstanding problems including post-LOCA H2 generation and control, a main steam line sealing system, suppression pool I
i bypass and testing, drywell structural and leakage testing, 1
quality classification of main steam radwaste and auxiliary f
.- )
J systems and interface definition and quantification.
I y
J
~
l
' 1-17 l .
Facility Modifications as a Result of Regulatory Staff Review 1 1.7 !
l As a consequence of our review, a number of design changes were- -
These modifications are discussed in greater'
- made to the GESSAR design. g
/
The principal. changes detail in appropriate sections of this report.
i which were made are as follows: ,.
An increase'in the wind loading, snow and ice loadings (1) and elevation of ground water with respect to the foundation mat will permit the plant to be used on more sites. i
'(2) The s,eismic instrumentation program has been augmented.
A main steam line sealing system apd third valve have been (3) . ...
a
, added. .
(4)
The RCIC system has been upgraded to an engineered safety feature.
.3 ill be verified by testing.
(5) ,The operability of active components w ;
Design measures have been 'taken to protect against the dynamic I
(6)
.l !
effects associated with pipe breaks.
A finite element method will be used to analyze various soil f (7) conditions to evalu' ate soil structure interactions.
! Fuel building has been upgraded to withstand tornado missiles.
(8)
,l
- Methods for Seismic Analysis comply with requirements of i (9)
Regulatory Guides 1.60 and 1.61.
i (10)~ Main Steam Line and Feedwater Piping Reclassification.
1 9
(11) Nhrk III containment changes to accommodate pool swell .
- testing results.
l
(
l :
! 1-18 .
(12) Increased dryvell design pressure margin. l (13) Tests to verify that controls on stainless steel are adequate I to prevent sentization.
a (14) GE agreed to preoperational vibration tests on Class 1 and 2
' piping systems.
1.8 Requirements for Future Technical Information 1.8.1 Development of BWR Technology The applicant has identified in Section 1.5, the research and These programs development programs applicable to the GESSAR plant.
are aimed at verifying the nuclear secam supply system and containment J.
a . .... The objectives, schedules designs and confirming the design margins.
for cotapletion, and current results are" summarized in referenced GE topical reports.
We conclude that the applicant'has identified and will perform the research and development necessary for the design and safe
, in the event
. )
operation of the plant on a timely schedule, and that, the results of any of this work are not successful, appropriate '
restrictions on operatieri can be imposed or proven alternate designs f j can be utilized to protect the health and safety of the public, !
I
.' 1.8.2 Outstanding Items of Review At the time of this report, we had not completed our review of l These ,
sertain items due to the lack of adequate information. l 1
) items are listed in the enclosure to the transmittal letter. I i
l l
, l .
I
2-1 2.0 SITE CllARACTERISTICS 2.1 Geography and Demography These will be reviewed for each application referencing GESSAR.
2.2 Nearby Industrial, Transportation and Military Facilities The proposed site will be evaluated on a case by case basis. Our review of the site will center on those items required by applicable version of the " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants." We will review the site location description, and boundaries, the population and population distribution, nearby
" military. bases, airports, oil or gas pipelines and chemical plants, as well as major transportation routes, to see what potential detri-t mental effects these could have on the plant.
2.3 Meteorology Details on the atmospheric diffusion characteristics of a proposed nuclear power plant site are required in order that a determination may be made that postulated accidental as well as routine operational releases of radioactive materials are well within AEC regulatory guide-lines. The meteorological characteristics of a proposed site are determined by staff evaluation of meteorological data collected at the site in accordance with Regulatory Guide 1.23, "Onsite Meteorological Programs". This procedure will be followed in review of any application for a site which will contain the GESSAR standard plant. The discussion which follows relates to the " envelope" of site meteorological conditions proposed by the applicant for the standard plant which provides an indication, in advance of the examination of a particular site, of the type of site for which the standard plant as proposed is suitable.
l I
2-2 2.3.1 Regional Climatology The General Electric Standard Plant is designed for an environmental temperature range of from -40*F to +115'F (-40*C to +46*C), which is adequate for over 90 percent of the contiguous United States. The design basis wind velocity of 130 mph at a height of 30 feet above grade with a recurrence interval of 100 years would be adequate for greater than 90% of the sites. The design basis maximum tornadic wind velocity of 360 mph and maximum pressure drop of 3.0 psi is ade-
"'quate for all areas of the United States. The design basis snow and ice load of 50 pounds / square foot is adequate for many areas within
( the contiguous United States, although such loads may be exceeded in several regions, especially in some northern portions of the country.
In these regions, some of the other loads (wind and seismic) are typically lower. Therefore, the total loads on the structure, even with the higher snow and ice loads, will be less than the total loads for other regions. In'other words, even where the snow and ice loads are greater than 50 pounds / square foot, that may not be the limiting design loading for the structures. Sites for which the extreme tempera-tures are outside the range presented above, for which the design basis
, wind of 130 mph may be exceeded, or for which the design basis snow and ice load may be inadequate will be evaluated on a case by case l basis Some east and gulf coast sites may be excluded since the winds associated with the Probable Maximum Hurricane could exceed 130 mph.
'2-3 2.3.2 Local Meteorology The applicant's design basis of atmospheric d1spersion conditions,-
equivalent to those presented in Regulatory Guide 1.3 - Assumptions i Used for Evaluating the Potential Radiological Consequences of a Loss-
\
l of Coolant Accident for Boiling Water Reactors,-will limit the plant's l
l- suitability to no more than fif ty percent of the potential sites within the contiguous United States, exclusive of other site conditions.
That is about fif ty percent of the sites that.we have reviewed have had meteorological conditions that are better and fifty percent have had worse conditions. This is not meant to imply that GESSAR is only good l
, for half of the sites in the United States since a larger exclusion -
boundary will compensate for less favorable meteorology as will a more leak tight containment or more sophisticated cleanup systems.
Therefore, case by case analysis will be done for about fifty percent '
of the potential nuclear plant sites within the United States. In any event, a meteorological data collection program, conforming to the provisions of Regulatory Guide 1.23 - Onsite Matco ological Pro-grams, will hcve to be established at each site to verify th'at the atmospheric dispersion characteristics of the site are within the design envelope of the plant.
2.3.3 onsite Meteorological Measurements Program The applicant has stated that the dispercion conditions at the site will be verified. Since the onsite meteorological measurements I
program is assumed to be site specific, GE does not include a description of an onsite meteorological measurements program.
l' 2-4 2.3.4 Short-Term (Accident) Diffusion Estimates The procedures used by the applicant (as presented in Regulatory Guide 1.3) to determine accident diffusion estimates are generally acceptable although the assumption of an elevated release from build-ing vents if wind speeds are less than 7 mph is not acceptable to the staff. An individual evaluation of atmospheric dispersion conditions associated with accidental releases from the plant buildings and vents will be necessary to verify that site dispersion conditions are Jwithin the plant design envelope.
2.3.5 Long-Term (Routine) Dif fusion Estimates i
The applicant does not provide any meteorological assumptions or
. t procedures to be used in evaluating the long-term atmospheric dispersion characteristics of a site. Diffusion estimates for routine releases from plant buildings and vents are to be based on meteorological data collected onsite. An evaluation of annual average atmospheric disper-sion values for routine releases will have to be made on a case by case basis at each site using meteorvi, _.m1 data collected onsite in an acceptable manner (see Regulatory Guide 1.23), over at least an annual cycle.
The staff concludes that the meteorological conditions proposed by General Electric envelope the majority of sites in the contiguous United States. Justification at sites where any of the parameters fall outside the envelope listed by GE will have to be presented by the
2-5 i
specific utility applicants. In addition, the final determination of the acceptability of any proposed' site, with respect to meteorological conditions as they affect Part 100 guideline dose ' calculations, will
~
be based upon evaluation of meteorological conditions' applicable to that site in. combination with the applicable exclusion' area and low population' zone distances.
2.4 Hydrology 2.4.1 Floods d
The applicant proposed that grade levels for all safety-related facilities be located at least one foot above the design bases flood
, level, including an allowance for coincident. wind ganerated waves,' and has referenced Regulatory Guide 1.59, " Design Bases rioods for Nuclear Power Plants," for criteria. . We conclude that these bases are acceptable criteria.
Protection of safety-related structures against locally heavy pre-cipitation has been identified as roof drains with 4 inches per hour capacity and overflow capability to limit standing water to 9.5 inches (about 49 pounds per square foot). Site specific drainage, identified as the plant owner's responsibility, will have to be carefully designed to preclude.ponding above plant grade adjacent to safety-related buildings.
2.4.2 _S_afety-Related Water Supply - Ultimate Heat Sink This is a subject that is the responsibility of and will be dis-cussed by the specific utility applicant.
l I
l 1
l 2-6 i
}
2.4.3 Groundwater 4 The applicant initially proposed that all foundation mats be located at or above groundwater levels. Although this criterion is acceptable, the staff noted that many existing or future sites would not meet the criterion. The applicant subsequently modified the cri-terion to provide for groundwater levels up to within 2 feet of plant grade.
2.4.4 Conclusions The proposed flood, safety-r. elated water supply, and groundwater criteria are acceptable to the staff.
2.5 Geology and Seismology ?
The geology, seismology and foundation engineering investigations required by Appendix A to 10 CFR Part 100 and the Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Revision 2 will be reviewed by the staff for each individual site for which an application is made. The staff believes that the proposed seismic design of 0.3g will be adequate for 70% of the nation east of the Rocky Mountains. The limiting condition associated with 0.3g is the potential for liquefaction. This can be overcome by compaction or other means. Due to the potential for significantly higher g values west of the Rocky Mountains, the staff concludes that GESSAR, as is, is not not intended for high seismic areas.
i l
l l
l J
- - _ - _- - _-_- _ a
3-1 3.0 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS 3.1 Conformance with AEC General Design Criteria General Electric Company presented in Section 3 of GESSAR, their evaluation of the design bases for GESSAR, with respect to the AEC's General Design Criteria (GDC) as contained in Appendix A to 10 CFR Part
- 50. Based on our evaluation of the preliminary design and of the proposed design criteria, we conclude that subject to the applicant's adoption of the additional requirements made by us, as discussed in
-this report, they are in conformance with the GDC.
3.2 Classification of Structures, Systems and Comoonents 3.2.1 Seismic classification Except for those items identified below, structures, systems and components important to safety that are required to withstand the effects of a Safe Shutdown Earthquake and remain functional have been properly classified as seismic Category I items. These plant features are those necessary to ' assure (1) the integrity of the reactor j coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents which could j l
result in potential offsite exposures comparable to the guideline j exposures of 10 CFR Part 100, i All other structures, systems and components that may be required )
1 for operation of the facility are designed to other than seismic l i
_____.______- - - ._ ____._____- _ _ ____-_____J
3-2 Category I requirements. Included in .this classification are those :
. portions of Category I systems which are not required to perform a safety function. ' Structures, systems and components important to safety that are designed to withstand the effects of a Safe' Shutdown
' Earthquake and remain functional have been identified in an acceptable manner in Table 3.2-1 of GESSAR.
The applicant has classified the discharge piping from the relief I
valve to the anchor point as Seismic Category I and the remainder of a
the piping from the anchor point to the suppression pool as non-Seismic.
-Category I. The staff did not agree with this.non-Seismic Category 1 1.'
classification since GE had not shown that failure of the discharge piping would not damage or degrade other safety related equipment.
In Amendment 20, GE justified their classifications to our satisfaction.
We conclude that non-seismic Category I classification is acceptable.
Analyses of reactor recirculation pump motor behavior following complete loss of cooling water were provided to the Staff. The GE analyses indicate that 1f the initial cooling water loss alarm and the subsequent bearing temperature alarm (about 6 minutes later) are both ignored, the bearings will continue to operate another 6 to 10 minutes before melting. The response further states that such melting will i
i 1 i i
3-3 l i not cause motor seizure and assuming the worst possible steel to steel !
friction, the motor will trip on overload caused by the added friction.
,- The resulting faster coastdown will result in a MCHFR decrease from 1
- 1. the normal 1.436 to 1.350. Preliminary review of these results indicate 1
that the consequences of the cooling water failure are acceptable.
After issuance of the PDA, the staff plans further detailed study of large motor performance under such conditions, including effects of assuming less severe realistic friction factors (following bearing failure)'.upon ultimats ,
s i trip and/or failure' mode of the motor. As these studies progress, further information may be required and certain system changes could be U
indicated.
i The applicant has classified the offgas system as non-seismic.
' Category I. This classification is unacceptable to the staff.
We will require the offgas system be designed to seismic Category I requirements in conformance with the staff technical position, as
- discussed in Section 11 of th.is report and Appendix B. We will report on the resolution of this matter in a supplement to this SER.
The basis for acceptance in the Staff's review has been conformance of the applicant's designs, design criteria and design bases for structures, systems and components important to safety with the Commission's regulations as set forth in-General Design Criterion 2, and to Regulatory Cuide 1.29, " Seismic Design Classification", technical g.
staff positions (such as appendices B and C), and industry standards such as the ASME code.
1 l
t l
.i 3-4 l
l The staff concludes that structures, systems and components import-i~ ' ant to safety that are designed in accordance with seismic Category. I requirements provide reasonable assurance that the plant will perform ,
t in a manner providing adequate safeguards of the health and safety of l 1
the . public .
l 3.2.2 System Quality Group Classification Except for those items . identified below, fluid system pressure-I retaining components important to safety will be designed, fabricated, trected and tested to quality standards commensurate with the importance of the safety function to be performed. The applicant has applied the
- , classification system identified in' Regulatory Guide 1.25-a
" Quality Group Classifications and Standards" to those fluid containing components which are part of the reactor coolant
. pressure boundary and other fluid systems important to safety where reliance is placed on these systems: (1) to prevent or mitigate the consequences of accidents and malfunctions originating within the reactor coolant pressure boundary, (2) to permit shutdown of the reactor and maintenance in the safe shutdown conditions, and (3) to contain radioactive material. These fluid systems have been classi-fied in an acceptable manner in Table 3.2-1 and 3.2-3, Figures 3.2-1 and 3.2-2 of'GESSAR and on system Piping and Instrumentation Diagrams.
The applicant has classified the discharge piping from the relief valve to the first seismic anchor point as Quality Group C and the remainder !
I of the piping from the anchor point to the suppression pool as Quality Group D. The staff did not agree that Quality Group _.
)
3-5 D is an acceptable classification for the major portion of this
. piping since failure of the discharge piping may damage or degrade-other safety related equipment. We will require the piping from the down-stream side of the relief valves to the suppression pool be classified
. Quality . Group C unless GE justifies that there are no unacceptable safety consequences resulting from failure of the Quality Group D portion'of the piping. 64 i:4:., junine taeir classification:, to our sc.tisfcetion in Imendment 20 and t'ncir design is accey' .:.
The applicant has classified the Liquid Radwaste and '0ffgas sys tems as Quality. Group D. We will require the applicant to nupplement his-
" i Quality Group D classification of these systems with those additional q
4 requirements identified in Section 11 of this SER and Appendix B. We will discuss the resolution of this item in a supplement to this SER.
I GE has stated that components of the main steam and feedwater
]
i systems important to the safety of the plant will be designed, fabri-cated erected and tes$ed in accordance with the quality and seismic i
design requirements described in the letter of. April 19,19 74, from J. M. Hendrie of the AEC to J. A. Hinds of General Electric. A copy
]
of this letter is enclosed as Appendix C. These requirements are acceptable as an alternate to the guidelines currently spec 1fied in Regulatory Guide 1.26 (March 27, 1972) and 1.29 (August 1973).
1 The basis for acceptance in the staff's review has been conform-ance of the applicant's design, design criteria, and design bases for i
pressure-rctaining components such as pressure vessels, heat exchangers,
~
storage tanks, pumps, piping and valves in fluid systems important to l
( !
3-6 safety with the Commission's Regulations as set forth in General Design Criterion I, the requirements of the Codes specified in Section 50.55a of 10 CFR Part 50, and to Regulatory Guides 1.26 and 1.29, technical staff positions, and industry standards.
The Staff concludes that flu.id' system pressure-retaining components important to safety that are designed, fabricated, erected and tested to quality standards in conformance with these requirements provide reasonable assurance that the plant will perform in a manner providing adequate safeguards to the health and safety of the public.
3.3 Wind and' Tornado Loadings 7 3.3.1 Wind Loadings t .
All the Nuclear Island Category I structures listed in Table 3.2-1 of GESSAR will be designed to withstand the effects of the design wind, and all Category I systems and components located within will thereby be protected from its effects. Category I systems and components located outside the structures and thus exposed to the wind, will be designed to withstand its effects. )
l The design wind specified for GESSAR has a velocity of 130 mph at
. j an elevation of 30 feet above grade based on a recurrence interval of )
100 years.
The procedures that will be used to transform the wind velocity into pressure loadings on structures, systems or components, and the t
l
.____________N
i l
b- .
3-7 associated distribution of wind pressures and drag coefficients will I be in accordance with the American Society of Civil Engineers Paper No. 3269,'" Wind-Forces on Structures." This paper has'been'widely used and recognized and has been accepted for use by the Regulatory Staff. _j i
The design wind loads will be combined'with other applicable loads as will be discussed in Section 3.8 of this report.
.3.3.2 Tornado Loadings l-All the Nuclear Island Category I structures listed 'in Table 3.2.1-of GESSAR will be designed to withstand the effects of the Design Basis Tornado, and all Category I systems and components located within will thereby be protected from its ef fects. Category I systems and com -
ponents located outside the structures and thus exposed to the tornado, will'be designed to withstand its ef fects.
The Design Basis Tornado specified for the plant has a tangential wind velocity of 300 mph and a transnational velocity of 60 mph. The pressure drop associated with the tornado is 3 psi in 3 seconds.
Furthermore, an appropriate spectrum of tornado-generated missiles is .
also postulated as will be discussed in Section 3.5 of this report.
The procedures that will be used to transform the tornado wind velocity into pressure loadings will be in accordance with ASCE Paper i
)
l
.i
3-8
. No. 3269 except that the pressure will be applied uniformly- over the full' height of the projected area of the structure and no gust factors
.will'be applied. The changes result in a more conservative analysis than the ASCE: paper. Category I structures will b'e designed for the pressure drop associated with the Design Basis Tornado which will be treated as a loed that varies with time. The tornado missile effects will be determined using procedures discussed in Section 3.5 of this' report.- The total'effect of the Design Basis Tornado on Category I~
structures, systems and components will be determined by an appro-priate combination of its individual effects, a
Tornado-generated loads will be ' combined with other applicable loads as will be discussed in Section 3.8 of this report.
t I
GE has stated that all the Nuclear Island Category I structures will be sufficiently separated from other plant structures and components that will not be designed for the Design Basis Tornado. A failure of or secondary missile from a' structure not designed for the Design Basis Tornado will not prevent the plant from being safely shutdown or result in an offsite exposure in excess of the guidelines of 10 CFR Part 100. The safety functions and structural integrity of Category I equipment and structures will thereby be assured.
3.3.3 Conclusions )
We conclude that the procedures that will be' utilized to deter-i mine the loadings on seismic Category I structures induced by the design wind and the Design Basis Tornado specified for the plant are acceptable since these procedures provide a conservative basis for l
engineering design to assure that the structures will withstand such environmental forces. <
( _ _ _ __ _ . _ _ _ . . _ . _ - . _ _ _ - - - -
+
3-9 The use of these procedures provides reasonable assurance that, in the event of a design wind or a Design Basis Tornado, the structural.
integrity of the plant seismic Category. I structures will not be impairsd. Seismic Category I systems and components lodated-within. these structures are thereby adequately protected and will be expected to perform their intended safety functions if needed. Conformance with these procedures is an acceptable basis for satisfying in part the requirements of General Design Criterion #2.
3.4 Water Level (Flood) Design 3.4.1 wFlood Protection The information provided in GESSAR states that "the design basis i, flood elevation 1s approximately one foot below the plant finished grade elevation including allowance for coincident waves and resultant runup". We understand this to mean.that flooding of plant features-will be precluded. We will . verify this flood protection on a case by case basir With this as a design basis for the standard plant, we conclude that the flood protection for seismic Category I structures, systems and components for those plants that reference GESSAR will be adequate.
'3.4.2 Design Procedures With the proposed plant grade one foot above the elevation of the design basis flood, Category I structures, systems and components will be protected from the hydro-dynamic phenomena associated with the flood. The hydrostatic ef fect of the flood, however, will be considered in the design of all Category I structures exposed to the water head.
All these structures will remain stable when subjected to either overturning moments or uplift forces of the flood.
I
l l
(
3-10 3.4.3 Conclusions i We conclude that the procedures that will be utilized to determine the loadings on seismic Category I structures induced by !
the design flood level specified for the plant are acceptable since these procedures provide a conservative basis .for engineering design to assure that the structures will withstand such environmental forces.
The use of these procedures provides reasonable assurance that, in the event of floods, the structural integrity of the plant seismic Category I structures will not be impaired and, seismic Cate-a gory I systems and components located within these structures are thereby ade-quately protected and can be expected to perform their intended i
I safety function if needed. Conformance with these design procedures is an acceptable basis for satisfying in part the requirements of -
3.5 Missile Protection Criteria Structures, shields and barriers that will be designed to withstand !
the effects of the various postulated missiles , are listed in Section 3.5.1 of GESSAR. The missiles that can potentially impact each struc-ture, shield or barrier are identified in Section 3.5.2.
I l
GE considered internally and externally generated missiles in !
their design. We did not include turbine missiles in our review since l the orientation of the turbine with respect to the safety related I 1
features of the site is important in assessing the effects of such 1
i missiles. Th'e effects of turbine missiles will be evaluated on a I
case-by-case basis.
The missiles selected by GE are appropriate and include i
k .
- i. o l .
.3-11 tornado borne as well as internally generated missiles, resulting'from the LOCA or equipment failures. The velocities proposed by GE for the. .
1
. various ' tornado missiles are unacceptable. . . We willLrequ1re the following velocities be used,in the design of GESSAR facilities. .These velocities are based' on our evaluation of maximum drag coefficients for each missile.
4" x 12" x 12'
~
.A - Wood-plank 200 lb 423 fps B - Steel pipe 3" 9, 10 ' long, . 78 lb 211 fps schedule 40' a
C - Steel rod 1" $ x 3' long 8 lb 317 fps 6" $, 15' long,-
~
D - Steel pip 2 285 lb 211 fps
, schedule 40 E - Steel pipe' 12" G ,15 ' long, 743 lb 211 fps schedule 40 F - Utility pole 13.5" 9 x 35' long 1490 lb 211 fps:
G - Automobile 20 ft frontal 4000 lb 74 fps area These missiles are 'to be considered as striking on both hori-zontal and vertical surfaces. Missiles A, B, C, D, and E are to be considered at all altitudes and missiles F and G at altitudes less than 30 feet above all grade levels within 1/2 mile of the facility
, structures. AnyL sites with elevations higher than plant grade within 1/2 mile will have to be examined on a case basis with respect to design against missiles F and G.
l l
l l
l
3-12 The analysis of structures, shicids and barriers to determine the effects of missile impact, will be accomplished in two steps.
In the first step, the potential damage that could be done by the missile in the immediate vicinity cf impact on concrete targets will be determined by the use of the Modified Petry formula. Fur thermore, secondary missiles, that could potentially be generated by spalling of the target, will be prevented by fixing the target thickness well above that determir.ed by penetration. In the case of steel targets, formulas developed by the Stanford Research Institute for estimation "of penetration of missiles will be used. These formulas have been previously reviewed and were found acceptable by the Regulatory I Staff'. 7 In the second step of the analysis, the overall structural response of the target when impacted by a missile, will be determined using established methods of impactive analysis, where the momentum of the missile is transferred to the target to determine the energy that has to be absorbed by the target.
The load of the missile impact, whether the missile is environ-mentally generated or accidentally generated within the plant, will be combined with other applicable loads as will be discussed in Section 3.8 of this report. ]
l We conclude that the procedures that will be utilized to deter-mine the ef fects and loadings on scismic Category I structures and (
l l __-_______a
l 1
l 3-13 l
l missile barriers induced by design basis missiles selected for the
- plant are acceptable since these procedures provide a conservative
! . basis for engineering design to assure that the structures or barriers are adequately protected against the effects of missile impacts.
The use of these procedures provides reasonable assurance that, in the event of design basis missiles striking seismic Category I structures or other missile barriers, the structural integrity of the structures and barriers will not be impaired or degraded to an extent that will-. result in a loss of requ' ired protection. Seismic Category:-
1 systems and- components protected by these structures are, therefore, adequately protected against the. effects of missiles. Conformance with t .
,these procedures and the use of appropriate missile velocities 'is an i acceptable basis for satisfying the requirements of AEC General Design Criteria 2 and 4.
3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping l
With respect tio systems located inside containment, GE states that l
the criteria to be employed for determination of the systems which are. (
evaluated, the locations and types of piping breaks which are postulated I and the protection measures against pipe whip to be provided will be consistent with the provisions of Regulatory Guide 1.46, ,
t
" Protection Against Pipe Whip Inside Containment."
l
. 1
l l 3-14 .
l i
The methods of analysis described in GESSAR will adequately account for the dynamic loadings on systems, structures and components that are associated with pipe rupture assumptions and will provide adequate l assurance that the containment structure, unaffected system components, and those systems important to safety which are in close proximity to "the systems in which postulated pipe failures are assumed to occur, i
will be protected. '
The dynamic analyses described in GESSAR for determination of restraint loading resulting from postulated pipe ruptures, will yield conservative results for the large clearance, large deformation re-straints described in GESSAR (i.e., a gap size of approximately six i
inches) when used with the thrust forces calculated in accordance with !
i l
f
3-15 the relationship given in Section 3.6 of GESSAR. Design limits proposed by GE in Section 3.6.3.1.5.1 of GESSAR for use in the design of the pipe whip restraints will result in deformation limits as conservative as ours for all methods and all material employed.
The methods used for formulating the hydro-dynamic forcing functions induced by pipe rupture and the dynamic analysis for the pdpe whip , motion provide an acceptable basis for restraint design.
The criteria used for the identification, design, and analysis of piping systems uhcre postulated breaks may occur consitute an acceptable decign basis in meeting the applicabic requirements of AEC General Design Criteria 1, 2, 4, 14, 15, 31 and 32.
The provisions for protection assinst the dynamic effects associ-ated with pipe ruptures,and the resulting discharging coolant provide adequate assurance that, in the event of the combined loadings imposed by an earthquake of the magnitude specified for the Safe Shutdown Earthquake and a concurrent single pipe break at one of the design basis break locations, the following conditions and safety functions will be accommodated and assured:
(1) the magnitude of a design basis loss-of-coolant accident cannot l
be aggravated by potential multiple failures of piping, j 1
I 1
l 1
L__------_-------
[
L 3-16
)
(2) the~ reactor emergency core cooling systems can be expected to perform their intended. function assuming a single failure.
. The applicant has stated in GESSAR that it will incorporate the -
criteria ~of the AEC letter from Mr. J. F. O' Leary dated July 12, 1973, in the analysis for high and moderate energy line breaks outside containment.
During the course of piping design for these high energy systems, con-sideration will be given to physical separation of pipes from other f safety-related equipment and instrumentation, as the primary protection against the effects of a postulated pipe break and single failure criteria.
a Acceptab'le specific application of these criteria to thoce portions of
_ piping which pass through containment have not been provided and so must
'h
- I be evaluated by January; 1975.
In implementing thcse criteria, the applicant will designate design basis break locations throughout all high energy piping systems. These l
~
postulated break locations will be chosen on the basis of highest relative stress, or significant changes in flexibility of the piping.
~
The protection provided'against the dynamic effects of postulated pipe breaks and discharging fluids in piping systems containing high i
energy fluids and located outside the containment is adequate to '
prevent damage to structures, systems and components to the extent considered necessary to assure the maintenance of their structural integrity. Such protection provides reasonable assurance that the safe shutdown of the reactor can be accomplished and maintained. l
3-17 In addition, for those piping systems not considered as high energy systems, the applicant will postulate leakage cracks in accordance with.the above referenced O' Leary letter to assure that essential equipment and components are protected from fluid spraying, flooding and con-sequent environmental conditions developed.
Except as otherwise noted above, the criteria used for the identification, design and analysis of high and moderate energy fluid lines outside containment where postulated breaks and cracks may occur constitutes an acceptable design basis for satisfying the applicable re,quirements of AEC General Design Criterion 4.
- j. 3.7 Seismic Design 3.7.1 Seismic Input
- The input seismic design response spectra (OBE and SSE) and the damping values applied in the design of Seismic Category I structures, systems and componento comply with the provisions of AEC Regulatory Guides 1.60, " Design Response Spectra for Noclear Power Plants" and 1.61, " Damping Values for Seismic Design of Nuclear power Plants",
respectively.
The synthetic time history to be used for Category I plant com-ponent design is adjusted in amplitude and frequency to envelope the response spectra defined in Regulatory Guide 1.60.
1 I
1 3-18 The applicant will revise Section. 3.7.1.2 to provide the response spectra which envelope the design response spectra shown in Figures q 3.7-1 and 3.7-2, of GESSAR for all damping values used in'the design.
Conformance with provisions .of AEC Regulatory Guides 1.60 and 1.61 provides reasonable assu)ance that for an earthquake whose intensity is 0.15g for OBE, and 0.30g for SSE, the resulting accelera-tions and displacements imposed on Category I structures, systems and components are adequately defined to assure a conservative basis.for the design of such structures, systems and components to withstand athe consequent seismic loadings. Compliance with the provisions of these Guides constitutes an acceptable basis for satisfying the
- requirements of AEC General Design Criterion 2.
1 -
l 3.7.2 Seismic System Analysis 3.7.3- Seismic Subsystem Analysis Modal response spectrum multi-degree-of-freedom and time history methods form the bases for the analyses of all major Category I struc-l tures, systems and components. Governing response parameters are combined by the square root of the sum' of the squares of each modal 4 l
response to obtain the modal maximums when the modal response spectrum method is used. The absolute sum of the modal responses are used for close,1y-spaced modal frequencies. The square root of the sum of the squares of the maximum co-directional responses are used in accounting for the three components of the carthquake motion.
Floor response spectra inputs to be used for design and test verification l- of structures, systems and components are generated from the time history method. Vertical seismic-system dynamic analyses are employed for all structures, systems and components where analyses show j..
l .
3-19 l significant structural amplifications in the vertical direct 1on. The system and subsystem analyses are performed based on elastic theory. ,
l The finite element approach is used to evaluate the soil-structure interaction effects for deeply embedded Category I structures. Deeply embedded has been defined by CE as being the case where the embedment is greater than 15% of the smaller horizontal dimensions of the foundation mat. Soil spring, multipic-spring shear beam and other equivalent methods are used to determine the soil-structure interaction effects of cases other than deeply embedded Category I structures. Non-linear a
stress-strain and damping relationships for soil are incorporated in the finite element analysis of soil-structure interaction.
The applicant will perform approximately sixteen cases of soil-structure interaction studies using finite element method for generating GESSAR seismic design envelope by early 1975. These cases will consitute a parametric study to verify their design envelope. The cases will cover a soil depth range of 75 ft. to 1200 ft., a shear wave velocity range of 690 fps to 3000 fps, a Poisson's ratio range of 0.30 to 0.38, a constant groundwater level at a depth of 30 ft.,
and a constant embedment depth of 40 ft.
For each of the 16 cases mentioned above, the sequence of analy-tical operations involved will be, (a) deconvolution analysis to obtain motion at base of soil profile, (b) one dimensional finite element analysis to verify free field motion and to establish maximum i
1 l
l I
l 3-20 !
. i element size, (c) establishment of a full finite element mesh to l
represent the soil surrounding and underlying major Category I struc-
' tures, (d) establishment of an appropriate model of the structure and combination with the finite element mesh to obtain the required soil-structure representation and (e) evaluation of appropriate !
response spectra and seismic design envelope spectra.
We conclude that the seismic analysis methods and procedures pro- .
1 posed by the applicant provide an acceptable basis for system and i subsystem seismic design.
3'. 7. 4 " Seismic Instrumentation Program The type, number, location and utilization of strong motion accelerographs to record seismic events and to provide data on the frequency, amplitude and phase relationship of the seismic response of the containment structure corresponds to the provisions of Regulatory Guide 1.12 (April 1974).
Supporting instrumentation will be installed on Category I struc-tures, systems and components to provide data for the veri-fication of the seismic responses determined analytically for such Category I items. {
We conclude that the seismic instru:acntation Program proposed by the applicant is acceptable.
1
1 3-21 3.8 Design of Seismic Category 1 Structures 3.8.1 Concrete Containment This section is not applicable to the GESSAR application.
3.8.2 Steel Containment The reactor coelant system will be housed within a free-standing steel cylindrical shell topped with a hemi-ellipsoidal dome and fixed at its bottom into a concrete mat covered with a liner plate. The steel containment will be enclosed by a reinforced concrete shield building. The containment will utilize the Mark III pressure suppression a
system which will be utilized to limit the posteLOCA containment pressure and temperature transients, i
The steel containment including all its penetrations vill be designed, analyzed, fabricated, constructed, inspected and tested in strict accordance with the rules of Subsection NE of the ASME Boiler and Pressure Vessel Code Section III, Division 1, and, in general, as augmented by Regulatory Guide 1.57, " Design Limits and Loading Combinations i for Metal Primary React'or Containment System Components."
i The containment will be designed for all the various load combina- )i tions that are censidered credibic, including appropriate combinations !
of accident and seismic loads. In addition, the containment will be designed to withstand a post-LOCA flooded condition to approximately 7'-0" above the cop of the reactor core and in conjunction with an i
I
__._________.___________W
3-22 l i
Operating Basis Earthquake. Such a flooding condition may be required to recover the fuel in the reactor af ter a LOCA.
Regulatory Guide 1.57 states that normal design limits should be used whenever the containment is subjected to'the concurrent loadings that result from flooding of containment for accident recovery and 1 the vibratory motion of 50 percent of the SSE. General Electric's design does not meet this criterion since the containment is not !
designed to meet normal limits for this loading combination but rather they meet a limit that is between the emergency and faulted limits.
'We have deternined that this deviation from Reg. Guide 1.57 is acceptable. Our basis for this determination is that this loading
~i combination will result in stresses that are below the yield condition cf the containment material and therefore the contain-ment would maintain its integrity should an OBE occur during post
, accident recovery.
The materials that will be used in the construction of the contain-ment will meet the requirements of Article NE-2000 of Subsection NE of the ASME Section III Code. The bottom region of the containment that will be submerged in the suppression pool will not be coated except for a narrow band at the water line. At the staff's request, the applicant committed to an inservice inspection program to detect any L corrocion of the steel shell, particularly pitting, and execute appro- j priate corrective measures. l 1
1
I
{
3-23 i
i Af ter the completion of construction and prior to operation the containment will be subjected to 'a structural proof test at 1.15 times the design pressure.' We conclude that'the criteria that will be used in the analysis,- design and construction of the steel- containment structure, to accout for the loadings and conditions that are anticipated to be experienced by the' ,
structure during its service lifetime, are in conformance with established criteria, and with codes, standards, and specifications acceptable to i the Regulatory staff.
1 1
The use of these criteria as defined by applicabic codes, standards,
, "and ' specifications; the loads and loading combinations; the design and' l
l analysis procedures; the structural acceptance criteria; the materials, l
! ; quality control and special construction techniques; and the testing and inservice surveillance requirements, provide reasonable assurance that, in the event of earthquakes and various postulated accidents occurring within and outside the containment, the containment structure will withstand the specified conditions without impairment of its structural integrity or safety function. A Category I concrete shield building will protect the containment from the effects of wind and tornadoes and various postulated accidents occurring outside tfie shield building.
Conformance with these criteria constitutes an acceptable basis for
.g satisfying in part the requirements of General Design Criteria 2, 4, 16, and 50.
____________________--___-__--___m
1 3-24 l
3.8.3 Containment Interior Structures The major containment interior structures include the drywell, the reactor pedestal and shield well, the refueling pool and operating floor, and various other intermediate floors.
The drywell will be a reinforced concrete cylindrical structure with a flat roof that is stiffened by two deep girders forming the refueling pool. It will completely enclose the reactor vessel and the recircula-tion system. Its primary function is to divert the steam released during a
,LOCA to the suppression pool. Because of this important function,on which the proper functioning of?the pressure suppression system depends,
. the Regulatory staff has requested that the drywell be treated to a .
I certain extent as a containment structure. Accordingly, the design and analysis procedures and the loads and load combinations will be similar to what is normally used and accepted for concrete containments.
The lower portion of the drywell wall, which forms the suppression pool wall with the horizontal vents emplaced in it, is to be constructed in a fairly novel form. The large number of 2-foot diameter vent pipes passing through this section of the wall makes the normal reinforced concrete construction method impractical. Instead, this portion of the pressure-retaining wall is proposed as an unreinforced composite section, with external steel plates carrying part of the circumferential and longitudinal stresses.
I
3-25 1 i
l 1
We are requiring that the drywell be subjected to a structural i
proof test at or above the design pressure that will verify the capa- ,
i bility of the completed vessel to withstand the me.ximum design pressure. !
The applicant has not yet committed to such a test. He will report en the resolution of this item in a supplement to the SER.
Guard pipes are provided for all RCPB pipes of a size whose failure could overprigsurize containment. These guard pipes will be designed, constructed and tested in accordance uith Subsection NE of the ASME Section III Code, The routing of safety relief valve discharge pipes to the suppression pool and the pressure loads generated by the operation of one or more "S/R -ralves have not been specified at this time. Resolution of this item will be reported in a Supplement to the SER.
i
(
The floors that are located above the suppression pool in the annulus between the drywell and the containment will be designed to resist the loads imposed by pool swell effects during a LOCA. As a result of the Mark III test program and relatei analysec, the RWCU pumps i:ere relocated from the annulus region to the auxiliary building, and structural modification.
will be made to accommodate the increased impact loading. For further discussion,see Section 6 of this SER.
The other interior structures will also be designed for appropriate load combinations that are acceptable to the Regulatory i
staff. These loads will include appropriate combinations of normal
3-26 operating loads, seismic loads, the loss-of-coolant accident (and other accidents involving high energy pipe ruptures) including temperature, 2 pressure, jet impingement, pipe whip and pipe rupture react. ion forces as discussed in Section 3.8.3.1.3 of GESSAR.
We conclude that the criteria that will be used in the design, analysis and construction of the containment internal structures, to account for anticipated loadings and postulated conditions that may be imposed upon the structures during their service lifetime, are in con-l j formance with established criteria, codes, standards and specifications that at acceptable to the Regulatory staff, a
The use of these criteria as defined by applicable codes, standards and specifications; the loads and loading combinations; the design and analysis procedures; the structural acceptance criteria, the materials, Sh quality control and special construction techniques; and the testing and inservice surveillance requirements, provide reasonable assurance that, in the event of earthquakes and various postulated accidents occurring 4
within the containment, the interior structures will withstand the specified design conditions without impairment of either structural integrity or the performance of required safety functions. Conformance with these criteria, codes, specifications, and standards constitutes an acceptable basis for satisfying in part the requirements of General Design Criteria 2 and 4.
3.8.4 Other Category I Structures Category I structures other than containment and its internal struc-tures that are included in the Nuclear Island will be built from 1
- l. l 3-27 structural steel and concrete. The structural components consist of slabs, walls, beams and columns. The design methods for concrete will follow those specified in the ACI-318 Code, and for steel will follow the AISC specifications with appropriate modifications requested by the staff (as deceribed in Attachment II to Section 3 of our November 1, 1973 letter to GE) to account for loading conditions peculiar to nuclear power plan-We conclude that the criteria that will.be used in the analysis, design and construction of all the Nuclear Island Category I structures sto account for anticipated loadings and postulated conditions that may l be imposed upon each structure during its service lifetime, are in i conformance with established criteria, coder, standards, and specifica-tions that are acceptable to the Regulatory staff. ;
1
'The use of these criteria as defined by applicable codes, standards )
I and specifications; the loads and loading combinations; the design and ;
}
analysis procedures; the structural acceptance criteria; the materials, !
quality control and special construction technigdes; and the testing )
and inservice surveillance requirements, provide reasonable assurance that, in the event of winds, floods, tornadoes, earthquakes and various postulated accidents occurring within the structures, the structures will withstand the specified design conditions without impairment of either the structural integrity or the performance of required safety functions. Conformance with these criteria, codes, specifications, and standards constitutes an acceptable basis for satisfying in part
( the requirements of General Design Criteria 2 and 4.
l 4
i I
3-28 3.8.5 Foundations ;
The foundation of the containment will be a concrete mat. It will be analyzed to determine the effects of the various combinations of loads expected during the life of the plant. The analysis will take into account bending moments, shear, and soil pressure for a plate on 1 an elastic foundation. The containment foundation will be designed and constructed in accordance with the rules of the proposed ACI/ASME Code for " Concrete Vessels and Containments" with certain modifications requested jbytheStaff. Foundations of other Category I structures, likewise, are reinforced concrete mats. Such foundations will be designed in accordance with the ACI-318 Code.
We conclude that the criteria that will be used in the analysis, i l
design and construction of all the Nuclear Island Category I foundations to account for anticipated loadings and postulated conditions that may ,
be imposed upon each foundation during its service lifetime, are in conformance with established criteria, industry codes. standards, and specific tions that are acceptable to the Regulatory staff.
The use of these criteria as defined by applicable codes, standards and specifications, the loads and loading combinations, the design and analysis procedures, the structural acceptance criteria, the materials, quality control and special construction techniques, and the testing and inservice surveillance requirements, provide reasonable assurance that, in the event of winds, tornadoes, earthquakes and various postu-lated events, Category I foundations will withstand the specified l
l
____ ___ ~
4 3-29 design conditions without impairment of structural integrity and stability or the performance of the required safety functions. Con-formance with these criteria, codes, specifications, and' standards constitutes an acceptable basis for satisfying in part the requirements of General Design Criteria 2 and 4.
3.9' Mechanical Systems and components
-3.9.l ' Dynamic System Analysis and Testing 3.9.1.1 Piping Vibration Operational Test Program GE has stated in GESSAR. that preoperational piping vibration tests a
will be conducted on the main steam line and the recirculation: system.
,,I-In response to our concern, GE has committed to perform preoperational vibration tests on all ASME Class 1 and 2 piping. With this require-ment met, the preoperational vibration test. program which will be conducted during startup and initia'l operating conditions on all safety related ASME Class 1 and 2 systems, restraints, components and supports is an acceptable program. The tests will provide adequete l assurance that the piping and piping restraints of the system have been designed to withstand vibrational
- dynamic effects due to valve closures, pump trips, and other operating modes associated with the design operational transients. The planned tests will develop loads similar to those experienced during reactor. operation. Compliance with this test program constitutes an acceptable basis, in partial ful-fillment of the requirement of AEC General Design Cr1terion 15.
t i
3-30 t
3.9.1.2 Seismic Qualification of Mechanical Equipment i
Proper functioning of safety-related mechanical equipment is essen-tial to assure the capability of such equipment to perform protective actions in the event of a safe shutdown earthquake (SSE) . The dynamic testing and analysis procedures wh'ich will be implemented to confirm that all Category I mechanical equipment will function during and after an carthquake of magnitudo up to and including the SSE, and that all equipment support structures are adequately designed to withstand seismic disturbances,,are acceptable.
4 Subjecting the equipment ant its supports to these dynamic testing and analysis procedures provides reasonable assurance that in the I
event of an parthquake at the qite, the Category I mechanical equipment, as identified in GESSAR, will continue to function during and after a seismic event, and the combined loading imposed on the equipment and its supports vill not exceed applicable code allowable design stress 1
and strain limits. Limiting the stresses of the supports under such loading combinations provides an acceptabic basis for the design associ-ated uith seismic events, as well as operational vibratory loading conditions without gross loss of structural integrity.
Impicmentation of these dynamic testing and analysis procedures con-8 stitutes an acceptable basis for satisfying the applicable requirements of 10 CFR 50, Appendix A, General Design Criterion 2.
i
r t
t 3-31 .
1 I
~
. 3.9.1.3. Preoperational Vibration Assurance Program for Reactor Internals With regard to flow-induced . vibration testing of reactor internals, l
. the applicant has. stated in GESSAR that the first BWR/6 plant of each i size will be considered a prototype design and will be instrumented 1 and subjected to both cold and hot two-phase flow testing to demon-strate that flow-induced vibrations similar to those expected during operation will not cause damage. The BUR /6 plants currently-scheduled for prototype testing are River Bend (size 218), Perry (size 238) and l Grand Gulf (size'251). Specific predictions and acceptance criteria I will be supplied at the operating license rev. leu (FSAR) stage of each of the citcd :1 i ants.
l i .
The preoperational vibration assurance program as planned for the-reactor internals provides an acceptable basis for verifying the design adequacy of' these internals under test loading conditions that will be comparable to those experienced during operation. The combination of tests, predictive analysis and post-test' inspection provide adequate
~
assurance that the reactor internals may be expected, during their service lifetime, to withstand the flow-induced vibrations of reactor operations without loss of structural integrity. The continued integ-rity of the reactor internals in service is essential to assure the retention of all reactor fuel assemblies in their place as well as to permit unimpaired operation of the control rod assemblies to permit safe reactor operation and shutdown.
The conduct of the preoperational vibration tests constitutes an i
acceptable basis for demonstrating design adequacy of the reactor internals
3-32 in partially fulfilling the requirements of AEC General Design Criteria i
1 and 4 and in conforming with the provisions of Regulatory Guide 1.20,
( " Vibration Measurements on Reactor Internals."
3.9.1.4 Analysis Methods For LOCA Loadings
'To confirm' the structural design adequacy of the reactor internals, including the control ' rod assemblies, the applicant has described t he dynamic analysis of the reactor internals, together with the unbroken piping loops, which will be performed under the combined effects of the postulated occurrence of a loss-of-coolant accident and a safe shutdown a
earthquake and of an SSE and a steam line break. The staff will require a more detailed description of the analysis for the combined effects of an SSE and a steam line brcah.. When this requirement has been satisfied, the dynamic system analysis which will be performed, provides >
an acceptable basis for confirming the structural design adequacy of the j reactor internals and the unbroken piping loops to withstand the combined dynamic effects of the postulated occurrence of a recirculation line break I plus an SSE plus a steam line break. The analysis will provide adequate assurance that the combined stresses and strains in the components of the i
reactor coolant systems and reactor internals, for these faulted conditions, e
i i
I l
b 3-33.
will'not exceed the allowable design stress and strain limits (faulted limits) for the material of construction, and that the resulting deflec-tions or displacements of any structural element of the reactor internals.will not distort the reactor internals geometry to the extent that core cooling may be impaired. The assurance of structural integrity of the reactor internals under a recirculation line break and-a steam line rupture concurrent with the most adverse loading event (SSE) p'rovides added confidence that the design may be expected to withstand a spectrum of lesser pipe breaks and seismic loading events. Comp 11--
"ance with the dynamic system analysis constitutes an acceptable basis for satisfying the requirements of AEC General Design Criteria 2 and'4.
b ASME'Codo Class 2 and 3 Components
.3.9.2 All safety related ASME Code Class 2 and 3 systems, components and equipment will be designed to sustain normal loads, anticipated transients, the Operating Basis Earthquake and the Safe Shutdown Earthquake within design limits which are consistent with those outlined in AEC Regulatory Guide 1.48, " Design Lim'its and Loading Conditions," The staff will require that the applicant define the loading combination for the upset condition as upset plant condition transients plus OBE unless it can be justified by other methods, as suggested by GE (such as a time history analysis), that such a combit ation is not required. When this require-ment has been satisfied, the specified design basis loading combina-tions as applied to the design of the safety-related ASME Code Class 2
l I
a:.
3-34 )
(-
l and 3 pressure-retaining components in systems classified as Category 1 I provide reasonable assurance that in the event an earthquake should !
1 occur at the site or other upset, emergency or faulted plant transients should occur during normal plant operation, the resulting combined stresses imposed on the system components may be expected to remain within the allowable design stress and. strain limits for the materials of' construction. Limiting the stresses under such loading combinations i p'rovides a. conservative basis for the design of the system components j to withstand the most adverse combinations of loading events without i
" gross loss of structural
- integrity. The applicant's design load combina-tions'and associated stress and deformation limits specified i 11 ASME Code Class 2 and 3 components constitute an accepthble bas s for i design in satisfying AEC General Design Criteria 1, 2 and 4 and are l l consistent with recent Regulatory staff positions.
The criteria used in developing the design and mounting of ASME Class 2 and 3 safety and relief valves provides adequate assurance that, under discharging conditions, the resulting stresses will be within the allowable design stress and strain limits for the materials of construction. Limiting the stresses under the loading combinations associated with the actuation of these pressure relief !
devices provides a conservative basis for the design of the system !
components to withstand these loads without loss of structural i
i integrity and impairment of the overpressure protection function.
l i
{
1 l
l I l l
3-35 I l
i
'h The criteria used for the design and installation of ASME Class 2 i i
and 3 overpressure relief devices,as modified by staff requirements, I t
constitute an acceptable design basis in meeting the applicable requirements of AEC General Design Criteria 1, 2, 4, 14, and 15 are consistent with the provisions of Regulatory Guide 1.67, " Installation of Overpressure l l
Protection Devices." I i
3.9.2.4 Comoonent Operability Assurance Program The applicant has provided in GESSAR a program to assure the opera-
.bility of active components which is acceptable to the staff.
Active components are defined as those pumps required to function and j valves required to open or close during or following the specified i
plant condition..
The conduct of the applicant's proposed operability assurance pro-gram will provide adequate assurance of capability of active i
pumps and valves in Seismic Category I systems, including those which may be classified as ASME Code Class 1, 2 and 3 3 to withstand postulated scismic loads in combination with other sig-nificant loads without loss of structural integrity, and to perform -
the " active" function (i.e., pump operation, valve closure or opening) when a safe plant shutdown is to be effected, or the consequences of I an accident are to be mitigated. The specified component operability assurance procedures constitute an acceptable basis for implementing the requirements
,of General Design Criteria 1, 2 and 4 as related to operability of ASME Code Class 1, 2 and 3 active pumps and valves.
. _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ = _ _ _ _ - _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ ,
(' i l
I 3-36 3.10 Seismic Qualification of Category I Instrumentation and Electrical Equipment )
Proper functioning of Category I instrumentation and electrical equipment is essential to assure the capability of such equipment to j initiate protective actions in the event of a safe shutdown earthquake (SSE) including, for example, operation of engineered safety features and standby power systems. The information presented in GESSAR is consistent with the 1974 draft o'f IEEE 344 " Seismic Qualification pf Class 1 Electric Equipment for Nuclear Power Generating Stations."
TheseismicqualificationtAstingprogramtobeimplementedfor Seismic Category I instrumentation and electrical equipment will provide adequate assurance that such eq'uipment may be expected to function l
properly during the excitation from vibratory forces imposed by the '
safe shutdown earthquake and under the conditions of post-accident operation. This program will constitute an acceptable basis for satisfying the applicab,le requirements of AEC General Design Criterion 2.
I
[
1 i
4-1.
4.0 REACTOR
- 4.1~ General The~ nuclear steam supply system includes a General Electric Company (GE)-
boiling water . reactor (BWR) which generates steam for direct use in the steam-driven turbine generator. The design of the General Electric
> Standard Safety Analysis Report (GESSAR) reactor is similar to the )
Grand Gulf.Huclear Station Units 1 and 2 and to the Perry Nuclear Power Plant, Units 1 and 2 that have been 'reveiwed by the Regulatory- staff. at '
a the construction permit stage.
The fuel and heat source consists of slightly enriched uranium 1.
dio::ide pellets contained in sealed zirconium alloy tubes of about-one-half inch in diameter. These fuel rods, which are;over twelve feet long, are assembled into fuel assemblies each. consisting of 63 rods in an 8 x 8 array within a square open-ended zirconium channel' box. Seven hundred and thirty-two of these fuel assemblies
- r.
form a roughly cylindrical core.
The 8 x 8 fuel assembly design utilizes smaller rods, and has a lower linear heat generation rate (LHGR) than the 7x7 fuel assemblies of previous BWR designs. The overall peak-to-average design power peaking factor of the 8 x 8 assembly has been reduced as.has the maximum LHGR, the maximum heat flux, the maximum fuel temperature and the maximum post-LOCA clad temperature when compared with the previous 7 x 7 design.
i "I
i r
I
/
4-2 The' core is supported in a domed cylindrical shroud inside the reactor vessel. Stean separators and dryers are mounted on the shroud dome. Two external, motor-driven, constant speed recirculating pumps inject high-velocity water into 20 jet. pumps which are located -
' in the annulus between the shro'ud and the reactor vessel. The high' t
velocity water from the jet nozzles entrains and imparts energy to additional water from the annular region. The combined flow enters the bottom of the reactor core and boils as it passes upward I
through the fuel assemblies.
The steam is separated from the-steam-water mixture which emerges from the core by'the steam separators and dryers. The-steam flows to the turbine-gene,rator through four 26-inch diameter ,
main steam lines. The heated condensate returns to the reactor through two- 24-inch feedwater lines and is injected into 'the annulus between the shroud and the vessel.
Control of the fission reaction within the core is achieved by the movement of neutron absorb 1ng cruciform-shaped control i
rods, and variation of the flow rate through the core, thereby changing the steam fraction and moderator density. Individual hydraulic drives permit the control rods to be axially inserted to any degree desired or to be inserted fully and swiftly upon r
receipt of a trip signal (scram). Core flow rate is varied by the flow control valves in the recirculation lines.
L
(
1- 3 4.2 Mechanical Design 4.2.1 Fuel Mechanical Design The 8 x 8 fuel assembly consists of 63 fuel rods and one unfueled, spacer-capture rod in a square 8 x 8 array within a square channel box. The' rods are spaced and supported at the top and' bottom by stainless steel tic plates. The rods are also held "
in alignment by spacer grids located along the assembly. Cladding.
4 is fully annealed Zircaloy and the rod contains a hydrogen getter.
The fuel pellet is a right circular cylinder whose height to diameter !
l ratio is approximately unity. It is chamfered and undished and -'
'l.
made of uranium dioxide of approximately 94% theoretical density.-
Gado11nia bearing pc11ets are used in the highest enriched rods :
and they are distinguished' from the rest of the fuel rods by means of an extended end plug design. A Zircaloy channel box contains the fuel assembly as a load carrying member and controls coolant '
flow and control rod upvement. The channel wall thickness has increased from the previous design. The benefit of the smaller diameter rod design of the new fuel, is to reduce the thermal performance requirements of the fuel and minimize fuel-pellet mechanical 1nteraction by use of chamfered pellets, reduced 1
pellet lengths and annealed Zircaloy cladding. Some of the main mechanical dimensions and~ parameters are given in Table 4.2.1 of I
r:
this report. l
(
4-4 TABLE 4.2.1 Fuel Assembly Data Overall length, in. 176 Nominal active fuel length, in 148 4
Fuel rod pitch, in 0.640 Space between fuel rods, in. 0.147
, Channel wall thickness, in.
0.120 Fuel bundle heat transfer area, ft 100.3 ii!
t Fuel Rod Data Outside diameter, in. 0.493 Cladding thickness, in. 0.034 Pellet outside diameter, in. 0.416 Fission gas plenum length, in. 12.00 Fellet immersion den-d'cy,:.gr/cc 10.42 i
i
l 4-5 1
Fuel assembly design concerns are directed to maintenance of basic assembly geometry for adequate coolant passage and preser-vation of cladding integrity to contain the fission products within the fuel rod.
In Section 4.2.1.3.5 of GESSAR, GE describes the loadings and design limits of the fuel assembly and cladding. They discussed the engineering design limits in terms of stress, strain, deflection, fatigue life and creep rupture. In addition, analytical methods to be used to demonstrate design adequacy are described. Such raaterial properties as cladding yield and ultimat'e stresses, and other thermal properties are given. We reviewed those design
" bases in detail and found that they provide an acceptable description of design bases for the 8 x 8 fuel assembly. Details of our evaluation k of the 8 x 8 fuel design are included in Appendix D of this report which deals with the 8 x 8 reloads. The only difference between the EWR/6 8x8 fuel and the reload 8 x 8 fuel are that the total active fuel length is 4 in. greater in the BWR/6 fuel and the fission gas plenum length is 0.75 in greater for the BWR/6 rods. These changes are not significant enough to change our general conclusions regarding 8 x 8 fuel given in Appendix C.
Prior to our final approval of the 8 x 8 fuel assembly design, at the FDA stage of our review process, we will need to review the following information: i
- 1) Pressure and temperature capabilities listed on pages 4.2-1 and 4.2-9 of GESSAR should be given in terms of specific values or curves as a function of time.
- 2) The analysis method that will be used to predict the combined ef fects of the LOCA and seismic events on the fuel assembly should be submitted. Specific stress, strain and deflection criteria should be given for this load combination.
1 1
i d
4-6
- 3) The analytical methods of creep buckling should be submitted and should include the creep rupture or creep-fatigue-rupture interaction curve.
- 4) The design limit for instability such as instantaneous static ,
I or dynamic buckling and creep buckling should be given. '
- 5) An analysis method for predicting the deformation of the channel box should be submitted. ,
- 6) A stress limit should be given for the peak stress which deals a with stress concentrations and transient non-linear thermal
~
stresses.
j, 7) A justification for setting a 0.060 inch fuel rod deflection limit should be given, i
I l
1
(
'd 1
\
1 4-7 1 I
General Electric performed mechanical tests which included:
fuel assembly handling and shipping tests, channel box removal and replacement tests, water lug shear tests and fuel assembly. bending stiffness tests. These tests verified the ability of the fuel to i
be handled with no damage.
General Electric has accumulated extensive fuel operating experiences with fuel whose range of design parameters envelopes the 8 x 8 fuel. Although the design of the unfue1.ed spacer-capture
~
, rods 1s new, it is based on experience with similar designs.
Fuel assemblies with eccentrically located spacer capture rods have been successfully operated in the Fumboldt Bay reactor.
..b ,
The methods used by CE to calculate the effects of fuel pellet densification have been previously submitted by GE in Topical Report NEDM-10735, reviewed and accepted by the staff.
To replace the poison curta1n previously used, UO ~Gd 0 rods 2 23 are introduced into a high enrichment assembly.- The thermal conductivity of such rods is slightly lower than that of UO r2ds.
However, the rods are expected to operate at relatively lower power than a UO r d.
2 A different end plug design is used to {
distinguish them from other fuel rods. We have previously reviewed t-the use of UO2 -Gd 230 r ds and found them to be acceptable.
General Electric has plans to do a test of the 8 x 8 design i
spacer grid and spacer-water rod locking arrangement. In addition
(-
they have a fuel surveillance program which is to be conducted on preselected 8 x 8 fuel assemblics during refueling outages.
-_-____________D
1
, 4-8 In the Final Safety Analysis Report, results of the above surveillance program should be discussed. Also, a* stress report should be provided for each component together with safety margin.
A calculation of cladding strain based on an empirical formula together with gross diameter measurement of irradiated rod burst test was submitted as a topical report (NEDO-10505, May 1972).
An updated report as well as re-analysis and evaluation is required as new information becomes available.' We conclude
,that the mechanical design criteria of the fuel are acceptable for GESSAR at the tin:e.
j; 4.2.2 Reactivity Control Systems Reactor power can be controlled either by movement of control rods or variation in reactor coolant recirculation system flow !
rate. The fuel rods will contain full length and partial length gadolinium oxide, a burnable poison, to supplement the moveable control rods in controlling the core reactivity throughout the core life. A standby liquid control system is also provided as e
a backup shutdown system.
Control rods (177 in number) are used to bring the reactor through the full range of power (from shutdown to full power operation), to shape the reactor power distribution, and to compensate for changes in reactivity resulting from fuel burnup. Each control rod drive has separate control and rapid l (
1
4-9 c .u . .
insertion (scram) devices. The drives have a common supply pump (and one parallel spare pump) as the hydraulic pressure source for normal operation and a common discharge volume for scram-r operation.-
A control-rod-ejection accident, to be distinguished from the H
rod drop accident, is precluded by a control rod housing support structure located below the reactor pressure vessel, similar to that installed on the other large General Electric r eactors. This
, , structure limits the distance that a ruptured control rod drive housing could be displaced, so that any resulting nuclear transient would not be sufficient to cause fuel rod failure.
'N i Reactor power can also be controlled through changes in the primary coolant recirculation flow rate. The recirculation flow control system can automatically adjust reactor. power level to station load demand whenever the reactor is operating between approximately 65% and 100% rated power. The recirculation flow control system is designed to allow either manual or automatic control of reactor power. This method'of reactor power control -
has been satisfactorily demonstrated in other reactors.
The Standby Liquid Control System (SLCS) is available to pump sodium pentaborate into the reactor vessel. See Section 9.3.4 of this SER for further discussion of tbc SLCS. This system is designed to bring the reactor to a cold shutdown condition from the full power i i 1
.J l
i o
4-10 t steady-state operating condition at any. time in' core' life,
' independent of the control rod system capabilities. The injection rate of the system is adequate to ' compensate for the effects of xenon decay.
On the basis of our review of the control rod, flow control ,
i and standby liquid control systems design, and the supporting evidence accumulated from operation-of similar systems in other General Electric reactors, we conclude that these systems will
, , meet the functional performance requirements and are acceptable.
The details of the proposed design of the new Rod Pattern Control' g System, which will' allow use of ganged rod motion, have not yet been submitted by GE for' Regulatory staff review. See Section 7 for further discussion of this system.
t f
p b-(
i
4 4-11 l
i 4.3 Nuclear Desir,n The GESSAR BUR /6 reactor core consists of 732 fuel assemblies 1 and 177 control rods arranged as shown in Figure 4.1-1 of GESSAR.
A planar view of the fuel lattice cell is shown in Figure 4.2-3 of GESSAR. The fuel lattice cell consists of four square fuel assemblies and a cruciform shaped control rod. A fuel assembly has an 8x8 square array of rods, 63 of which are fuel rods; the 64th rod is a water-spacer rod. The cruciform shaped control rods contain 76 stainless steel tubes (19 tubes in each wing of the cruciform) filled with vibration compacted boron carbide powder. Moderator / coolant (H y 0) occupies all space not taken
,up by fuel rods, control rods, and structural material. All of the water gaps between fuel assemblies are of the same size. Some of the water gaps, i h ,
. which do not include a control rod, are provided with guide tubes for both fixed and moveable neutron flux detectors. Guide tubes are located in the space near the corners of two adjacent fuel assemblics.
There are a number of noteworthy features of the fuel lattice cell which are applicable to the first fuel cycle. These are: (1) the fuel rods are of four differ'ent uranium-235 enrichments, (2) the average enrichment of the uranium-235 isotope in a fuel bundle is 2.0% by weight, and 0) a number of fuel rods will incorporate axially distributed I gadolinia.
t We lave reviewed and evaluated the design bases for the GESSAR reactor. The design bases consist of both safety design bases and t
power generation design bases. The general requirements of the safety design bases are (1) that sufficient negative reactivity feedback be
(
.provided to prevent fuel damage as a result of abnormal operational transients, (2) that nuclear characteristics as required be exhibited
,f
~
4-12
(
to assure that the reactor has no inherent tendency toward divergent or limit. cycle operation, and (3) that the excess reactivity of the core be ' limited. sufficiently to assure that the reactivity control systems are capable of making the reactor suberitical with the highest worth. control red fully withdrawn The general requirements of the power generation design bases are (1) that sufficient reactivity be provided to reach the desired burnup for full power' operation, (2) that continuous, stable regulation of core excess reactivity be allowed, and (3) that sufficient negative reactivity feedback be provided to facilitate normal maneuvering and control.
a In. addition to the general, safety and power generation bases, the GESSAR reactor. is -designed to meet a number of specific GE design j . bases.
. These are listed 'belony
- 1. The pouer reactivity coefficient must always be negative.
- 2. The moderator void reactivity coefficient must be negative.
- 3. The Doppler reactivity coe.fficient must be negative.
, 4. Control rod operating patterns and withdrawal sequences must be specified so that control rod worths are sufficiently low to L
' prevent damage to the reactor system in the event of a rod drop.
- 5. The maximum control rod withdrawal speed must not be greater than 3.6 in/sec.
6.
9 Control rod withdrawal increments must be limited so that a rod movement of one increment does not result in a reactor period which cannot be controlled by an operator.
- 7. The power generation rate must be controlled so that the linear heat .
't generation rate of 13.4 kW/ft is not exceeded and so that MCPR is not less th'an the operating limit for the plant.
w:- a
", 4-13
- 8. The control rod system must be capable of shutting down the reactor (Kdf < 1.0) at any time with the highest worth rod stuck out.
- 9. Sufficient burnabic poison must be' included in the nuclear design !
to ensure that the shutdown requirements can be met throughout the core life.
- 10. The backup shutdown system must be capable of making the reactor suberitical at a temperature of 20*C. It must be capable of inserting at least 600 ppa of natural boron between a minimum rate of 6 pp / min and a maximum rate of 25 ppm / min.
Based on our review, we conclude that the nucicar design bases for the GESSAR reactor are acceptable, since they are in conformance with GDC 10, 11, 12, 26 and 27.
4.3.1 Power Distribution Ue have reviewed and cvaluated the information presented on I
power distribution. The pcuer distribution is a function not only of the nuclear design, but also the reactor operating state. Consequently, an infinite number of power distributions are possible for a GESSAR reactor. Constraints are placed on the power distribution in order to limit the linear heat generation rate to 13.4 kW/f t and to keep -
the MCPR above the operating limit. Target peaking factors for these design limits are given 'in Table 4.3.1. The operating conditionc are i periodically monitored to ensure compliance with the design limits.
The in-core neutron monitoring system is composed of the Source Range Monitoring (SRll) subsystem, the Intermediate Range Monitoring ,
(IRM) subsystora, the Local Power Range Monitoring (LPRM) subsystem, the Average Power Range Monitoring (APRM) subsystem, and the Traversing In-core Probe (TIP) subsystem. The SPd! range varies from the source range to about 10~ % of full power. The IRM's cover from 10 to 20%
4-14 of full pover. The LPM range varies frora a few percent to 150% of full power. The APPJf's provide a continuous indication of average reactor power from a few percent to 150% of rated reactor power. The APPJI subsysten is baced on a subset of the LP.T! detectors. The T1P subsystem is used to calibrate the LPPJis and to provide detailed axial flux distributions.
A study of power distributions in boiling water reactors is given in Appendix 4A of GESSAR. Appendix 4A, which is still being revieved by the staff, indicates that the GE design methods are capable of adequately representing operating reactor states. The design methods are compared with measured data for both gross and local power distribu-
"'tions. The effect on pouer distributions of rod patterns, fuel burnup, flou variations, void distribution, xenon, hot and cold reactor t conditions, and load follouing are discussed. The errors and uncertainties associated with the analytical methods are also discussed and have been accounted for in the evaluation of fuel performance by i GE through the process computcr.
We conclude that discuss 4.ons of the power distribution in Sectien 4.3 and in Appendix 4A of CESSAR are acceptable provided that questions and concerns arising from the staff revicu of Appendix 4A are satisfactorily resolved prior to issuance of a PDA. These questions are directed toward the statistical analysis of reactor data in establishing and
, accounting for errors and uncertainties. These questions are being addressed by GE in Topical Report NED0-20340 and resolution of our concerns will be accomplished as a part of our review of that topical report. We further conclude that the information presented concerning the monitoring of power distributions is acceptable.
4-15 4.3.2 Reactivity Coefficients We have reviewed and evaluated the information presented on the reactivity coefficients. The most important reactivity coefficients which determine the stability and dynamic behavior of the GESSAR reacter are the Doppler reactivity coefficient, the moderator void reactivity coefficient, and the moderator temperature reactivity coefficient.
The power reactivity coefficient, which is associated with stability to power oscillations due to xenon and other causes, is a function of the Doppler and moderator void reactivity coefficients.
The Doppler reactivity coefficient is a reactivity change associated
, with the Doppler broadening of absorption resonances and is caused by changes in temperature. The Doppler reactivity coefficient is negative for the GESSAR reactor. The ausa.uce magnitude of the coefficient increases with both increasing moderator temperature and increasing void fraction because the resonance escape probability is inversely proportional to the water to fuel ratio. The Doppler reactivity coefficient also becomes more negative as a function of fuel burnup due to the bulldup of plutonium isotopes. Values of the Doppler reactivity coefficient are gf;c_r. -in Tcbic 4.3.1 of this report. In various transient analyses, the Doppler reactivity coefficient is taken to be 0.126c/F and is multiplied by a design conservatism factor of 0.9.
The GESSAR reactor has a large negative moderator void coefficient of reactivity and a modcrator temperature coefficient of reactivity which is much smaller in magnitude. These ceafficients are obtained
( from partial derivatives of the infinite multiplication factor,
4-16 i
neutron 1cchage, and control fraction
- with respect to the variables of temperature or void content with the reactor near critical. Of the two, the moderator temperature coefficient is less significant 'and plays a role only near the inlet region of a hot operating reactor where the void content is smallest. This coefficient may become slightly positive near the end of the fuel cycle. The strong modcrator void coefficient of reactivity, on the other hand, gives the GESSAR reactor a nunber of inportant characteristics such as (1) the capability of using coolant flow control for load follouing, (2) the inherent ability to self-flatten the radial power distribution, and (3) stability
" to xenon. induced spatial power oscillations. Values of the void reactivity coefficient are given in Table 4.3.1 of this report. In various transient analyses, the moderator void coefficient is taken to be -9c/% void and is multiplied by a design conservatism factor ranging in value from 1.2 to 1.4.
We have revieued this information and conclude that the discussion in GCSSt.R of the reactivity coefficients is acceptabic. We find that the i important prompt (Doppler) and void reactivity coefficients are negative i
throughout a fuel cycle. We further conclude that the absolute magnitudes I i
of these coefficients are sufficiently large to ensure the stability of the !
GESSAR reactor during power operation. l 4.3.3 Control Requirements and Control We have reviewed and evaluated the information presented for the GESSAR reactor on control requirements and control. The excess reactivity designed into the initial core is controlled by a control i
- The control fraction is defined es the ratio of the leno.th of control rods inserted into the reactor to Lhe total inserted length of all of the control rods .
~
1 4-17 rod system supplemented by the use of a burnable poison, gadolinia, in a number of fuel rods. The gadolinia is uniformly distributed in a UO 2 fuel pellet but hcs an axial distribution within a fuel rod.
The reactor is designed to permit the energy extraction of 12,000 to 19,000 INd/T everaged over the initial core loading and depending on the initial uranium enrichment. The excess reactivity is needed to compensate for reactivity losses due to moderator heating and boiling, fuel temperature incrcasec, equilibrium and peak xenon, sanarium poisoning, fuel depiction, and'other low cross section fission product poisons.
The control rods provide i number of important, operating functions.
They are a means for (1) rapidly decreasing the core reactivity during i
- areactortripbybeingdrive$intothecore, (2) bringing the reactor into the power operating range from either cold or hot shutdown conditions by planned rod citi.draval, (3) compensating for fuel depletion by planned rod withdraual, and (4) shaping the pouer distribution by selectivo movement. The control rods are capabic of shutting doun the mactor (Eggf < 1.0) throughout the entire first fuel cycle for -
the most limiting condition, that is, for the reactor at 20*C and f
for the highest worth control rod stuch out. The uncertainty associated with the calculation of the shutdotm margin was estimated by GE to be about
)
O.007 AK by comparing calculations with measurements on critical cores l having no exposure. l Control rod withdrawal sequences are selected prior to operation in order to optimize core performance and to achieve low individual rod vorths. The maximum controlled rate of reactivity addition during )
l l
l
4-18 startup is 0.0011 AK/sec. This value is based on the withdrawal of ,
an in-sequence rod assuming a total rod worth of 0.0100 an, a peak incremental rod worth of 0.00033 AK/in, and a maximum rod speed of 3.6 in/sec. Reactivity addition rates are considerably reduced at hot operating conditions from those under startup conditions due to the effects of void forriation and redistribution as a rod is withdrawn.
The control of the reactor is not only dependent upon the movement of control rods but also toon changes which occur in various system parameters. Eccause the pressure changes caused by turbine throttle operation bring about reactor power changes in a direction opposite to changes in reactor pressure, the reactor is operated as a constant pressure device. The plant output is increased or decreased by changing the reactor circulating vater flou and/or noving the control rods.
As indicated previously, reactor startup from cold or hot conditions is accomplished by withdrawing control rods and keeping the recirculating unter flow at a fixed value. The reactivity differences betwaen the hot standby condition- (5% pouer, 30% flow), as defined by GL, and the cold critical condition are 0.069 on and 0.041 LK for beginning and end of cycle, respectively. These reactivity differences include temperature, void fraction, and xenon changes. By adjusting the recirculating water flow, the reactor power can be varied over approximately 35% of the power range. The power change produced by varying the recirculating water flow is nearly uniforn and is based j l
on curves developed during the reactor startup phase which correlate reactor power and flow for various control rod patterns. Control rod changes may also be made in the power range in conjunction with changes 1
4-19 in the recirculating water flow, however,' load following -is usually accomplished by varying recirculating water flow. Spatial power disturbances, such as those caused by xenon redistribution, present no special control problem to the GESSAR reactor. The large negative power coefficient provides strong inherent damping of such disturbances or oscillations.
The GESSAR EITR/6 incorporates a standby liquid control system to satisfy the requirements of GDC 26. This system is capable of injecting a natural boron solution at the rate of 6 to 25 ppa / min and can bring the system coolant to a concentration of at least 600 ppm. Based on
" the reactivity worth of the boron, this liquid control system, independently i
of any control rod action, is capabic of shutting down the reactor.from I full power throughout the fuel cycle.
He conclude that the discus,Sicn in GESSAR of the control requirements l and control is acceptable. We find that there is sufficient shutdown margin throughout the fuel cycle. We agree with the applicant that spatial power disturbances will be strongly dauped by the large negative i
power coefficient. We conclude that power changes by control rod move-ment and/or changes in recircu1?tfor water flou can be made in an acceptable manner uith respect to effects on the power distribution.
We further conclude that adequate control of the excess reactivity
, exists throughout the fuel cycle. Finally, we conclude that a second shutdown control system requirement is met by the standby liquid control system.
4.3.4 control nod Patterns and Reactivity Worths
(
We have reviewed and evaluated the information presented for the GESSAR El!R/6 on control rod patterns and reactivity worths. We find
4-20 f
specified rod withdrawal sequences are designed to limit rod worth so that
, the ejection of any control rod from the fully inserted position resuirs in a peak fuel enthalpy of not more than 280 cal /g. The selected rod patterns
, at any time will satisfy this requirement on the peak fuel enthalpy if.the control rod worth is no more than 0.01 4K even if the rad ejection velocity reaches its_ maximum value of 2.79 ft/sec.
As contrasted to other power producing reactors, the rod withdrawal I
sequences for a boiling water reactor are complex. In the startup range, I the control rods are withdrawn to 50% control dansity leaving a checkerboard pattern. ,0nce a control rod has been selected for withdrawal in the i startup range, it is withdrawn from its fully inserted to fully withdrawn position. The maximum in-sequence rod worth always occurs when the first control rod of an in-sequence group is withdrawn. The maximum out-of-sequence control rod worth will occur as follows: (1) all the. control rods of an in-sequence group have been withdrawn, (2) a single rod from the next in-sequence group is withdrawn, and (3) the operator makes a single error by withdrawing the out-of-sequence control adjacent to the in-sequence control rod withdrawn in step 2. The withdrawal of control rods during startup are performed in conjunction with permissives from _ the Rod Pattern Control System (RpCS).
l l
l
4-21 In the power range, once a checkerboard control rod configuration has been achieved, the concept of in-sequence and out-of-sequence control rods is no longer meaningful since all interior control rods will have.
- approximately the same reactivity worth. The worth of an interior control rod is about 1.5% Ak/k in the hot operating state; however the amount of 1'
reactivity which can be added due to a dropping control rod is restricted.
since only partial withdrawal of all the remaining rods in bands occurs.
Control. rod withdrawals in the power range are also restricted to limit the total power peaking factor. Control rod patterns are varied from ti=e to time to maintain uniform burnup in each fuel assembly.. In the power i
range the worst singic operator error is defined as the selection and full !
withdrawal of the maximum worth control rod. This results in two ways of inserting high reactivity into the reactor. The first way is by the high I
worth rod itscif and the second way is b,y an adjacent rod drive be'ing completely withdrawn but with its control blades in the fully inserted position.
In the startup range the maximum in-sequence and out-of-sequence control rod worths are computed by means of full core, three group, XY diffusion caicuIctions. Homogenized cross sections are used for each fuel hindle. l These cross sections are generated by using the GE standard lattice design methods for the controlled or uncontrolled fuel bundle. The effects of the axially distributed gadolinia are included in the XY diffusion calculations by using average cross sections and axial l bucklings obtained from a one-dimensional, three group, axial diffusion calculations.
c l
l
4-22 In the power range, the control rod calculations are affected by the formation of steam voids in the moderator. The maximum control rod worth I is calculated by means of three-dimensional XYZ diffusion theory for a control rod fully inserted or fully withdrawn for a constant void distri-bution.
The initial void distribution is obtained from a three-dimensional coupled nuclear-thermal huraulics calculation with the maximum worth out- !
of-sequence control rod fully inserted.
The information presented on the scram reactivity function is co nfusing.
Section 4.3.3.2.2 indicates that the scram reactivity function used to terminate abnormal operational transients is shown in Figure 15.1.1-1 while Section 4.3.3.2.6 indicates that the curve in Figure 4.3-3 for end of equilibrium cycle conditions is used for analyzing operating plant transients. Thus, we are uncertain which scram reactivity function has been used in the i
analyses of various trcnsients and accidents described fu Chapter 15.
If the scram reactivity function actually used is one of GE's generic curves, it should be clearly identified and presented in such a manner that reactivity ($) versus control fraction and scram time may be easily 'i read over the entire range of values.
We conclude that the information presented in C"SSAR on control rod patterns and reactivity uorths is acceptable. Although the control rod patterns and withdrawal schemes are quire complex, we find that the Rod Pattern Control System and the nuclear instrumentation can limit the worth of a control rod and the power peaking factor. Finally, we conclude that the restrictions on the rod patterns will limit the incremental control rod worth to approximately 0.01 K and that no dropped rod would produce a peak enthalpy of 200 cal /g even if the rod were dropped at 279 ft/sec.
_ _ _ _ _ -__U
- e 4-23 We conclude that the scram reactivity function which will be used in the transient and cceident analpes ne i to be fully discussed and more precicely defined.
4.3.5 Criticality of Fuel Assemblies We have reviewed and evaluated the information presented for the GESSAR BWR/6 on the criticality of fuel asscablies. The criticality analyses are performed assuming a higher-than-normal average fuel enrichment and also assuming that there are no control rods or gadolinia.
For the dry condition, the multiplication factor, K,ff is 0.50. In the fucl handling facilities two fudl bundles give K 0.74, four bundles eff give K % 0.90.
ggf Sixteen to twenty fuel bundles represent a critical
( array.
Procedural controls preicnt personnel from arranging four fuel bundles in a square array. See Section 9.1 of this report for further discussion of fuel criticality.
Uc concle.'.e that the discussion of criticality of fuel assemblics as presented in GESSAR is acceptable. Ue find that the procedural controls outlined are sufficient to prevent K,gg from exceeding 0.90 under normal conditions of fuel handling and storago.
4.3.6 Vessel Irradiation We have reviewed and evaluated the information presented for the GESSAR BWR/6 on vessel irradiation. A one-dimensional, discrete ordinates transport code is used to calculate the neutron fluence at the pressure vessel assuming continuous reactor operation at rated power for 40 years.
~
{
~
4-24 1
A radial power distribution representative of conditions throughout -
the life of the plant was used. Axial power distributions representative of beginning and end of cycle were assumed. The calculated fluence at the pressure vessel for neutrons of energies above one MeV is about 2.4 x 10 18 neutrons /cn2, We conclude that the information presented on vessel irradiation is acceptable.
4.3.7 Analytfen1 Methods Ue have revicued and evaluated the information presented for the
, CESSAR EWa/6 on the analytical methods. The basic calculational procedures used by CE for generating neutron cross sections are part of its so-called Lattice Physier. ;1od el. In this model the many-group fact and resonanco energy crocs sections are computed by a GAM-type of program. The fast energies are treated by multigroup integral collision probabilities to account for geometrical effects in fast fission. Resonance energy cross sections are calculated by using the intermediate resonance approximation uith energy-and-position-dependent Dancoff factors included. The thermal cross sections are !
computed by a THFRHOS-type of program. This program accounts for !
the spatially varying thermal spectrum throughout a fuel bundle.
These calculations are performed for an extensive combination of parameters including fuel enrichment and distribution, fuel and moderator temperatures, burnup, voids, void history, the presence or absence of adjacent control rods, and gadolinia concentration and
, distribution in the fuel rods. As part cf the Lattice Physics Model, l
i
-_____-_2
l l
l 4-25 l
three-group, two-dimensional XY diffusion calculations for one or four fuel bundics are performed. In this way, local fuel rod powers can be calculated as well as single-bundle-or four-bundle (with or without a i control rod present) average cross sections. )
The singic or four bundle averaged neutron cross sections which are obtained from-the Lattice Physics llodel are used in either two-or three-dimensional diffusion calculations. Two-dimensional, XY calculations are usually performed in three-groups at a given axial 4
location to obtain gross pouer distributions, reactivities, and average a three group neutron cross sections for use in one dimensional axial .
calculation. The three-dimensional diffusion calculations use 1.5 energy groups and can couple neutron and thermal hydraulic phenomena.
These three-diacncional calculations are performed using 24 axial nodes and 1 radial node per fuel bundic restiting in about 14,000 to 20,000 spatial nodes; however, at the design stage geometrical synmetry is used to reduce the size of the calculation. This three-dimensional calculation provides t.he best simulation of the GESSAR BUR /6 and yicids gross three-dimensional pouer distributions, void distributions, control rod positions, reactivities, eigenvalues, and also average cross sections for use in the one-dimensional axial calculations.
The one-dhnensional axial calculations are space-time diffusion calculations which are coupled to a single channel thermal-hydraulic mod el.
This axial calculation is used to generate the scram reactivity function for various core operating states. This one-dimensional l
g space-time code has been compared by GE with results obtained ucing the industry standard code, UIGLE.
4-26 I
The Doppler, moderator void, and noderator temperature reactivity coefficients are generated in a rudimentary manner from data obtained from the Lattice Physics Model. The effective delayed neutron fraction and the prompt mode neutron lifetin,e are computed using the one-dimensional space-time code. The power coefficient is obtained by appropriately combining the void and Doppler reactivity coefficients.
The behavior of the GESSAR BWR/6 to any induced power oscillations is discessed in GE Topical Report APED-5652. The effect of spatially varying ..enon concentrations on the stability of the GESSAR BWR/6 is specially discussed in. reference 22 of Section 4.3 of GESSAR. These l
studies show that the GESSAR BWR/6 is stable to any xenon-induced power oscillat1ons because of the damping effect of the large, negative, spatially varying void coefficient.
Appendix 4A of CESSAR gives a considerable amount of information j
on the comparison of calculated local and gross power distributions with measured data. The factors which influence the power distribution are discussed as well as uncertainties in the measurements and calcula-tions. Houever, Section 4.3 of GESSAR does not provide any comparisons of calculations of Rgg with measured data for hot and cold conditions i and with and without equilibrium xenon and samarium present. Comparison with experimental data of calculated control rod worths in the cold condition !
shutdown margins for various conditions, the reactivity worths of the distributed gadolinia, and reactivity coefficients for various conditions is similarly lacking.
f i
We conclude that the discussion of the analytical methods in GESSAR indicates that they are sta te-of-the-art. However, we believe that the analytical methods need to be more fully described and documented
L i
1 4-27 I
in terms of the equations, numerical techniques, and methods of solution.
We believe that the neutron cross section data base needs to be fully described and documented prior to issuance of a PDA. Furthermore, we believe that the analytical methods need more experimental verification and documentation over as wide a range of boiling water reactor parameters and operating states as possible and must include a discussion and evaluation of the uncertainties involved.
4.3.8 Summary of Evaluation of Nuclear Design Ue believe that our revie7 has estaW ished that sufficient information a has been presented in GESSAR to conclude that the nuclear design and operational boundaries are understood and that the reactor can be expected to meet required limitations over the appropriate range of operation.
In particular, vc believe that suf ficient information has been presented on such reactor characteristics as power distributions, reactivity coefficients, and control for use du steady state limits and in the trancient ard accident analyses described in Chapter 15 and i that the Technical Specifications are compatible with these characteristics. {
llovever, there are a number of areas in the CESSAR nuclear design l
where uc require additional information prior to issuance of the PDA to i complete our review. Specifically: i
- 1. The effect,'if any, on the nuclear design of the major design modifications which were incorporated into Chapter 7.
- 2. The scram reactivity function which is used in the transient and accident analyses in Chapter 15.0 needs to be identified and k
- more fully described.
i 4-28 i .
l
- 3. Questions and concerns related to uncertaintica and errors arising from the staff review of the power distribution study given in i Appendix 4A need to be satisfactorily resolved.
- 4. The analytical raethods described in GESSAR need to be raore fully described and documented.
- 5. A comparison of analytical results uith reasured reactor
, , characteristics should be presented along with the associated f I
uncertainties.
a 2".
t I
(
)
- _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ . I
4-29 TABLE 4.3.1 NUCLEAR DATA SUIT!ARY GESSAR ;
Design Peaking Factor .
Maxiraum Fuel Bundic to Average Fuel Bundic 1.40 Axial Peak-to-Average 1.40.
Local Peak-to-Average 1.13 Total Peak-to-Average 2.22 a
Water-to-Fuel Volume Ratio .- 2.50 Uranier Weight per Bundle (1b) 415 Maxiuuu Core Reactivity, All Rods in (Keff) <0.965 j Maximum Core Reactivity, Strongest Rod Out (Keff) <0.99 Reactivity of Movable Control Rods, Cold (4K) 0.17 Range of Reactivity Coefficients Fuel Doppler Coefficient (6h/h/'P) -1.2 to
-1.3 X 10 -5 k
Moderator Void Coefficient (6h/k/% void) -1.0 t
-3
-1.6 x 10 l 1
l'
4-30 4.4 Thermal and Hydraulic Design The thermal and. hydraulic characteristics of the GESSAR reactor are similar to those for other BWR/6 reactors such as the Grand Gulf Nuc1. ear Station, Units 1 and 2 and the Perry Nuclear Power Plant, Units 1 and 2 which have already been reviewed at the CP stage by the Regulatory staff. In addition, the GESSAR reactor thermal and hydraulic characteristics are similar to those for previous BWR designs, such as LaSalle County 1 and 2, Bailly Unit 1, and Zimmer Unit 1 nuclear facilities,
, except that the fuel assemblies have an 8 x 8 rather than a 7 x 7 array of rods.
The core thermal and hydraulic design bases are formulated to limit the local power density and coolant flow within the core to values such that the fuel damage limits are not exceeded during normal operation or operational transients. One damage limit io the critical heat flux. The present critical heat flux limits are calculated using the correlation reported in the GE topical report
}
APED-5386, " Design Basis for Critical Heat Flux Conditions in l l
Boiling Water Reactor," issued in 1966. This correlation, known as f
the Hench-Levy correlation, is based on experimental data taken over the range of conditions representative of BWR's. The correlation l was formulated as a lover limit line to the then existing rod bundle critical heat flux (CHF) data. The minimum critical heat j l
flux ratio (MCHFR) is defined as the ratio of the critical heat j
(
l
_ __ ______o
1 4-31 flux correlation value at the corresponding fluid conditions to the actual maximum calculated heat flux occurring at a given point in the fuel assembly at any time during operation, including reactor anticipated transients. A MCHFR of 1.0 conservatively predicts that cooling of the fuel rod is maintained in the nucleate boiling heat transfer regime.
In addition, GE has nearly completed an extensive series of
. critical heat flux tests in the ATLAS test loop, on full-scale 8 x 8 heater bundles with varying inlet conditions, and axial and radial power distributions which are representative of a expected conditions in a BWR. These tests have provided the foundation for developing a new critical heat flux correlation called GEy.L (General Electric Critical Quality X - Boiling Length) which GE proposes as a replacement for the Hench-Levy correlation. The basic fona of GEXL is identical to the well verified CISE (Italy) correlation. In contrast to the Hench-Levy correlation, which is a lover limit line to the data, GSXL is a best fit correlation to the ATLAS data. GEXL is used to determine the bundic pousr (critical power) at which boiling transition will occur. It is proposed that, when approved by the AEC, GEXL replace the presently used Hench-Levy correlation.
A new thermal design method (GETAB, General Electric Thermal Analysis Basia), which uses GEXL and appropriate design parameters to determine the maximum power capability of a fuel assembly during normal operation and abnormal operational transients and accident I
conditions, is also proposed. The figure of merit chosen for reactor design and operation is the critical power ratio, defined as the ratio of the critical bundle power to the operating bundle power. '
1
)
J 4-32 f
I We have reviewed the GEXL, GETAB, and Hench-Levy correlations as well as the rod bundic data and the results of our review are i given in Appendix E of this report. General Electric Company has begun i applying GETAB to their operating plants. They have instructed i
the owners to establish operating limits by either the Hench-Levy correlation or GETAD, choosing the method that provides the more conservative results. GE filed an auendment to GESSAR ,
at the end of August, 1974, that discusses the application of GETAB to GESSAR. We presently feel that if a GESSAR reactor a
were currently operating, thermal margins should be predicted on the basis of the method (i.e. , either Hench-Levy or GETAD) which yicids the more conservative recults. Theref ore, use of a
the Ucuch-Levy correlation for the GESSAR reactor vould be acceptable. !
However, upon submittal of the evaluation of GESSAR using GETAB, we will re-examine the acceptability of current critical heat flux models. i The current GE design basis (Hench-Levy) for normal operation is that the MCnrn calculated for any point is greater than 1.9 during normal operation and greater than 1:0 during anticipated transients.
These limits provide considerabic cargin between expected conditions t
, and those required to cause fuel clad damage since the critical heat flux correlation presented in A?ED-5286 is conservatively based on a limit line drawn below'nearly all of the available
(
L- _ _ _ _ _ _ . _ _ _ _ . _
l 1
4-33 J l i 1 0 1
1 1
experimental data points. The maximum linear heat generation rate 1.
reached during normal rated power operation is not expected to cxceed 13.4 kW/ft, corresponding to a MCHFR of 1.9. Analysis of 1
anticipated operational transients shows that the lowest MCHFR, a value of 1.3, occurs following a loss of feedwater heater. A MCHFR of 1.0 would occur at 16.0 kW/ft. Center line molting begins at 21 to 22 kW/ft.
In GETAE, the uncertainties associated with the parameters
,affecting steady state bundle power are treated statistically in order to satisfy the criterion that, during a transient, 99.9%
of the rods in the core will not experience boiling transition.
I For example, in a typical reactor, a MCPR (miniraum critical power ratio), of 1.04 would meet the 99.9% criterion. To accommodate transients, which are ec1culated from conservative, non-statistical assumptions, a resulting LMCPR is added to the base MCPR in order to arrive at the required steady state MCPR uhich might typically be 1.2.
The scope of our thermal-hydraulic design review included the l
design criteria and thermal-hydraulic performance. The applicant's '
thermal-hydraulic analyses vere performed using analytical methods and correlations that have been previously reviewed by the staff and found acceptable. Differences between the proposed core design l (and criteria) and those designs and criteria that have been (
L
4-34 I
previously revieued and found acceptable by the staff were reviewed.
We found that all such differences were satisfactorily justified by the applicant.
The staff concludes that the thertnal-hydraulic desiga of the core conforms to the Commission's regulations and to applicable Regulatory Guides and staff technical positions and is considered acceptable for the PDA. In Section 4.4.3.5 of GESSAR, CE presents typical values of st. ability and hydrodynamic performance and references a
calculations that predate introduction of the BIm/6 dee.ign. CE should update the stability analyses prior to submittal of the first BWR/6 FSAR.
O l
I e
5-1 ,
5.0 REACTOR COOLMIT SYSTDI 5.1 Summary Description The principal components of tite reactor coolant system are the reactor pressure vessel, the reactor recirculation system, the main steam and feedwater lines, and the pressure relief systen. These items form the major components of the reactor coolant pressure boundary (RCPB). The pressure boundary also contains portions of the Reactor Core Isolation Cooling System, the Residual Heat Removal System and the Reactor Water Cleanup System. Portions of these systems as well as other piping that extend from the reactor vessel out to the second cuter-mostisolationvalveareconsid$redwithinthereactorcoolantpressure boundary.
5.2 Integrity of Reactor Coolant Pressure Eoundary 5.2.1 Design of Reactor Coolant Pressure Boundary Components The staff will require that the applicant define the loading combina-tion for the upset condition as upset plant condition transients plus OBE unless GE can justify that such a combination is not required. GE feels they can do this by employing a time history analysis. When this requirement has been satisfied, either by compliance to our position or justification of their position, the design loading combinations specified for ASME Code Class 1 RCPB components will have been appro-priately categorized with respect to the plant conditions identified as Normal, Upset, Emergency or Faulted. The design limits proposed by the
(
applicant for these plant conditions are consistent with the criteria recommended in AEC Regulatory Guide 1.48, " Design Limits and Loading
l 5-2 Combinations for Seismic Category I Fluid System Components." Use of ;
the criteria recommended in AEC Regulatory Guide 1.48 for the design of the RCPB components will provide reasonabic assurance that, in the event an earthquake should occur at the site, or other system upset, emergency or faulted conditions should develop, the resulting combined stresses imposed on the system components will not exceed the allowable design stresses and strain limits for the materia]s of construction.
Limiting the stresses and strains under such 1c.ading combinations pro-av ides a basis for the desimi of the system components for the most adverse loadings postulated to occur during the service lifetime without loss of the system's structural integrity. The design load combinations and associated stress and deformation limits specified for ASME Code Class 1 components constitute an acceptable basic for design in satis-fying the relcted requirements of AEC General Design Criteria 1, 2 and 4.
5.2.1.1 _C_omp]iance with 10 CFR Part 50, Section 50.55a Components of the reactor coolant pressure boundary as defined in 10 CFR Part 50, Section 50.55n, have been properly identi-fied and classified as AS!1E Section III, Code Class 1 components. These components within the reactor coolant pressure boundary will be con-structed in accordance with the requirements of the applicabic codes and addenda as specified by 10 CFR Part 50, Section 50.55a, Codes and Standards.
i
5-3 The staff concludes that construction of the components of the reactor coolant pressure boundary, in conformance with the Commission's l
regulations, as discussed above, provides reasonable I l
assurance that the resulting quality standards are commensurate with !
the importance of the safety function of the reactor coolant pressure boundary and is considered acceptabic.
5.2.1.2 Applicable Code Cases The specified ASME Code Cases, whose requirements will be applied in athe construction of pressure-retaining ASME Section III, Code Class 1, components within the reactor coolant pressure boundary (Quality Group Classification A), arc acceptable to the that compliance with the requirements of these Code Cases in conformance with the Commission's regulations is expected to result in a component quality level commensurate with the importance of the safety function of the reactor coolant pressure boundary and is considered acceptable.
- 5. 2.1.3 Design of Active Punps .and Valves Sec Section 3.9.2.( of this report.
- 5. 2 .1.4 Loose Parts Monitor occasionally, misec11aneous items such as nuts, bolts, and other small items have become loose parts within reactor coolant systems. In addition to causing operational inconvenience, such loose parts can damage other components within the system or be an indication of undue wear or vibration. For the past few years we have required many
(
l 5-4 applicants to initiate a program, or to participate in an ongoing pro-gram, the objective of which was the development of a functional, loose parts monitoring system within a reasonable period of time.
Recently, prototype loose parts nonitoring systems have been developed and are presently In operation or being installed at several plants.
General Electric Company has stated that they have initiated a long-term program for the purpose of development of a vibration conitoring system for light water reactors. The objective of the program is the 4 development of a system requiring sensors on only the outside surface of the reactor pressure. vessel to provide continual monitoring for the impact and vibration of loose parts during reactoe operation. We have concluded that this is* acceptable for this stage of the GL6SAR review.
5.2.2 Overpressure Protection The pressure relief system prevents overpressurization of the reactor coolant boundary under the most severe operational transients and limits the reactor pressure during normal plant isolation and load rejections so as to prevent opening of the spring safety valves. The valves of the i pressure relief system also are part of the Automatic Depressurization i
System, which is a subsystem of the emergency core cooling system !
described in section 6.3.
The pressure relief system consists of 22 dual purpose safety / relief valves. All are mounted on the main steam lines within the primary containment drywell between the reactor vessel and the isolation valves.
(
l l
i 5-5 l All discharge through piping directly to the suppression pool. The ,
1 valves are all spring-loaded with the set pressures in the range from 1135 to 1205 pcig. At the set pressure of the highest set valve, the valves have a combined capacity equal to 126% of rated steam flow.
The valves are also actuated at relieving set pressures within the range of 1095 to 1135 psig. These valves contain pneumatic actuators and can be operated either by automatic or remote manual controls at any pressere above atmospheric. For overpressure relief, a pressure aswitch ou cach valve initiates the pneumatic actuator at the relieving set pressure. Eight of the valves can be pneumatically actuated by a
'l signal from the Automatic Depressurication System (ADS). Each of these valves is equipped with a pneumatic accumulator and a check valve in the supply line so that the valve can be actuated even if the pneumatic supply faila.
The cbility of the pressure relief system to prevent overpressuriza-tien of the reactor coolant pressure boundary is evaluated assuming that: (1) the plant is operating at design conditions (105% of rated steam flow and a reactor vessel dome pressure of 1045 psig), (2) the most severe operational transient occurs (closure of the main steem line isolat ion valves), (3) the direct scram signal from the . valve ;
1 position suitches fails and scram is effected by the fastest indirect !
scram signal (high neutron flux), and (4) at least one valve is inoperative.
(
1
_.y.._ _ _ . _ . _ _ . _ . . . __ __. _ _ . . _ _ . __ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- /1 <
(
5-6 >
The bachup reactor high neutron flux scran is conservatively. applied as a design basis in determining the required capacity of the pressure relieving dual purpose safety / relief valves. Application of the direct position scratus in the design basis could be used since they qualify as acceptabic pressure protection devices when determining the required i
safety / relief valve capacity of nuclear vessels under the provisions of the AS"I Code. Since the peah vessel pressure is'at least 50 psi less i than the ASME Code allovabic pressure of 1375 psig, at the vessel bottom,
the overpressure protection system is acceptable.
The ability of the syatem to limit pressure during anticipated i operational transients belou the spring set-point of the safety-relief valve is analyzed ase,uming (1) the plant is operating at design conditions (105% rated steam flow and 1045 psig pressure in the steam dome), (2) .
the most severe operational transient occurs, (3) scram is effected by the direct sf p:31 fron the valvas' position suitches and (4) all safety /
{
relief valvce function properly. Locause the analysis of the most severe operational transient, closure of the turbine stop and bypass I
valves, results in a calculatd peak vessel pressure of 1216 psig, the I
overpressure protection is acceptable.
A small fraction of the pressure relief valves at some EhR plants L
have inadvertently opened during certain transients. An evaluation of these i
inadvertent openings indicates that the potential exists for the same mechanim !
to prevent these valves from opening when required. Even though these 1 _. ,
(
failures have not resulted in everpressurization or compromised the integrity I
of the reactor containment system, they do represent a deviation l
1 from the anticipated performance of an essentici safety system I i ;
[' __ - - _____ _ - _ _ _ _ _ _ - - - _ ____
l
5--7 i (overpressure relief system) that ftes safe.ty implications such as
~
excessive vessel cooldown rate, increased probability of fuel coolant loss and potential for a co:: con mechanism causing failure to .open.
Changes in design, equipmerg, inspection and testing can be made to improve the safety and safety / relief valves' performance on GESSAR plants. GE has propos-d a new valve design. They stated that the- !
I safety / relief valves to be used on GESSAR plants will be ". . . balanced (
,, type, spring leaded safety valves provided virh an auxiliary power
-actuated device which allous ope,ning of the valve even uhen pressure is -
<r >
less than the safety-set pref.sure of the valve." GE further states i that valve problems on operating ElTR's were ".. . associated principally
,, with multiple stage pilot operated sa fety/rclief valves. These never,
/ pouer operated safety valyce employ significantly fewer moving parts verted by the steam, and are therefore considered an improvement of the provloesly used valves."
1 Design details and ' drawings of the valves must be provided to the Regulatory staff, In addition, appropriate " Bench" tent data must be pro- ]
vided which verifies improved performance. GE vill be required to maintain a survcil2ance program once the neu valves become operational on any BWR, GE has agreed to provide most of the above information. They have not yet made an acceptable commitment to provide and report on their surveillance program.
I We uithhold final conclusions regarding ability of the pressure
{
( relief system, in conjunction with the reactor protection system, to 1
I
)
I
J l
l 5-8 provide adequate protection against overpressurization of the primary coolant boundary and unnecessary operation of the valve, subject to 4 4 a commitment by CE to perform and report on an appropriate surveillance progrcm. We will review this information when it becomes available and report the results of the review'in a supplement to car SER.
5.2 3 Fractu;c Toughness 5.2,3.1 CompJinnce with Ccde Requirements d
We have reviewed the materials selection, toughness requirements, and extent of materials testing proposed by General Electric to provide assurance that the ferritic materials used for presstire retaining con-ponctits of the rcector coolent boundary vill have adequate toughness under test, normal operation, and transient conditions. The ferritic materials will meet the toughness requirements of the ASME Boiler and Pressurc Vessel Code,Section III.
The fracture toughbess tests and procedures required by Section III 1
of the ASME Codt for the reactor vessel and other ferritic components provide reasonable assurance that adequate safety margins against the possibility of nonductile behavior or rapidly propagating fracture can be established for pressure retaining components of the reactor coolant pressure boundary.
l 5.2.3.2 Operatinn Limitations ,
We have reviewed the operating limitations that vill be imposed on i
o
l I
8 5-9 the plant, and conclude that the reactor will be operated in a manner that will minimize the possibility of rapidly propagating f ailure, in ~l i
accordance with Appendix G, 10 CFR 50.
The use of Appendix G as a guide in establishing safe operating limitations, using resulte of the fracture toughness tests performed in accordance with the Code and AEC regulations, vill ensure adequate safety marginn during operating, testing, maintenance, and postulated accident conditions. Compliance with these Code provisions and AEC sregulationc, constitute an acceptable basis for satisfying the require-ments of AEC Ccncral Decign Criterion 31, Appendix A of 10 CFR Part 50.
- 5. 2. 3 .3 Recctor Vessel Materini Surve111cnce Program The toughncsc properties of the reactor vessel beltline material vill be monitored throughout its service life with a material surveillance program that will comply with the Appendix H,10 CFR 50, and is consistent uith programs Lbat have been fcund r.cceptable for other 50 plants. Tha progrna ein meet the requirements of AST11 E-13.5-73. GE hr s said l l
that the predicted neutron fluence for this reactor 18 vessel is only 3.3 x 10 nyt. The program is acceptabic with respect l 1
to the number of capsules, number and type of specimens and retention of archive material, The surveillance program constitutes an acceptabic basis for monitoring radiatien induced chang,2c in the fracture toughness of the reactor vessel material, and will satisfy the requireacnts of AEC
( Design Criterion 31, Appendin A of 10 CFR Part 50.
't 5-10 i
- 5. 2.4 General Material Considerations We have revjewed the proposed materials of construction for the reactor coolant pressure boundary to ensure that the possibility of serious corrosion or stress corrosion is minLaized. The materials used are conpatible with the expected environment, as proven by extensive testing and satisfactory service performance.
5.2.5 h'ater Chemistry Control Further protection against corrosion problems will be provided by oontrol of the chemical environment.
The composition of the reactor coolant will be controlled; and the proposed maximum contaminant levels, as well as the proposed pll and conductivity requirements have been shoun by tests and service experience to be adequate to protect against corrosion and stress corrosion problema. The water cheuistry control is in accordance with the provisions of Regulatory Guide 1.56,
" Maintenance of Waler Purity in Boiling Uater Reactors."
We have evaluated the proposel requirements for the external insula-tion used on austenitic stainless steel componente, and conclude that they will be in conformanc'e with Regulatory Guide 1.36, " Nonmetallic Thermal Insulation for Austenitic Stainless Steel."
The controls on chemical composition that vill be imposed on the reactor coolant, and the use of external thernal insulation in conformance with Regulatory Guide 1.36, "Nonnotallic Thcrnal Inculation for Austenitic Stainless Steel," provide reasonable ascurance that the reactor coolant I
i a
i 5-11 pressure boundary materials will be adequately protected from conditions that would lead to loss of integrity from stress corrosion.
5.2.6 Inservice Inspection Program To ensure that no deleterious defects develop during service, cciccted wc]ds and wcld heat-affected zones will be inspected periodically.
Ocacral Electric has stated that the design of the reactor coo] ant I systen incorporates provisions for access for inservice inspections in !
.pecordance with Section XI of the AS:'E Boiler and Pressure Code, and that methods uill be developed to facilitate the remote inspection of those arecs of the renctor vessel not readily accessibic to inspection personnel. The conduct of periodic inspections and hydrostatic testing of pressure retaining components in the reactor coolant pressure boundary in accordance with the requirements of ASliC Section XI Code provides reasencble assurance that evidence of structural degradation or loss of leaktight in,tegrity occurring during service vill be detected in time to permit correctiva action before the safety function of a component is comprorgised. GE also states that the inservice inspection i program for Clans 2 and 3 componenta will fully satisfy the provisions i
of Regulatory Guide 1.51, " Inservice Inspection of ASME Code Class 2 i
. ?
and 3 Muclear Po.icr Plant Componcats." Compliance with the inservico inspections required by AStlE Section X1 Code constitutes an acceptable
{
basis for satisfying the requirements of AEC Cencral Design Criterion ,
i g 32, Appendix A of 10 CFR Part 50 t
1 5-12 5.2.7 RCPD Leakage Detection System Coolant leakage within the primary containment may be an indication of a small through-uall llaw in the reactor coolant pressure boundary.
The leakage detection system proposed for leakage to the containment will include diverso leak detection methods, will have sufficient sensitivity to measure sma?1 Icaks, will identify the leakage source to the extent practical, and will be provided with suitable control room alarrs and readouts. The major components of the system are the
.containne.nt atmosphere particulate, iodine, and radiogas . monitors, the dry-well floor drain' sump system, and tha drywell cooler drain system. Indirect indication of leakage will be obtained from the contaimncnt pressure and temperature indicators. The leahage detection systems prcposed to detect leakage from components and piping of the reactor coolant pressure boundary are in accordance with AEC Regulatory Guids.l.45, " Reactor Coolant Prenture Loandary Leakage Detection Systems" and provide reasoncbic assurance that any structural degradation resulting in leak-age during service will be detected in time to perr.it corrective actions. Compliance with the recommendations of AEC Regulatory Guide 1.45 constitutes an acceptable basis for satisfying the requirements of AEC General Design Criterion 30, Appendix A of 10 CPR Part 50.
- 5. 2. s neactor vescel an:1 App trtenances Uc have revieved all factoro contributing to the structura) integrity of the reactor vessel and we conclude there are no special censiferations
( that make it necessary to consicer potential vessel failure for a GESSAR plant.
5-13 I
The bases for our conclualon cre tb:t. the deci;;n, stress analysis, fabrication, inspection, and quality assurance requirements of a GESSAP reactor vessel will conform to the rules of the ASME Boiler and Pressure Vessel Code,Section III, and all applicabic Code Cases.
The stringent fracture toughness requirements of the ASME Code,Section III, will be met. Also,. operating limitations on temperature and pressure will be established for this plant in accordance with Appendix G, " Protection Against Non -Ductile failurc," of the ASME Doi]cr and Preunure Vcss21 Code,Section III, and Appendix G,10 Cri; 50.
The integrity of a GESSAR reactor vessel is assured because the a
vessel: .~
. i (1) Will be designed and fabricated to the high standards of quality i
required by the ASLIE roll 6r and Pressure Vessel Code and pertinent Code Cases listed above.
(2) Will be made from mat erials of controlled and denenstrated high quality.
(3) Will be inspected and ' tested to provide substantial ac>urance that !
the vessel will not fail because of material or fabrication deficiencies.
q (4) Uill be operated under conditions and procedures and with protective "
devices that provide assurance that the reacter vessel design con- "
ditions will not be exceeded during norcal reacter operation or during raost upsetu in operation, and that the vessel vill not fail under the conditions of any of the postulated accidents.
(5) Will be subjected to nonitoring and periodic inspection to deraon-( I l
strate that the high initial quality of the reactor vessel has not deteriorated significantly under the service conditions. ,
1 l
l
l
, 5-14 5.3 Thermal Hydraulic System Desinn 5.3.1 hulytical Methods and Data
{
' The analytical methods, thermodynamic data and hydrodynamics data used are similar to those used in the Grand Gulf, LaSalle, Bailly and Zicm:er designs and are acceptable to the staff. These are also pre- ;
sented in Section 4.4.
5.3.2 Anad Follouinn Chnracte_ristics The load following characteristics of the reactor coolant system )
provide the capability for one of the principal modes of DWR operation.
The docign of the EWn includes the ability to follow load demands over a reasonable range without recuiring operator action. The power can be controlled over approximately a 35% powet- rmuge by flow control. Be-cause of the negative void coefficient, load following is accomplished by varying the reactor recirculation flow. To increase power, the recirculation flow rato is increased thus sucaping voids from the nod-
~
erator and inercsaing core reactivity. As reactor power increases, more steam is formed and the reactor stabilizes at a new power icvel with the transient excess reactivity balanced by the new void formation.
Conversely, when 1 css power is required the recirculation flow rate is I i
, reduced. The resultant formation of more voids in the moderator auto- <
1 matica11y decreaces the reactor power to that commensurate with the l new recirculation rate. l t
1
( ______
5-15 -
The transient effects of such events as loss of full or partial coolant'flou, Joad changes, coolant pump speed changes, and startup of an-ir.cetive loop are discussed in Section 15.1 of this report.
5.4 Compopynt and S6 system Des Qn 5.4.3 Reactor Pecirculatj_on Syeten The reactor recirculation cystem consicts of two loops external to i
the reactor vessel, within the dryucil, that provide automatic load follaeing capability over the rang? of 65 to 100 percent of rated poecr.
Jhc leopa provide the piping path for the driving flow of water to the 20 reactor vessel jet pumpc. Each loop contains one high capacity (con-stant speed) mot _or-driven purp, a flow control valve, two niotor opercted ate valves (for ptnr.p w.intenance), and a bypass around the discharge gate and flow control valves. In each loop leaven the vessel in n 22-inch suction line and enters the suction of the recirculatico perap (. hich is below the vessel water).
The water is discharged at a head of 865 feet (at a flow rate of 35,!.00 s -). The fim, control valve varies the flow rate over a 35% I' power rouge normally from 65 to 100 percent power. The water from the recirculation pu:.:ps flows to 20 (10 per loop) jet pumps which are located in the reactor vessel and accelerates a portion of the flow in the annulus. Watar not accelerated by the jet pumpo returns to the recirculation pump through the sucrion Jines. There are various systen interlocks on the flou control valves and bypass valves that provide cssurance that adequate pump NT'Sil will be available and protect the pump from bearing or cavitation damage, i
5-16
(
5.4.2 Main Steam Line Flow Restrictors Each steam line is provided with a venturi-type flow restrictor within the drywell (between the reactor vessel and the first main steam line isolatdon valve). The restrictors limit flow to 200 percent of the rated flow should a main steam line break occur outside the primary l
r containment. The purpose of the restrictor is to limit the coolant !
blevdown loss prior to isolation valve closure to reduce the probabil-ities and consequences of fuel failure in addition to reducing the forces on the reactor internal structure during blovdown. The a
restrictors arc designed and fabricated in accordance with the ASME Code,Section III and are acceptable.
5.4.3 Main Steam Line Isolation Valves (USLTV)
Rapid acting isolation valves are located on each steam line on each side of the primary containment. On various signals from the plant protection system these valves close and isolate the reactor coolant from other portions of the plant. At the same time isolation i occurs, the'same signals from the plant protection system are sent to i various backup and emergency systems so that they automatically function as described in Section 6.3.
The analysis of a sudden, complete steam line break outside the drywell is described in GESSAR and shows the fuel clad is protected if the valve closes in 5.5 seconds or less.
l 1
i 1
l v
l
i 5-17 l
5.4.4 Reactor Core IsoJat son Coolant System (RCIC)
The RCIC system is a backup, high pressure source of reactor coolant that will operate independently of the normal plant ac power supply.
Its operational purpose is to provide an alternate source of reactor coolant to the vessel and to provide sufficient coolant to remove residual heat following reactor shutdoun and loss of feedwater flow without requiring depressurization of the recctor. The RCIC consists of a pump driven by a steam turbine taking stcan from one of the main steam lines upstrean of the isolation valve and adjacent to the reactor. i The pump takes suction from either the condensate tank or the suppression pool and discharges it to the reactor vessel through a head spray nozzle.
The system is designed to Class.I standards and is capable of being tested while the reactor is in operation. It has also been classified as an Engineered Safety Teature since it functions as an ECC system during certain postulated events such as the control rod drop accident, with the HPCS as primary backup system to the RCIC.
The P.cactor Core Isolation Cooling (RCIC) sys tem includes the piping, valves, pump, turbine, instrumentation, and controls used to maintain unter inventory in the reactor vessel whenever it is isolated from the main feedwater system. The HPCS provides a redundant backup for this function. The scope of review of the RCIC system for the GESSAR plant inc3 udes piping and instrumcutation dia2 rams, equipment layout drawings, and functional specifications for essential components.
(
i i
l 5-10 I The drawings, component descriptions, design criteria, and support-ing analysis have been reviewed and have been found to conform to {
Com.dssion Regulations as set forth in the General Design Criteria, i Regulatory Guides, and technical positions. The RCIC system has been found to conform to Regulatory Guide 1.29, "Scismic Design Classification."
The kCIC system and HPCS system jointly conform to General Design Cri-teria 2, 4, and 34. The two sy' stems have been found capable of trans-ferring core decay heat following a feeduater isolation and reactor l i
shutdown, from the reactor to the suppression pool, so that the core 1 i
Minimum Critical Heat Flux Ratio (MCliFR) does not decrecsc below 1.0 and '
the pressure within the reactor coolant pressere boundary does not exceed e
110% of design presnure. This capability has been found to be availabic even with n loss of offsite power and with a single active failure.
The staff concludes that the dasica of the Reactor Core Isolation Cooling System conforms to the Co:rmission's regulations and to applicable Regulatory Guides and staff technical positicus and is considered acceptable.
5.4.5 Residual llect negaval System The Ros3 dual licat Removal (RHR) system is designed for two principal normal modes of operaticn besides the safety related modes. For normal usage, the the RHR systen functions to remove reactor decay and residual heat during either a nornal shutdova or following isolation of I
l l: .,
l I'
/
5-19 i
the reactor. In one safety-related mode of operation, the EHR system (as LPCI) restores and maintadns coolant inventory in the reactor vessel I
after a loss-of-coolant accident. In the other safetv-related mode of operation, the RHR System provide's'a containment spray for condensing steam in the containment during ,t.lic post-LOCA period. These safety-related modes of operation are further discussed in Section 6.3 of this report.
The RHR system consists of tro heat exchangers, three main system i pumps, and associated valves, piping, controls and instrumentation. All- ;
a t functions 1 components arc desigged to satisfy seismic Category I design )
requirements. The main systen punps are sized on the basis of flow required durin~g the LPCI mode of operation which is the mode requiring the racximum 1
flou rate. The heat exchangers.cre sfacd on the basis of their . . .
I duty follouing a LOC.\. ;
j i
Twn loops, each consisting ~of one heat crchanger and one RHR pump ;
and au:111 cry equipuent, are phycically separated from cach other in the recctor building. A third loop, alco consisting of a pump and associated pip 1ng, can pump RHR service water directly into the reactor if necessary.
During reactor isolation, the RIIR system can be operated in the con-densing mode to condense reactor steam; hence, the RHR system operates in conjunction with the reactor core isolation cooling system (RCICS). With the rcactor isolated, reactor steam normally is directed to and condensed in the suppression pool via the relief valves and the ROIC turbine I
(
s-
i 5-20 exhaust piping. However, the suppression pool temperature under these conditions is limited in order that the water temperature rico due to a postulated subsequent design basis loss-of-coolant accident would not cauce the pool temperature to exceed 170'F during the reactor blovdoun.
The condensing mode of RHR operation relieves the burden on the suppression pool by transferring a portion of the steam generated by l decay heat to the RHR service water. The condensate is either dumped to the supprecsjon pool or returned to the reactor vessel through sthe suction of the steam-turbino driven RCIC pump. Shortly after shut-doun, both heat exchancers are used to handle essentially all of the decoy heat. After about 1-1/2 hours, the capacity of one heat exchanger is adcquate and the other rey be transferred to the suppression pool cooling code which utdidzas the RHR heat exchangers to cool the suppression peo] unter by transferring heat to the PJ1R service water.
This mode can be uc,ed in conjunction uith the condensing rnode or to
~
provide long tern suppression pool cooling following a loss-of-ccolant accident bloudoun.
The shutdown cooling mode and a reactor vessel head spray mode are operated during normal shutdown and cooldown. Reactor water is diverted from one of the recirculation loops, through the RHR pumps and RHR heat exchangers (shell side) where heat is transferred to the RHR service water (tube side), then the cooler reactor water is returned to the reactor vessel via a recirculation loop. Part of the cooled reactor
( ,
1
}
5-21 I
water flow is diverted to a reactor head spray nozzle where 1t main-tains saturated conditions in the vessel head volume by condensing the steam generated by the hot vessel valls and internals.
The system is protected againct overpressure.~.ction by relief valves and can be automatically isolated to protect the core from low water level in case of a becah in the cleanup syste.n. It is also automatie-ally isolated uhen the Standby Liquid Control System is actuated.
The scope of our review of the RHR System for the GESSAR plant incl ded u pi ip ng and instrumentation diagrams, equiprent layout drawings, failure node and effcets analysis, and performance specifications for' g essential components. Our revieu has included the applicant's proposed design criter a and design bases for the RHR and his analysis of tha adequacy of those criteria and bases and how well the design conforms to those criteria and bases.
The drawings, corponeut descriptions, design criteria, and supporting analyses ass'ociated with the RHR system have been revietied . The RHR system is not single failure proof in the shut-down cooling mode, and, therefore, does not conform to the intent of AEC General Design Criterion 34. An example of a single failure which could render the RHR system inoperable is a failure-to-open of one of the isolation valves in the RHR line leading from its
(
m O
( 5-22 l
associated recirculation loop. Such a single failure could place the reactor in the unusual position of being unable to achieve a cold shotdown condition uithin a reasonable period of time. It is possibic that some " bootstrap" type of operation outside of the RER system could be effective in achieving a . degree of shutdown capability (such as with the ECCS); however, ue interpret tlie intent of GDC 34 to be that the system normally utilized to place the plant in a cold shutdown condition (the RHR system) be single failure proof.
a Also, since the RU2 system is a lou pressure systea for which over-pressure protoetion is required, any design modification shall not f
reduce the level of pretection against uverpressurization, The ItRR r,ystet v313 be required to be modifled so as to be im:uunc ,
to singic active failures before submittal of the first E'JR/6 FSAR.
l 4
I i
4
l i
5-23 i
)
54,6 E R P.nacto n ,tcr cleej.y Synte,s 5.4. 6.1 _Sys_tsDfj._c,ripgon The reactor water cleanup system (RUCU) is used to maintain the l l
i chenieni purit.y of the reactor coolent. The portions of the reactor water cicanup syster.1 up to and including the outcrmoct isolation valve d
are part .of the reactor coo 3cnt prensure boundary. The applicant's {
des i!;it objectives for the system are to: (1) prevent excessive loss of reactor ecolant; (2) prevent the release of radioactive noterial from 1
the re.netor; (3) rcnove solid and disec1ved ir:.puritica from the coolant; and (4) dincharge excess unter during pouer transients. In addition,
. the syntcu is designed to minimine toupcrature credients, to conserve i
reactor heet, and for mairitvin6111ty during reccter cporation, The RWCU system flev rate is 154,000 lb/hr. The TECU system will cone ist of tuo 50%- capacicy pun.ps, regenerative 01:d non-regenerative f heat exchangcrs, and two 50% capacity filter-dimeralizers. The deair.erali::ed water may be cent to the recctor through the shell side .
l of the regenerative heat exchanger, to the main condenser hotuell, or i to the liquid radvaste system.
The PSCU-systen is isolated from the receter by tuo motor-driven isolation valves that close automatically in the event leakage is
(
I l
l i
i i
, 5-24 l
detected by the RCPE Leakage Detection System. The outermost drywell- !
isolation valve will also close automatically in the event liquid poicon is injected to the reactor by the Standby Liquid Control System or if the outlet temperature of the non-regenerative heat exchanger exceeds a predetermined level. Reverse flow is prevented by check valves in the return line to the fecdwater system and downstream of the RCh"J system pumps. Strainers in the outlet from the filter-denincralizera prevent resins frou cntering the reactor in the event
,of failure of a resin support. In the event of lov flow or loss of flov in the system, flou is maintained in each filter-demineralized by
^
its own holding pump. Sample points are provided in the inlet to and outict from cach filter-denineraliver to determine dcmineralizer DF.
The reactor water cleanup system will be used to aid in maintaining I the reactor water purity and to reduce the reactor water inventory as ;
required by plant operations. The scope of our review of the reactor water cleanup system in,cluded the system's capability to meet the anticipated needs of the plant, the capability of the instrumentation and process controls to ensure operation within limits defined in Regulatory Guide 1.56 cod the seismic design and quality group classifica-tion relative to Regulatory Guides 1.26 and 1.29. Our review has included single line diagrams and schematic dier. rams along uith descriptive information concerning the system design and operation.
l l (
t l
.i 5-25 The basis for acceptance in our review has been conformance of the i
applicant's designs and design criterla to the Cor.:niscion's Regulations !
and to applicable Regulatory Guides as well as staff technical positions and industry standards.
Based on the foregoing evaluition, we conclude that the proposed reactor water cleanup system is acceptable.
,A
~
e t
0 l
I 1
I-l 6-1 1
h l
4
- 6. 0 ENGINEERED SAFETY FEATURES l
6.1 Cencral The purpose of the various engineered safety features (ESF) is to provide a conplete and consistent means of assuring that the public will be protected from excescive exposure to radioactive caterials, should a major accident occur in the plcnt. In this section of our report, we discuss the reactor containment systan, the emergency cooling syster c, and the previsions for taintaini: the habitability of the co:itrol rec.: after poctulated accidents. Discussions of other en;.inecred a
safety features are provided elscwhere in thic report, as related to the pr.rt icular syate.as they directly serve. As will be seen, certeln of thcsc ESF syctema have functions for normal plant opert.tions as well ta their safety-related fuactions.
Syntern and conponents deeignated as engineered safety features arc decigurd to be capable of performin; their function of assaring safe shut dmin of the reactor under the adverse conditions of the various postulated design basis accidents described in Section 15 of this report. They are designed to seisnic Category I standards and they must function even with a complete loss of offcite power.
Components and systems are provided with sufficient redundancy so thct a singlu influre of any component or system, vill not result in the loss of the plant's capability to achieve and maintain a safe shutdoun of the reactor. The instrumentation cyctems and energency power syste.as are designed to the sane neirmic, redoniency, and quality I
________~
1 a
1 l
l 2
}
l 6-2 !
l n ]
l requirements as the systens they serve. These instrumentation and
.l l
onsite power systcms'are described in Sections 7 and 8, respectively, 1 of this report.
- 6. 2 Containment Systems The contaimacnt system for GESSAn includes a reactor containment structure, containment hect removal systens, a containment isolation sys' ,.a, a combustible gas control system, a shield building surrounding i the pria:,r- containment and the standby gas treatncnt systen. The l
derign of the containment systeu for CESSAn is siullar to the design l a
of the system for the previously reviewed Grand Gulf Puclear Station '
which will be the flrct n.:clcar reatjon to utilize the Mark III contclar nt dc sir;u.
1 The safety issucs raised in the cource of our revicw of the pro;'osed contnjnnent systemc, are barically the same as those issues rained durin3 our revicu of Grcnd Gulf. During the revicu of Grand Culf, the bacle analyt. teal apr.roac'.: cnd the dea Q n margins for the I
Marh III containment vere established. 1hc secpe of the large-scale '
Marh III test program uas clao uutermined. Ecsed on a successful 1 l
reco'ution l
of these safety icsuen, the Grand Culf containncnt design j i
was found to be acceptahic pending final validntion with large-scale .
test data. A cinilar approcch van taken for GESSAR as described below I i
and fernt the basis upon which our evaluation was performed.
{
1 l
I i
l
.]
1 6-3 i
l
\
l l
s I mgth 7 Utillo tNG n ,
l
)'*
- .'4 i
CONTAINt.'.f NT
,/
UITE M PCOL
' REACTOR i
O f) / f'l
/ ,.' ., ,..
/. e. ,
I'.
.: n g
<[.,.. y a
, 8< H2 ACTOR-ChitELD W A L L
/
, /,.4 h ;v,
/
cayweLL
,,, l , / el F t I a' L. j
' h > e M,,'. / ,-
1 e /
> , T n V 'W E n a.. .
"V L.- \
. 4 WalR WA1,L N '.
./
l . HORl2CNTAL r
e vests
.4 j P,-
t . .
s > -NJ,
. ' . ?> <jb ;f t ;
!,.),
t -
- 7. b,:v :T. .C.. ."F..~ '
K.P T.:,bj .Y ',.
.,,,'g.tl*
p *t h,
. ll-I
.- /: il ". *.* ' :. .';_ 1"d' fe,;x.--s, ,Y..
,.Uef Nf5tO*4 g
/ v*
9, 3P, , .,-
- <vre zwe m m_u erte ;!---
T ;j - *s-
- As; .. .c , a r?
h 2.< h ,,.-as - - . c .4.
4 NOL
- 1 i
' i.
FiCl!RE 6.2.1 TAARK lli CCTJTAINM.9NT (Containment and Shieid Bui!cHng) l I
l L----- -__
l l
l 6-4
(
\
6.2.1 Containment Functional Desien Inc contaituacnt function:1 design refers to the performance capability of the rcactor containment structure following postulated loss-of-coolant accidents. For GESSAR, a Mark III type containment maintainc a fission product boundary in the event of a loss-of-coolant accident. Figure 6.2.1 chows the principal features of the M:;rk III containue.nt concept. Tbic der $cn utilizes the effect of water pressure suppression and consists of separate dryvell and containment volures connected through a suppression pool by hori-a zontal vento. This dcsi;n t is bcsically the sane as that employed for the recently re /icued Crand Gulf f acility. It results f rom an i
ongoing develop.wntal progran of the prcosure suppt ession containment as discussed in Section 6.2.1.1 of thic report. Our review in this area included the temperature cad pressure responses of the drywell and containment to a spectrum of loss-of-coolant accidents; suppression pool dynamic effects during a loca-of-coolant accidcnt; or follouing the actuation of one or ir. ore reactor coolant systea pressure relief valves; the consequences of a losc-of-coolant accident occurring within the containment but outside the dryuell; the capability of I
the containacnt to withstat.d the effects of steam bypass of the suppression pool; and the enternal pressure capability of the drywell and containment and the systens provided to limit external pressures.
The review has considered General Electric's proposed design bases
(
6-5
(
and design criteria for the containment and the analytes and test l'
, data in support of the aduquw_y of the critcria and bases.
The contaj ouent system is divided into tuo najor subvolumes, i a dryac11 cncioning the reactor system, and the primary containment )
i l
surrounding the dryuell and containing the suppressica pool. The j l
contein..mnt and the drywell voluces are connected, through the 1
suppressica peal by an crray of harizontal vents in the drywell j
{
wall. The w pprresden poel nececs as r heat sink in the unlikely l l
(
event of a d.nsign basic lossiof-cocient accident.
a The pri:ary contr.inment is a free-standing steel' structure concisting of a vertical. cylin&r., domed top, and a flat base. The i f ft3 and net f ree volurae of tb.: priuar; c<> n t a ire:icn t is 1.16S x 10' 4
i the design precuure ia 15 paig. To satisfy its design bacis as a j l
ficcion product Icahnge barrier the primary containment is designed '
for c ler.hnne rate of 17l of the volere per day at 15 psig.
An additional structure called the shicid building, surrounds the prieary contaimient. ILs. purpose, in conjunction with the fuel 4 building and part of the cuxiliary building, is to provide a volune which fiscion product leabe.gc from the primary containment following l l
a postulated less-of-coolant accident can be diluted and held up j l
, prior to relcace to the environ:.cnt. Our evaluation of the shield )
l building desisu in ine.luded in Section 6.2.3 of .this report. I
< i 1
i 1
I l
l i
l l
1
i 6-6 i
Located within the primary containment is a substructure, called the dryvell,_which encloses the reactor and reactor coolant system. ]
The dryuell is an unlined concrete structure, enciesing a net free <
volume of about 274,500 ft 3 and.. designed for a differential pressure of 30 psid. The purpose of the drywell is to channel steam released during an unlikely loss-of-coolant accident through the vent matrix i
cystem to the suppression pool for condercation. k'hile not a fission '
product barrier the drywall must be free to gross leakage for adec,uate a
performance of the precsure suppression feature. ,
Since, for the liark III'Mosign, the containment completely surrcunds the drywell, high energy lince penetrating the dryvell must pacs throu;,h the containment volu5e. 1bese lines are designed to low streso levcis and high qtlality standards to preclude rupture inside the !
containment but outside the drywell. As an additional margin, the applicant has pre.vided guard pipes on certain high energy lines bewteen the dryuell and containment. The guard pipes will be designed to the same prescure ac the enclosed process pipe.
Because the pressure suppression concept relice upon a controlled channeling of steam through the suppression system, the possibility of bypass paths must be minitized. Our evaluation of potential bypass sources and containment bypasa capability is discussed in Section 6.2.1.8 of this report.
(
6-7 i
The suppression pool is a' 360-degrec. annular pool located in {
the bottom of the containmour'and retciced between the containment .
I wall and the dryvell weir wall. The weir wall is a 360-degree, reinforced concrete wall located inside the dryuell and 30 inches ,
'from the drywell wall. An t.dditional volume of suppression pool j water'is stored in the upper containment pool, located on top of the drywell, duria;, normal opecction. This water is added to the i supprc naion pool folluving a LCCA by the suppression pool tiaheup 1
l sys tem discussed in Section 6.2.1.4. During a norrual operation '
a 129,550 ft3 of wcter is contained in the suppression pool and 34,150 ft3 of mahcup t:cter in stored in the upper pool. The suppression l
-1 pool serves as a heat sinh fer postulated transients and accidents and as the source of coolicy trater for the emergency core cooling l l
sys t eais. In the cacc. of trancients that result in a loss of the j i,
raain beat sink, energy would be transferred to the pool by the dis- l 1
char:;a pipInc, from tlye reactor prescurc safety / relief valves. In ;
the event of a leeu-of-coolant accident within the drywall, the !
i horizontal vent system in the drywell vall would provide the energy tranuf er path.
Located in the vertical section of the drywell vall and below
, the suppression pool water 1cyc1 arc 120 horizontel vent holes of 27.5" dia:ceter and arrarced in 40 circunf erectici colunns of three vents. In the event of a losn-of-coclant accident the pressure will
(
l 1
l 6-8 !
t !
o rise in the drywell due to the release of reactor coolant, and force the level of water dotm in the weir annulus. When the water level l l
has been depressed to the level of the first row of vents, the differential l pressure vill cause air, steam, and entrained water to flow from the drywell into the suppression pool. The steam will be condensed in the poc1 and the air driven f re.m the drywell will be compressed in the primary containment'. Thc net ef f ect coul'd result in appro::inately a 4 pni rise in average conteinnant pressure. Pea!: differential pressure is calculated by the applicant to be 21.8 psi, a
Figure 6.2.2 illustrates the dryvell and containment pressure response as a function of tinc followiig a design basi.s loss-of-coolant ac ci d en t . !.n c1re chcun in Fi:;uro 6.2.2 the chart-tern, containment j response is choun in termc of two reglene.; onc representing the volume between the compression pocl and the Hydraulic Control Units (HCU) floors, cud tha other representing the reminder of the containment volunc. I Although thi., response decs rot r : i, - ;r o si;,i.ificantJy cifect peak drvi c11 differentint pressure it doc: result in pressure loads on sort. !
containT. cat internn3 structurcs. These considerations are discussed !
more fully in Section 6.2.1 9.
Following the initial phase of the accident, containment and dryvell prescure vill continue to rise due to the input of core decay and sensible heat to the suppression pool. The long- tern pressure rise will be limitcd to 11.9 psir, by operatien of the re-
[ dundant containment heat renoval system. Therefore, in thw pressure l
l I
6-9 l t responso analysis of this type of containment two limiting conditions must be concidered; the short tern drywell differentici pressure '
l and the long-term containment shell pressure. Our evaluation of 1 the applicant's analytical raethods for each of these time periods (i.e. , both long and short term) is discussed in Sections 6.2.1.2 i
and 6.2.1.3 of this report. The General Electric Company has also l ceupleted cr.cil-acale tests and is perforraing f ull-scale tests to support the brl. III short-terra at:aly tical model. Our review of these test programs is discussed belo.
a Loth the drywell and contain=nt are divided into a number of subec.:,portrents by internal structures. Our evaluation of the sub-t compartment dccigna is discussed in Sectio':. 6.2.1.7 of this report.
6.2.1.1 Pevieu of t'." Contein,ent Techno1cav Two basic prcosure suppression designs have preceded the Mark j.
III containment; d.e., the Mark I, or "lightbulb-torus" and the Mark II, or "over-under". A con,mrison or deaign parenaters for the three '
concainnont types is provided in Table 6.2.1. The uetwall and dryvell I 1
1 of Mark I and II vere connecteJ by a vent system which entered the l 1
i supprcscion poc] vertically and was at a constant submergence. For i both decipns, the design basis loss-of-coolcnt accident for containtert
, response was a rceirculation line break. In both Mark I and II
{
t containments, the peak drywell prescurc' occurred at about 10 seconds following the accident, which was af ter vent electing, and during the vent flow part of the transient. Uetuell peak pressures occurred j
(
o' I
6-10 I
in about 10-20 seconds due primarily to the compreccion of drywell air in the uctrall, licrh II containacnts alco experienecd a short-term drywell deck differential prcruurc uhich dould occur either at the time of vent clocring or later in the vent fleu trans.4cnts. Generally those a
plants with rel.ctively large vent areas had vent clearing controlled
.t peak <.;e.1 d if f er.:n't ial pre :. vros. In the lon<: t ern both the drywell v 6 .
and i. :Lecil recchs.d a ruoi.J: ry rc.nh prNsure due to continued dei ny u
/ heat gencreN on; houavor this transient ua.<> 3ess severe than the a ,
1 short L ore.?cnd t he cefoi e vas n0L controlling for establishing co'nt.ir;:9.tdeygnpressmcc.
I For con t u i ra;. .nt. .." aly :J n, "'1ho Gene':al F3 cetric Pressure Suppreycie:-
4 Contr.lusent Am]ytice.1 Mu?e)" as described in EEDD-10320 and ite
- s '(
supp14xnts ,uus u.c d. ' lair,rcodel connists of five separate cubmodels; blo'..<cun, dryrell, uet ael.1, vent cle.aring and vent flou. Eased on a reviev of th.& at:alytJeal methodc cc.elayed in th: model, correlation with Ilruboldt and Eodein I,c" test recuits, and comparison uith C05,TE:TI-PS results, the s taff has previoucly concluded that the GE model uas conservat ive and therefore acceptable. for containecnt analysis.
The 11crh III type containment proposed for GESSAR is different from the :iark 1 cud II types of conta:Lennts in three basic ways.
Tirct the D'..'R/6 t ype reactor system t hich is proposed for GESSAR, has relatively larger stern 31nes than the previous IER core desir,ns.
The effect of this is that the postuisted loss-of-coolant accidents l
I i
l I
e_-_________ ._
r
._ c 3 n t o ) f c t )
3 l n ' 3 n d t l e "0 t r I) e c
f em 3 f e y
h n t t 2 I R
' i.
r o
0 0
d i
si a
x 0 0 gi g a u t 2 s d M C t f
n 5, r l t en e
6 7, ns op
/
% 8 f
U L 4 8
~ i e
4 eo 1 3 l 0 5 9 0 1 f t 7 0 A t c . 6 (5 7 8 0 P( r 2 3 N c( 1 1 1 1 3 3 4 d d e c c e r r o o f f n n i i 3 e e M r r G
I d 3 d 3 3 S ee t ee t t E nt f nt r f D
) i e y ie y t
- . l r 0 r. n l r 0 0 g a :
': 2
'T I ~c 0 i d - c 0 0 i d P t l : c P ;
ec 0, p
/
l n eo 0, 0, n p
/
2 6 3 E -
c, 4 8 t
7- "
~1 ec 4 5 ec 3 6 5 3 2 4 0
.I .. t 0 5 t 0 0 8 4 7 0 s
- 1. , M ~- (
? ~1 4 0 s 1 1 4 0 2 2 2 2.
6O _. T E.C L il B T A T.
T r
O 7
O: )
S y l 2 l 3 I r I t I t 3 P r c f c f t
/
e h y h f y t P . F s 3 g a s 0 g a U 2 2
' I 0 i d 0 3 i d M t t O l s C " e 0, p
/
l e 0, 0 s p
/
2 f f
-' n e ? 3 e 9 0, 5 3
9 8 2 7
1 A : t 3 6 t 1 5 6 2 0 0 T
I v. s 1 5 0 e 1 3 5 0 3 4 3 .
r S
(
n n a o o e i i r
. t t c c a u e u e t r r r r n t u t u r e s s s s e v n s n e s w /
o e e o e m e o a a c -
c r e c m u r e p e a e u p t u l p t r e r L f l a f l o a a r a L o o n L. o l
r T o v n r a a C v g m v g m k k E e i k e l i k r a t a s
- Y p r a T p r o s e y a e n e 1
I i e e r. y i o e e h r e r P t a d l U t a p d l t b v b
~
l I
i 6-12 associated with the main steam line rupture and recirculation line l rupture result in nearly cquivalent peak drywell pressures. There-I l
fore both of thece postulated pipe breaks must-he considered in '
determining the DBA-LOCA for Mar'E'111 containment pressure response.
Second, the vent system connecting the dryuell and containment utilia.es a circunferential arrangement of horizontcl vents at three dif f erent elevations which 2 eads to an additionci functional de-pcnJcnce on vent c1 caring cad vent flow pheno:nona thnn the "crh I and II types. In nddition, hecnuso of the relatively large vent a
arcas provided, the peck dryac41 diff eren^dal pressore is vent clearing contrciled; i.e., the highest differential prcncure across the dryvr]1 occurn Curing " tnt clearing. Thic placco added e. phosis on the dyn a ;ca of veut cler:ri.ng but reancen the Jr.pcet of vent flou casurptio,s en drywc11 pressure.
Thl:d, cs th.: volut.e of t'hc conta!weent la chout five times thu of rhc 6> <ril, tha cocpreccion of drywell cir into the centain-ment durin? vcnt flou resulta in only a sr.ull (cbout four psi) rise in containment pr uar.u c e. This EDc11 cffcCt leads to a long-tcrc cont ainc.cnt pech pressure uhich is not specifica1]'/ related to the .
I sies of tl c reactor coolant bruch or the short-terrt, pressure responce. l j
Eccausa of the above, the staff hac concentrated its review in 1
those are.' chere previous analytical rodele or testing cannot be {
extrapointed to the Mark III design. These itcas cre covered in I
-]
detail in the following sections.
(
6-13 t
6.2.1.2 Short-Term Pressure Renponso As diccussed above both the main steam line break and recircu-
'lation line break result in nearly equal poch drywell pressures.
For the postulated doubic-ended rupture of a 26" main steam line the applicant has assumed a blowdown profile which is separated into an initial one-second period of stenm only blowdown followed by two-pla.nc, 31guid v. iter anj stear ble.edo..n due to liquid Icvc1 swall in the renetor vest.e1. During storm blewdorn, the mass and energy a
input rates to the contain:,ent were calculated assucing criticci flou of an idec1 gab. The tuo-phase blordown rcte uns based on the frict'onicss Moody critical f]ev trodel and the averege density of
. fluid inventory within the recrtor vt accl . The staf f ha:-; previously revieucJ and f ourd ucceptable these assu*ptir,ns for determining bicudoun rates.
The tice at ubich the liquid ]ccel in the revetuc vessel swells to the clevation of the steau line notrice following the break, detcrn:in>a the tire at which the rc Je3 che.nces f rom stun to two-phase bicudown assumpticac. Peak dryvell dif ferc.utJ n1 pressure can be sencitive to the icvel rise ti::- since tv.o-phace blerd. min yields a greater rate of steam addition to the dryvell th::n steam only blowdoun ;
and also introducca liquid vr.ter into the vent flev. Ecth of thesc l cffects incrcace dryeell presourss. I l
c' e
_ _ _ . _ _ _ . _-_---------4
6-14.
l In the GESSAR containment analysis CE has assumed a level rise time of one second and based on this assumption has calculated that the peak drywell differential pressure would be 21.8 psid (Figurc 6.2.2). GE has alto provided studies of level rise time as a function of operating conditions which indicate that the most rapid level rise would be about one second assuming a hot standby condition.
Follouing the postulcred design basis loss-of-coolant accident, the drywell pressure will rise and accelerate the water in the vent annulus. At about 0.92 secondo, the first row of vents will be cleared of water and a mixture of air, stcan, and water will flow
(
into the suppression pool. The vater in the vent annulus will continue to accelerate dounuard resulting in clearing of the second rou of vents at about 1.17 seconds and the third row at about 1.52 seconds. The peak dr>vell differentini pressure occurs at 1.40 seconds (main, stcan line break) and is a result of sufficient vent area being uncovered to reverse the pressure transient in the drywell. Due to this phenomenon the peak pressure is predominantly controlled by the dyncmics of vent clearing and only partially influenced by vent flow assumptions.
In the analysis of the vent clearing transient, General Electric used the vcnt clearing model described in the " Mark III Analytica] Investigation of Small-Scale Tests Progress Report",
[
11EDM-10976, and " Fourth Quarterly Progress Report: Mark III 1
1- _ . _ . _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
9 6-15 i e, i 9 O O N w W
=
- f. g% _
- g. m
- ... - .. y
-s s E. i e
m e
m fg oc >w ldi-g+
N. . ' "
b C Ea i f 6 D ou '
e
> > e z - u m u t; p J e c 3 O O -c ,
il i < x 3 5 I h 5 O O O f < a c E
[
'E O n W o o > E -
2 3 d' 8 ~
E 8 "
jl 8 8 s - r 5
~
/-/ a r
a I
5
{.
/ t -
E
- m M
O M u >=
c' ~ I w -
[ <
2 b -
~
~
h W
- v. i
.[ -
d
- w n.
a
/ t -
a U 4 I. 2 O >
y y e
pl ~~t 5 -
.y
- u. r- a.
a o f -
- u. e s.
o u
M o e v'g
- Z w
k 5 ##
g b w
.J u. /
/*
WO o
q
=
N /' E #,5 e
c s
=
c T,-
w f%.M t
1 s
sgu m . .:
r@
es u
h cN % d b !c.' 2 - C)
.% w i o-- - -
m ,
~
e ,s !
w--
- d ]
p n N =.
B i d - w
" A ,v, v c e e c; . p
.a!,.
F JS .a e M
i
.J Z W<U<.L _
LL. I l-a W
w f& 1 l
- h E
t ub5 Cum 1 $
o 2 -
O O g I
j g l j %
! I _t i i I -
O b
@ $ 0"
( w ,auncsaua
)
I.
C _ ___ _ _ .
6-16 !
I Confirmatory Test Program", NED0-20345. In this model the vent system is nodalized into six control volumes representing the vertical weir annuluc and horizontal vents. Conservation of mass and momentum is applied to each control volume to determine fluid accelerations and vent clearing times. An effective vent length is used to simulate the effects of suppression pool inertia and turning loss coefficient are applied to account for changes in flou path direction and area. The loss coefficients currently used in the model are derived from generally accepted data and a
General Electric intends to verify these coefficients during the large-sca]e liark III testing program.
The Ccnaral Electric vent flou model has also been revised to consider the more complex lbrh III horizontal vent geometry; houcver, the basic thermodynamic flow assumptions used for previous water--pressure suppres.sion designs, remain unchanged.
Ior GESSAR the vant flou vas computed on the basis of parallel path flow splits which are a function of the number of uncovered vents and geometric loss coefficients. These loss coefficients will also be confirmed experimentally on the large-scale facility.
Based on these analytical models, General Electric has determined that the postulated rupture of a main steam line !
would result in the highest dryuell differential pressure and has calculated this pressure to be 21.8 psid. GE has stated in GESSAR that the drywell will be designed for a pressure of 30 psid l
I l
l
l l
l 6-17 which provides a n,argin of 37% above the peak calculated value. !
Both the Ecgulatory staf f and our consultants, the Aerojet Uuclear Con:pany, have revieued the ana]ytical tuodel used for the pressure respon.se calculation. Based on our review and our consultants' rcce...aendatiout ue bslieve that this t.mtsin is adequate to account for possibic unen tainticc in the GE vent c1 caring and vent flow r.odels.
a -
As shoun in Figure 6.2.2 the short terc. containtcnt response is calculated for two regionc; the wetucil, rhich includes the volur.u betueen the cupprer.nir.n pcol and t,1.c Hydrnalic Control Unit (11C0) floor, and the retrairer of the contcincent volume. An indiuted en Tigarc 6.2.2, t he two vo]uce analycis doec not show aN s it ,ni ficen t effect on penk calcoleted drywc11 differential prcococe. At t h i ', t in, , ho w ver, GE has not made acai]abic the detailt. of t iir. nualytical approach or justificatloa that may h,e appropriate in tcrus of rclevcut test data from tha f
lorge scalc ?arn III test fnellity. L'hcn this information is provided, we vil] conplete our review of this aspect of the
, analysic and report our conclusions in a supple:nent to the SER.
(
)
l1 6-17a )
l 1
(
t l
l l
{
i l
GC has also provided analyses of the drywell pressure [
response for a postulated rupture of the recirculation suction line. CC has calculated that the peak drywell differential prescure for this break (19.9 psid) ic less than chat calculated for tbc stecn line break. l The chort-tcrn bloudo'.m rate is a sensitive parameter for a Mark Ill contain:nont cince the drywell pressure peaks very early "in the transicut. CE has used a.ppropriate assue.ptions in the recirculatic.a line blowdoen nodel to accurately represent the
, ; chart-teru ef fect by inclutird.fhe unas inventory of the recir-culation line. Th:. nadel m:cc a 1 roah arca equ"1 to the cross scetion area of the suction line to account for the reactor vessel side of the break and an equivr.icnt brea?: aron equal to one half t h's euction line area th t si: ulater the e.;terna? rceirculatien loop contribution. Th$ effective break area for the loop side is derived from an anelycio of the pre; cure-tiac history within the h
l
(
6-18 f
loop following the break and which calculates a ness flux approximately 50% of the mass flux predicted by the Moody flow model using the initial fluid conditions within the pipe. 'Such modelings remain a valid consideration until the initial nass inventory' within the external loop is depleted (~2 seconds).
Subsequent mass flow then becomes Itmited by the effects of critical flow through the jet pump nozzles. In sur.cary, the total bloudoun is calculated on the basis of a single effective volume representing the prinary systcu, cass flux based on the !!oody correlation and a brech arca profile which equals 1.5 the suction line area up to 2.0. seconds anu then becomes equal to the suction line plus jet pec;p nozzle area subsequent to 2.0 seconds.
h'e have reviewed GE's recirculation 1dne blowdown nodel, and have also performed independent calculations of the nass flux 1 i
from.a recirculation suction line break using the RELAP-4 computer l I
code and nodeling assumptions as prescribed by 10 CFR Part 50.46, "Acceptanet Criteria for Emu cer.r: Core Cooling Systens for Light k'ater-Cooled Nuclear Power Reactors", January 4,1974. The results of these calculations show total nnss release rates substantially I
lover than those calculated by GE. On this basis we believe that j the blowdown rates presented in CESSAR are acceptable for use in the analysis of drywell pressure response to the postulated rupture j of a recirculation suction line. Further, since the peak drywell
(
l f
6-19 differential pressure calculated by CE for this break is less than that for the main stenn line ve conclude that the latter is the most liraiting pipe rupture for the dryvell. Since the drywell has been designed for a difforcatial pressure of 30 psid which provides a nargin of 37% above the peak calculated value. and since we feel that this margin is adequate to account'for the uncetteintler in the GE vent clenring and vent flow r.toJcls, we conc 3ude that the proposed dryecil design pressure is acceptabic. ,
- 6. 2.1. 3a gng Tern Presourc Response Follouinn the short-term bicudoma phase of the accident, suppression pool temperature and containment pressurc will increase e
due to the coatinua' input of decay an.1 sensibic heat into the containment. Refcirint; to Figure 6.2.2, at about 100 seconds af ter t he accident the dryvell pressure has stabilized to approv.1-mately 3 psi above the contaiv ant pressure. This differential pressure corresponds to the subenrpence of the first row of vents, at some later tiec the dryvell and containment pressures will
. equalize due to the return of air from the containment.
During this tJme period the ECCS pumps, taking suction from the suppression pool, have reflooded the reactor pressure vessel 8
up to Ll'c level of the uain stenn line nozzlcs. Subsequently ECCS teter will overflow out the breah and fi21 the dryvell up to the top of the voir wall, establishing a recirculation flow
( I I
i l
________-_____-_a
6-20 I
path for the ECCS coolant. Also during this time the Suppression Pool Mahcup System has added water to increase the suppression !
pool inventory in the long term (Sec Section 6.2.1.4).
At about 30 minutes following the accident, the containment cooling mode of the Residual Heat Removal (Rl!R) Eystem is activated and suppression pool vater is circulated through the RHR heat exchangers, establishing an energy transfer path to the service water system and ultimate heat sinh.
u In the long-tern analysis, GE has conservatively accounted for potential post-accidcnt encrcy sources. These include decay heat, sensible heat, ECCS pump bear, cud metal-vater reaction cnergy. OE bac acnumed that the only heat sink avuilubic in the containment is the suppression pool and the only mechanism for heat rejection is the E!R heat exchnnscre.
The long-tere model also assurmd that the containment atuosphere is caturated and equal to the suppression pool temperature at any time. Therefore, the contaimaent pressure ;
i is equal to the partial pressure of air and the saturation l i
pressure of unter corresponding to the pool temperature.
Based on the above assumptions the applicant has calculated the peak contalument prescure to be 11.9 psig. The design pressure of the containment is 15 psig which allous a 28% margin chove the peah calculated value. On the basis of our review of the cpplicant's
( analysic and the prescure margin we conclude that the containment !
design pressure for this plant is adequate.
l l
6-21 l
t l
6.2.1.4 Suppression Pool Mckcup Systen (SPMS)
GE has recently proposed the addition of a SPMS. The Suppression Pool Maheup Systen provides water from the upper containment pool to the suppressica pool follouing a loss-of-coolant accident. This increase in long-term suppression pool inventory provides additional
'l pool heat capacitance, a mininwa long-term drywell vent coverage f i
of tuo feet, and r.ccounts for any post-accident entrepnent of water in the contcinoant or drywell. l Two 24" lines connect the upper pool, located on top of the 1
drywell, to the suppression pool. Each line contains two normally cloccd valvcc in series uhich open on a lou-low st.ppression pool level nignal in coincidence with a LOCA sinnal permiusive.
Dumpin'; of the upper pool would stcrt tuo to three minutes following the bei,inaing of ECCS flou and would require about five minutes for . compl etion .
The origin of cuch a system for the Mark III containment is re3nted to the pool dynr.nic forcos imposed on containment structurec i following a loss-ef-coolant accident. Test results have indicated that the c : tent end ragnitude of cuch forces are proportional-to the submergence of the horizontal vents. Therefore, decreasing i
the vent suboorgence, by a reduction in pool vater volume ,results in Ics, cevere structural design requirencnts en containment i internal structures.
(
l 1
l l
6-22 i
Such a reduction in normal suppression pool volume is possible in a Mark III containment since the containment accident response is separable into short and long-term modes. The short term, or blowdoun, phase requires less pool mass for acceptable temperature increments due to steam condensation than the long-term phase requires for absorption of decay and sensible heat. Therefore, the suppression pool can be maintained at a reduced volume during norcal operation and during the early phase of an accident if its inventory is augmented for the long term.
a Since GE only recently proposed the suppression pool nakeup concept, we have not had sufficient time to complete our review which will include a failure mode and effects analysis. We will report our conclusions in this area in a supplement to the Safety Evaluation prior to issu:nce of the PDA.
6.2.1.5 Extarr,1 Pressure Deci s The drywell structure is desigred for an external pressure of 21.8 psid. A drywell vacuum breai:er systen is provided to control sup-precsion pool water Icycl in the weir annulus and prevent inadvertent floodin; of the drywell. The systen is not required'to operate to maintain the structural integrity of the drfv;11. C2, ho'tever, has not c1carly deceribcd the design of the vacuu.a brec::ers.
1 .
The containment vescel is dr.isned for an e::ternal pressure of 0.8 psid. A containment vacuut relief syeten, consisting of four, 36" lince, is provided to naintain external prercures within design limits.
g Each vacuum relief line connects the containment
6-23 to shield building annulus and contains one check valve in series with a motor operated globe valve.
We have revicued the dryuell design external pressure and find j that it is acceptable since it represents an upper limit on possible external pressures by assuming complete depressurization of the dryvell to 0 psia. The inclusion of a Eryvell vacuum breaker system, houever, is a concern cince it introduces a number of dryuell penetrations which could potentially be sources of bypass lechage between the dry-well and containnont volumes. Therefore, to leduce the potential for hypass
, Ical:nge, and since the system vill no longer be required to limit enternal pressures on the dryuall, uc also conclude that CE should specify a systen ubich utilizes a minimum number of small dryucl1 penetrations comparable in size to the minicun bypass'capabillty. The resolution of this icsue vill be reported in a Supplement to the Safety Evaluation.
s We have revieued the applicant's sizing analysis for the containment vacuum breakers and ue find that certain assumptions used may not be sufficiently conservat-ive. Uc are currently pursuing additional analytical studies with GE in order to determine an appropriate basis for sizing of the containment vacuum
- breakers. This item is generic to all plants uith Ibrk III containments utilizing the free standing steel shall type of construction. We vill report resolution of these items in a Supplenent to the Safety Evaluation.
l
/
l l
, \
l
{ J
6-24 6.2.1.6 Test Progran The General Electric Company is presently conducting a full-scale test program to experimentally establish the vent clearing and vent flow performance of the Mark III containment concept.
Full-scale testing was started in November, 1973 foJ1owing completion of a two-year small-scale test program. The staff and its consultants, the Acrojet Muclear Company, have discussed the test progran with General Electric and visited the test facility on several occasions.
a A total of 67 small-scale tests of the Mark III concept have been performed by GC since June 1971. The test arrangement simulates the containment at a scale of approxir.ately 1:2000 on a volume basis. Sen11-scale test data have been reported in " Hark III Confirmatory Test Progran Progrees Report", MED'i-10348 and "Narh III Analytical Invantigations of Small-Scale Test Progress Report",
NETI-1097 6. Correlations between test data and analytical predictions for vent c] caring times indicate reasonabic agreement in this scale.
The full-scale tcs: facility has been volumetrically scaled to cochup an 8 sector of a Mark III containment including one column of three full-sice vents. The initial test series began in November 1973 and considered steam blowdouns for one, two, and three 28-inch vont configurations. The objectivca of this cories were to establish a correlation betucen experimental data and the I
General Electric vent clearing model and to verify the values of
J 6-25 loss coefficients used in the vent clearing and flow models. '!
Data from these tests are currently being evaluated and correlated '
with the analytical r,odels. Follouing the initini test series, !
a group of air tests were performed with the full-size vents to obtain pool svell and impact loading data. Additional test series are planned ubich will consider the following phenomena:
- 1. A test serics consisting cf a co3uun of five 16-inch vents for continucd investigation of pool suc11 uill be conducted. The y results will be used to select various combinations of 1, 2, i
and 3 vents to determina the effects of vent spacing, vent ]
interaction and e:: tended blevdouns, t
- 2. A ccrics of liquid blowdowa tests will he conducted to indicate comparability to the vent clearing and vent flow performance as deterrined in the steam blowdoun scrics.
- 3. A series of crall break tests will be conducted to investigate pool stratification,n and vent chugging effects.
i
- 4. Tests vill be performed with the suppression pool at an initial elevated temperature to determine steam condensation capability.
- 5. A multi-vent tect series will be conducted uhich vill employ a vent test section of three colunms of three, nine-inch vents to consider vent interactions, vent c1 caring, and pool maldistribution for a sub-scale mockup of a 24* cector of the containment.
Based on our revicu of the IIark III containment concept and the
( associated test program we consider the present testing to be a continuation of earlier prossure suppression testing done by CE
6-26 in support of previous BWR containment designs; i.~e., Mark I and Fbrk II. The emphanis of the large-scale tests has been directed at those aspects of a Mark III containment which are innovative and which therefore must be demonstrated experimentally and correlated to analytical predictions. We conclude that the design of the i large-scale facility and the scheduled tests to be performed will provide a sufficient data base to establish the performance !
characteristics of the Ibrk III and to validate the analyt.ical I
approach taken by GE in its accident analysis of Mark III contain-a meat. Our cone. ult. ant s concur i.. tni., tvalua tien. .:e will report in SEP suppl er .m t c. , further test res.ilt s be::<n.m e vn !) abic .
6.2.1.7 Subcompartment pres:wre Ann.lyscs i
Within both the dryvell and containment, internal st,ructures form subcompartner.ts or restricted vc 'unes which are subject to dif ferential pres.sures following postulated pipe ruptures. In the dryvell there are tuo such volumen; the annulus fort.ca by the reactor vousc1 and the biological chic 1d, and the dryuc11 head region which is a cavity surroundin;; the reactor pressure vessel head. In the containment varfous coracnents such as the valvec, heat exchanpcrs, and filrer/deminerallnors of the Reactor Water !
Cleanup (RWCU) cystem are located in individual compartments.
GE has not submitted all of the assumptions uecd in the sub- f compartment pressere analyz.ec, nor have they pre m ted the results of those analyses at this time. Uiwn these analyces are provided
( we vill complete our reviet.' and report our conclusions in a '
Supplement to the Safety Evaluation prior to issuance of the PDA.
1 l
l-t 1 ;
1 1
6-27 '
i 6.2.1.8 .Secan P," ness of the Senorension Pool Possible bypass leahnge paths frca the dryuell to the outer containment have been considered in our revicu of the Mark III i containment. The control of such byposs paths de inportant to ensure that the dceign pressure of the containment is not exceeded I for postulated design basis accidents.
There cre three potential sources of stenn bypass of the suppression pool ascocirted uith the Merk III contain .cnt sra used in a CCSSAR. First, sinco t he drycell is of reinforced concrete con-struction, the potential e::ists for crackins of the drywell structure under accident loading condition'. c This can ' allow direct leakage of bloudoun oteau to the contaira r.t volu:e. Second, the design of the corbestible gar cont.rol system.s could allow the opening of
, direct f1w paths betosen the dryucll and cc: tain~.ent for the dilution of hydre ;cn. /.lthough thnse syctcen crc dc.stEnad to operate aftec blowdcyn'is complete, residual stecting in the reactor vessel coatinocs after blowdoun duc. to the addition of deeny and sensible heat to the ECCS coolant. This energy could be added directly to the contaim:ent atnoaphore. Third , ' parts of the Reactor Unter Cleanup (F.'CU) syster.i are located within the princry containnunt but outside the dryw]1. This system has high energy i I
pipe lines, connected to the reactor primary system, which do not j have guard pipes. Therefore, postulated ruptures in these lines I
would result in bicwdova of reactor coolant directly to the con-4 tainment atnosphere without benefit of energy absorption in the l
l suppression pool. l l
l l
6-28
.t In the case of postulated R'.1CU system pipe breaks the applicant has provided design fectures to terninate the blevdown prior to exceeding the decian .linits of the containment. Two isolation valves in serfeu are provided on both the RUCU suction and return 1 lines uhich vill automatically isolute the RilCU system from the primary reactor syaten. Isolation sfgnals will be generated by tuo lechage detection syrtens; ont based on Rh'CU system flou couparisono and another Laced on compartment temperatures. In addition, a flou linitc1 is provided in the suction line to limit the rate of blowdown prior to isolation. based en sensing Icakage a n .: a ualni u, GL :.as uncuJ.ated t h.it. the con'..ain:..ent prescure respance acroning a .Mrd pino rupturn would be lcas the.n 5 psig, which is below the centainnant design pressure of 15 poig. '
As discusced in Section 6.2.5 the design of the combustibic gos centrol cystetus is not complete. 1le vill reviau these systers to ensure that the potential for inc.dvertent bypa s of the suppression pool is nininize.1 and that an intentional bypassing (i.e., post blo'..'down) is uithin the containment's capability to tolerate bypasc leakage. The resolution of this arcr. uill be reported in a Supplement to the Safety Evaluation prior to issuance of the FDJ In regard to bypcss leahngo associated with potential cracking of the dryuell or other sources around penetrations, we conclude that the GESSAR containment should have an allowable bypass area i
l i
6-29 !
. 1 l
l
)
of approximately one square foot (A/vE~) for the spectruu of reactor coolant system breaks. The allowable bypass area is considered to be that leakage aret betueen the dryvell and containn.ent which would result in containtient pressurization to design pressure following.a postulated loss-of ~c'colant accident. To mitigate the effects of bypass, a heat renoval systen is necessary. For l t
CESSAR cuch a system is the contairacnt epray synten which is 1
an operating mode of the Residtial llent Renoval (IM!R) system. j a GE has shova that starting the containment sprays following a 10-minutedelaytosatisfythc.syd~ tem'sECCSfunctionprovidesamininum bypass cdpibility of about 0.0 ft (A/v2) for small primary' system
.' breaks. Weconsidercuchcap$bilityadequete. CE has also made a commitment to autenatically actuate the sprays uhen required. We find this to be acceptable pending our review of t he design details.
In addition, uc u)11 requito a cormitment to perforn a leahage test of the drywell at approxir.:ately desih n pressure prior to pla' t operntlen, and lou pressuro leakage tests of the drywell periodically during plant lifetine. The full pressure test vill inpose loads on the dry, ell uhich are a substantial fraction of the accident loads and will prevido added ascurance that the dryecil, as constructed, !
conforms to its design bases. The acceptance criterion for the '
teste should be based on the measured leakage being less than the
(
l i
r
6-30 leakage corresponding to a flow path of A//E = 0.08 ft at the test pressure. He conclude that.if the leakages are not in excess of the indicated limits, there would not appear to exist any potential bypass paths in excess of the design capability of the Mark III contain..cnt systcm. The resolution of the automritic suitch to th: con-tainnent spray uode and the high pres:.ure lenker,e test will be repcrted in a Supple: Tot to the S'.fct-; hyr.luntion pzior to ierut.n:e of the TbA.
6.2<1.9 Pool Dynarates rollowing a loss-of-coo 3 ant accident in the dryuell, the drywell e
air vill be congreased due to blowdoun rass and energy addition to the volunc. Following vent clearing, an air /cteam/ water mixturn t
vill be forced from the dryvell thrcunh the vent system and injected into the suppression pool, approximately 7-10 fact below the surface.
The eteen corponent of the flou mixture vill condense in the pool while the air, being non-condensibic, will be relcaecd in the pool as high pressure behb]cn. Tbc continued addition and expanalon of air causes the pool voln...e. to swell resulting in an acceleration of the surface vertien11y upward.- Due to the effect of buoyancy, air bubbles will rise fa9ter than the pool water mass and will l eventucily breah through the swollen curface and relieve the driving force behind the pool. I The proposed GESSAR containment design included internal structures and floors at lov elevations which could be subject to loads due to pool dynamic effects. Floors for the R'<.'CU pumps l~
I I ,
L- 1
6-31 !
I i
were located at the -5'3" elevation, approximately four feet above the pool surface, and the Ilydraulic Control Unit (RCU) floors at i c1cvation ll', or approximately twenty feet above the surface.
To determine the magnitude of pool dynaraic loads, GE performed a series of air tests, using the large-scale Mark III test facility in February and Ibrch,1974, and is presently conducting a series of 1/3 scale tacts. The air tests indicated that large impact loadc, due to pool suell, could be expected on the lovest floors.
, Rather than dens 3ning the structures to these requirements GE has proposed relocati:ir; the PJ.'CU pumps in the auniliary building. He find this to be an acceptchle raodification. The air tests alen e I indiccted that siraificant differential pressure loads could be developed across the hCU floors and that pool dynaele forces were proportienc1 to the sula.eg:ence of the horizont:1 vents. In this regard GC has proposed that the suppression pool water level be lowered 'to allow 3 esc sevci e structura2 design require.r.ents. GE vill be (va3cating the l'CU leading, conditions more extensively in the 1/3 ccale tect series and we vill revieu and evaluate the results of the test program and establish appropriate design nargins as part of our onsoing revicu effort for the Mark III containment.
In the courss of our revicu we have also identified the potential I
for flow caldistritation eff ects within the drycell and vent systen.
These effects could occur in the following two ways. First, blochage
(
6-31a of part of the vent annulus or vent holes vould result in limited flow through the restricted part of the vent systen uith consequently higher flows in the remainder of the vents. Second, the bloudown from a postulated pipe break in one area of the drywell could preferentially clear those vents nearest the break and delay c1 caring of those vents furthest away.
CE has cubr.,itted analp as of the effects of partial vent bicchage av.' non-uniform veut c1ccring to the staff. The results a of the vent b3cchage analysis indicate that calculated dryuc11 peak prensure is not particularly sensitive to this effect for bicchen;c of up to 20% of the vent 1.cles (on a circu ,ferential i
ba sir.) . The effect of 20:l bloche;:e is onir a 0.5 psi increase in peah dryuell differential pressure. In the cane cf potential flow unidistrib tion or localicat f mb the ef fect of preferential vent c]cariuc vas ninulated by accu..inp a varichle trat er 1cyc1 along the circumference of the weir annulus. The results of the ant.lynin indiccte that there is virtually no rct effect en drywell peak proscute since those annu]us sections eith increased submergence (delayed veci clearing) of f set the jnf3uence of those annulus sections with reduced subr:ergence (prenature vant clearing).
Based en thn2 results, th ' s taf f ~u L ti a t por tu : r' t ora in the vent flo.! distribution will not significantly affect the peak dryuc11 prencore and th:refore the analycis of the desipu basis
( loca-of-coolant accident prescutcJ in GZS2.n la ac:xpla:>2 e fer establishing the drywell design pressure.
\ _ __- ________--_-__--_ - - _ -
6-32
(
Flow maldistribution effects could also be manifested in the containment where circuaferential or radial waves could be generated in the suppression pool due to non-uniform vent flow and/or non-uniform air addition to the pool. Analyses of these phenomena have been performed by CE ubich indicate that no significant vent uncovering due to waves is expected. In addition, GE has shoun that the liark III conlahuaent has a significant tolerance for vent uncovery for the postulcted conditions leading to wave fornation.
,Uc find thcoe results to be acceptabic.
6.2.1.10 pM.prle.i.d Con _chicions The apr 11 cant hat calculated the sho e and long-tcrn drywell and contcinnent pressuru ci described above. Eaced on our revicu of the appliccnt's annlytical methods, we conclude that the drywell design differential pressure and containment design prensure are adequate. Ue have reviewed the large-ccele tect facility design and the sch:Juled test.n to be perfor.aed snd find that this program should provide sufficient data to confirm the unalytical vent models used in the short-term analysis. In addition, as part of our ongoing revicu ef fort for the Ibrh III containment, uc will be reviewing data from the test facility directed at pool dynamic cffacts and parametric studies.
Our revicu of other aspects of the containment functional design has renalted in the fo]Iouing conclusions:
( 1. The applicant uill he required to reduce the rw.bec and size of the dryvell vacurm breakers.
i 6-32a 5
- 2. The applicant will be required to provide additional justifica'tion i
for sizing of the containment vacuum breakers.
- 3. Uc. vill review the adequacy of subcon.partacnt decian pressures when this inf onnation and supporting analyses. arc provided by CE.
I
- 4. The cpplicant vill be re. quired to nerform a lenhcr,e test of the dryuell at about the design pressure.
- 5. The derign of the containacnt and containment internal structures to cecernodate pool dyncr.ie forces ic bei- c, dcyc1cped on the basin of large-scale tect data. We will continue to review a
nud evaluette this information as it becor..ec avcilchle, l t c ::.c., 1 Lhron o u l- i.111 bc 1e,oi s.1 ,.riur : o iscud nc,,, of t.he 1 D.i . I:t, 5 ne. 6 re:nirtio p: f o; to t: e fir : I';/s :>.< 133 rson, 6.2.2 Con t a i__nie. .:.t_ Wr.t - -n._ ~ _c_
y
.c. 1..
The contcinr.:ent heat renoval system includes the piping, valvec end techanic.11 components used to reove encrgy from the containacnt followlar; p loss-of-coolant accident. Per GESSAR this systen is t he Recif ual Heat Removal (RiiR) Syste.a which, when opercting in the supprest. ion pool cooling er containment -
}
cpray raade, renovce energy from the containment to limit long-tera .
l containment post accident tcuperatures and proscures. Our revicu I
in this area included process c.nd inctrurentation dicgrees, ;
~
descriptive infor: nation con.mernio ; syrtem functioning and interaction with eccential supportin; systecc, General Electric's
( proposed systen design bases an-! criteria, and analysta in support of the adequacy of those basrc and critoria.
I r
i 6-32b i
The PJin Systeta consists of two heat exchangers and three pumps. One heat exchanger and one purp forn an independent loop
.and each loop ic physically ee;arated and protected to nininite the potentia] for nin;lo failures causing the loss of function of the errire systen. The third 'pury is located in a separate roon cod c:.n be couaceted to either loop. The RllR Systen is decip,r:d to Catencry I scirric criteria.
Oy rating in the centalirn ht ccoling node, the RRn pumps take
, sve t ino frem t! c suppression pool, pees the flou through the RUll hent exebangers, and direct tb [ cooled vater to the suppression pon] , the rorctor vesac1, or t.hp centcinm:nt sprcy headern. The Jucntio;is of auction and returb lines in the suppression pcol facilitate tni::ity; of the return water uith tbc. total pool inventory before the return vater becomes availabic to the suction lines.
Strafuers are provided on the cuction line inlets.
Analyses of the heat ren:nal capability of the RHR system have bcen presented based ca a service uater temperature of 100'F and cu overall b at exchanger duty of Si0 Btu /*F sec. These analyses indiccte adecuate heat rem.wal ecpability to limit the suppression pool temperature to 179'F cnd the containment pressure to 11.9 psig follouins a postulated loss-of-coolant accident. We find these values to bc vithin accer.tabl:' linita cad are appropriate interface values. Detailed characteristics of the heat exchanger design are 1
l
- _ _- a
6-32c site depciident and therefore, each applicant for a standard plant euct }srovide the above infernation and demonstrate that the W
availablehect renoval capacity for his plant is at.lecsc as great as th:t precified above. i GnerakElcetrichasstatedinGESSARthatadequatenet pocir:ive h'te cioa head is availabic at. the R!IE pump inlets assuming.
the contairrent is at atuer.pheric prescure and the pool is at caturation tengerature. These assumptions are consistent with the requireuento cf Rer.,ulatoty Guide 1.1 cnd thereforc acceptable.
Prov!siens are node in the containment her.t renoval systen to pernit a inservice inspcetion of synten components and functional testing of active'coqion mpc.,
I?e concludo thnt the contalnuent heat rmoval syst er can be opc:nt.cd in such.a nenner es to provide adequcte cooling to the con tair.mont following a len?-ef-coola'M accident and conforms to
, z\ ~
General Design Criterio 33, 39, and 40, and is acceptahic.
6.2.3 secor.ct n ev Centrin ant Fur. t t en n2. Degff,n. I The neondary conca$rannt systin includes the structures and i i
systers ened to centrol and treat radioactive leakage from the
) f .,
prinary conti facent in the event of a loss-of-coolant accident.
i i. ,;
For CESSAR Sh,cecch:lcry containment syst;m consists of the sMc1d s jn buildings the fuel builling, parts of the au>:iliary building, Lt1c SkTield Euildiq (nnulus Recirculct '.un and D.hauct Systc= (SWCCS) j n . W.
6 ' 'r
(
( l
(,
q
/
> L A.
- ),
(
' 'L2 .
l i 6-34 and the Standby Gas Treatment System (SGTS). Our review in this crea iucluded schetnatic flow diagrams, descriptive information concerning cystem functioning.and interaction with essential supporting systems, Cencral Electric's proposed systen design banca and criteria.
The shield building is a cylindrical, reinforced concrete c.crecture corp 1.ctc3 y enclosing the contc ltawat vessel. The Shf e.td bui]dlig. Annulo.4 P.acircu3ation and Osb> cst System nain-a tains the anr.alut forned by che chield building, an1 containment
- at a no;;ati ve prescarc (opproxhately -5" w.n.) during nernal oper.-;1an and follo',ing a posteleted loss-of-coolcut accident.
Fol3 ouing on accident, a varial:1c. fraction of the SDEES enhaust
- is di?:ected to the Stendby Cas Trentmaat System (SGTS) for filtrntfon p-ior to release to the atnosphere and the renainder is recirculated to the annulus fcr additional holdup and dilution.
CE bowever has not specified the credit acalgned for m12:in:; the j recil:ulci.Jon flou in the analysis of the radiological consequences I j
1 of a loss-of-u,olant accident. We uili require that CC provide and justify the celected value. In doing so, GE must also establish the fraction of total containment leakage which leaks to the annulus volume (cn opposed to the fuel building or ECCS and PJ:CU rooms) and which therefore can be recirculated. We will reputt on the resolution of thcsc itcma in a suppleuant to our SER.
( i l
I O
6-35 The Fuel Building and the ECCS and RUCU rooms in the Auxiliary
.Luilding are also part of the secondary containment system. These volunes are maintained at a negative pressure (-0.25" w.g.) during operation by normal plant ventilation systems. Following a postulated accident, the ECCS and EUCU rooms will be maintained at a negative pressure by the Standby Gas Treatuent System (SGTS), which filters the exhaust flow prior to release to the atnoophere. Tha Fuel Building uJ1l be connected to the SGTS only in the event that high a
rediction in the exhaust flow is detected. Uc find this unacceptable and we vill require that the SGTS be aligned to the fuel building upon receipt of a loss-of-coolant accident cigccl. (Additional l.
I deficiencies relative to the SGTS are discussed in Section 11 of thic report.)
The Shield Euilding Annulus Recirculatica and Exhaust System provides active conponunt redundcncy, is designed to seismic Category 1 criteria, and is located within scinaic Category I structures. Redundant components are separated and protected.
The Standby Gas Treatment System (SGTS) is designed to seismic Category I criteria and is located within seismic Category I structures. The SGTS consists of redundant exhaust fans and filtration trains cach consicting of a demister, heat coil, prefilter, HEPA filters and charco.'1 filter. The system is designed to maintain the secondary containment volumes; i.e., the shield l
l
1
\
._ . i..
6-36 1 e
I -
1 building annulns, fuel building, and ECCS and RWCU rooms, at a j negative pressure following a LOCA and to process the exhaust flow prior to release to the atmosphere. Redundant components are separated and protected.
- Following a postulated loss-of-coolant accident the pressure in the secondary containment volumes could increase due to inleakage and the starting time required for the SGTS. Additionally, the annulus pressure and temperature will increase due to heat transfer dthrough and expansion of the primary containment shell. GE has provided an analysis of the annulus pressure transient which considers the above phenomena and which indicates that the annulus will be maintained at a negative pressure of about -1" w.g. or greater.
We find this to be acceptable. However, we will also require GE
, to provide similar analyses for the fuel building and ECCS and RWCU room volumes. .
A certain fraction of the total containment leakage could bypass the secondary containment system due to leakage through lines or penetrations which are open to the containment atmosphere i
following an accident and which pass through the accondary con-tainment boundary. GE has identified several suc'n leak paths and has made a comnitment that the total nmount of bypass
, leakage will be less than 1% of the total containment leakage.
I i
! - ~
L-
i l . p l
(
i 6-37 9
1 Based on our review we find that GE may not have identified all potential bypass leak paths. . We will require GE to reevaluate their design with regard to bypass leakage and we will also require GE t'o make a commitment to perform periodic leak tests, beyond the require-ments of Appendix J of 10 CFR Part 50, of each potentia 1 1eak path and make repairs as required. For each proposed plant referencing GESSAR we will also review the applicant's test program to determine the adequacy of the testing methods and review the test results to determine the capability of the design to maintain bypass leakage within the prescribed limits.
6.2.3.1- Containment Air Purification and Cleanup Systems There are two engineered safety feature air cleanup systems proposed.for GESSAR. They are the Standby Gas Treatment System
, and the Control Room Air Cleaning Unit.
The staff has analyzed the engineered safety feature filtration systems designated by the applicant to operate in emergency situations with respect to the positions in Regulatory Guide 1,52, l
" Design, Testing, and Maint'enance Criteria for Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants." We find the applicant's design is in
)
agreement with these positions and have used an adsorption efficiency ]
of 99% for iodine removal in our accident consequence computations (see'Section 15). I J
1
. . y s .
pay- l
. -. c.,
I!
6-38 i
L i
6.2.4 Containment Isolation System I: 1 (, .
The containment isolation system includes the containment isolation valves and associated piping and penetrations necessary to isolate the primary containment in the event of a loss-of-coolant accident. Our review of this system included the number and location of isolation valves, the valve actuation signals and valve control features, the positions of the valves under various plant conditions, the protection afforded isolation valves from missiles and pipe whip, and'the environmental design conditions specified in the design of components.
d The design objective of the containment isolation system is to allow the normal or emergency passage of fluids through the contain-ment boundary while preserving-the integrity of the containment
( .
boundary to prevent or limit the escape of fission products from a postulated loss-of-coolant accident. GE has specified design-bases and design criteria as well as the isolation valve arrangements used for 1,ol.cion of primary containment penetrations.
1 No manual operation is required for immediate isolation of the containment. Automatic trip valves are provided in those lines which must be isolated immediately following an accident. Lines that must remain in service following an accident for safety reasons are provided with at least one remote manual valve. The containment isolation systems have been designed to the ASME Section III, Class 1 or 2, code and have been classified as Category I I
seismic design systems. 1
L 6-39 l GE, however, has not provided a consistent description of Instrument line penetrations of the primary containment. We will require GE to provide this information and justify that the design conforms to the positions of Regulatory Guide 1.11. In addition, the environmental design criteria specified for isolation valves and other safety related equipment located in the drywell and containment are ,
l not clear. We will require GE to clarify their environmental design criteria and bases and will report on the resolution of this item in a supplement to this SER.
GE has proposed that the containment purge system be operated
" continuously during normal operation to limit the buildup of activity and allow plant personnel unlimited access for surveillance and maintenance. The purge rate would be 4,300 cfm through 42-inch
( supply and exhaust lines. Fast-acting isolation valves are installed on the supply and exhaust lines to provide rapid containment iralation in the event of a loss-of-coolant accident.
The Regulatory staff considers that continuous purging of the primary containment through large penetrations is undesirable.
However, the staff could find continuous purging to be acceptable if the applicant takes the following appropriate design measures to ensure rapid and reliable containment isolation in the event i
of a loss-of-coolant accident: 1
- 1. A charcoal filter system should be provided to allow for filtration of the purge exhaust flow. (See Section 11.3 of this report.)
k' ,
~
i l
6-40 1
l
- 2. The isolation system should be designed to the quality and j
(
redundancy criteria appropriate to engineered safety features.
- 3. Redundant instrumentation and control systems should be provided l and actuated by diverse parameters. ;
- 4. A reevaluation of the purge and exhaust flow rate requirements should be conducted and minimized such that the penetration sizes can be reduced.
- 5. A comprehensive testing program should be proposed with the intention of demonstrating the performance and reliability requirements for these isolation valves.
, 6. The isolation signal shall be purge line radiation.
We are currently discussing this matter with GE and will address the final resolution in a supplement to this Safety
( Evaluation.
6.2.5 combustible Gas control The combustible gas control systems include the piping, valves, components, and instrumentation necessary to detect the presence of combustible gases within the prima'ry containment and to control the '
concentrations of these gases. Our review of these systems included the potential sources of combustible gases and their yields, the .
accumulation of gases within each volume of the containment systen, the capability to monitor the concentrations of the gases, and the capability to control and reduce the combustible gas concentrations by suitable means. -
1 4
4 g y j i
. _- L. .
6-41
(.
f.
- l. Following a loss-of-coolant accident, hydrogen may accumulate l within the containment as a result of metal-water reaction between the fuel cladding and the reactor coolant, and as a result of radiolytic decomposition of the post-accident emergency cooling water. GE has used the same assumptions as Regulatory Guide 1.7 to calculate the rate of hydrogen released by radiolycis and consistent with the Regulatory Guide has assumed that hydrogen is released as a result of 5% metal-water reaction of the fuel cladding although the~ analysis of the emergency core cooling system indicates that the metal-water reaction will be limited to much less than 1%. For the determination of containment hydrogen concentration, I GE calculates the rate at which hydrogen is released by metal-water reaction by arbitrarily maintaining the cladding peak temperature 1
of 2300*F using a core power distribution more conservative than 1; the'ECCS/IAC power distribution (which is the acceptable limit for emergency core cooling analysis) until the equivalent of 5%
of the cladding is rea'cted. With these assumptions the Regulatory Guide 1.7 hydrogen flammability limit (4%) would be reached in the
~
drywell at approximately one hour following the accident. GE has proposed a redundant hydrogen mixing system and a hydrogen recombiner to limit the hydrogen concentration within the containment to below 1
4 v/o. The design of these systems .:Ls in accordance with the l provisions of Regulatory Guide 1.7. However, the Regulatory
- - m: . . . . . ,
y
L 6-42 1
(
staff has recently completed a review and evaluation of the current guide and' has proposed, in part, that several revisions be made to the design parameters specified by the guide. Of particular relevance to the GESSAR d'esign would be the revision of the assumed.
extent of metal-water reaction from 5% to either five times calculated or 1% of the cladding mass, whichever is greater.
The-proposed, revised Regulatory Guide 1.7, has been placed in the Public Document Room and is currently under review by the ACRS Regulatory Guide Subcommittee.
a We'have. informed GE of the contemplated revisions of Regulatory Guide' 1.7 and advised GE to consider the potential impact of such changes on the design of the GESSAR combustible gas control systems.
4 Accordingly, GE 1,s in the process of studying different design options and in particular will be investigating the possibility I
~
of eliminating large penetrations of the drywell which could result in unacceptable bypass leakage paths. We will report further on this it'em in a supplement to the SER.
Due to the aforementioned potential changes in system design criteria the staff has not made a determination regarding the !
currently riroposed combustible gas control design for GESSAR. )
I Rather, we will wait for a finalized version of"the revised guide !
. I to be approved and proceed at that time to review GE's modified ]
I I
I
'( r
- ghe ewimus+ee y
l
. ._ . . 6.
I 6-43
(
design in light of the new guide. We believe that this issue can be resolved in a timely manner and we will report our conclusions in a supplement to the Safety Evaluation.
6.2.6 Containment Leakage Testing Program The GESSAR containment design includes the provisions and features necessary to satisfy the testing requirements of Appendix J, 10 CFR Part 50. The design of the containment penetrations and isolation valves permit individual, periodic leakage rate testing at the pressure specified in Appendix J,10 CFR Part 50. Included d
in the proposed program of leakage rate testing are those penetrations that have resilient seals, such as, airlocks, equipment hatches,
, and fuel transfer tubes.
( The proposed containment leakage testing program complies with the requirements of Appendix J, 10 CFR Part 50. Such compliance provides adequate assurance that containment leaktight integrity can be verified throughout the service lifetime and that the leakage rates will be' periodically checked during service, on a timely basis, to maintain such leakages within the specified limits.
Maintaining containment leakage rates within such limits provides reasonable assurance that, in the event of any radio-activity release within the containment, the loss of the contain-ment atmosphere through leakage paths will not! be in excess of the j
j l
[s l
$$ $ w *- 6 m *
@ b & 6estems e-+---
- sum g
L.3 1
6-44
(_
' acceptable limits specified for the site, i.e., the doses will be well within 10 CFR Part 100 guidelines. Compliance with the requirements of Appendix J constitutes an acceptable basis for satisfying the requirements of General Design Criteria 52, 53, j
-and 54, of Appendix A of 10 CFR Part 50.
6.3 Emergency Core Cooling System (ECCS) 6.3.1 System Description The ECCS subsystems provide emergency core cooling during those postulated accidents where it is assumed that mechanical
~
" failures occur in the primary,c.oolant system piping, resulting in the loss of coolant from the vessel at rates greater than the available coolant makeup capaclty using normal operating equipment. ;
The ECCS subsystems are provided in sufficient number, and with adequate independence, diversity, reliability, and redundance that, even if any single active component of the ECCS fails during i
a loss-of-coolant accident (LOCA), adequate cooling of the reactor ;
core will be maintained. k The ECCS consists of two high pressure systems and two low pressure systems. The former are the High Pressure Core Spray (HPCS) system and the Automatic Depressurization System (ADS).
The latter are the Low Pressure Core Spray (LPCS) system and the Low Pressure Coolant Injection (LPCI) system, which is one mode of the Residual Heat Removal (RHR) system. The ECC systems for the GESSAR reactor are functionally identical to General Electric 1969 product-line facilit'ies (LaSalle, Bailly, Zimmer).
( !,
[
\; ,
,1 6-45 I
('
All the ECCS are initiated by a high drywell pressure signal or a ' reactor vessel' low water signal, except for the ADS. Initiation l
of ADS requires coincidence of both of these and a third signal, !
4 indicating pressure at the discharge of at least one low pressure i
ECCS pump. The ECCS is designed to provide adequate core cooling and to limit the peak fuel rod cladding temperature for the complete
-spectrum of break sizes and locations up to and including the design basis loss-of-coolant accident.
The ECCS can operate independently of the offsite electrical a power from the onsite diesel generator and battery systems. All
. evaluations have been made assuming that only onsite electrical.
power is available. In addition, ECCS performance capability has been shown to be adequate assuming a failure'of any single active component within the ECCS. This single failure criterion has been applied in addition to and coincident with the assumed coincident loss of offsite power.
The HPCS system consists of a single motor-driven centrifugal pump and associated system piping, valves, controls and instrumentation. i The system is designed to operate from offsite power or from a separate diesel generator. Suction is taken from the condensate tank or the suppression pool and piped to a spray sparger over the core
. (via two entry points at the shroud). Nozzles spaced around the sparger spray the water over the top of the core and into the fuel s
'L h i
l 1
6-46
( assemblies. The system is designed to function over the entire range of reactor coolant system pressures and break sizes. For small breaks the system will maintain reactor water level. For intermediate breaks that do not depressurize the reactor vessel rapidly, the system will depressurize the vessel. For large breaks, rapid depressurization occurs and the EPCS cools the core in the spray cooling mode until sufficient inventory is accumulated to terminate the transient.
The pump characteristics are selected to satisfy requirements a for both high pressure, low flow rate deliveries for small breaks and low pressure, high flow rate deliveries for large breaks.
When the cooling system is activated, the initial flow rate is
{ established by primary system pressure. As reactor pressure decreases, the flow rate will increase until the required core spray flow rate is achieved when the differential pressure between the reactor vessel and primary containment reaches 200 psi. The pump is, designed to deliver 6110 gpm at 200 psid, 1465 gpm at 1140 psid and has a shutoff head of 1370 paid.
The ADS reduces the reactor pressure so that flow from the LPCI and LPCS can enter the reactor to cool the core and limit
~
the fuel cladding temperature. The ADS utilizes eight of the 22 safety-relief valves in the nuclear pressure relief system.
Automatic opening of these valves requires coincident signals l
k 6-47
(
of reactor vessel low water and high drywell pressure along with a high discharge pressure indication of any LPCI or LPCS pump, but only after a timer delays operation of the relief valves for two minutes. If the operator determines that the initiation signal is false or depressurization is not required, the timer may be recycled. The ADS is redundant to the HPCS and is only required if the HPCS cannot maintain reactor water level following a LOCA. Similar to the HPCS, the ADS is not required for large breaks.
The LPCS system consists of a motor-driven centrifugal pump (that can be powered by either normal offsite power or the standby ac power systep); a spray sparger in the reactor vessel; and piping, valves, instrumentation and controls to convey water from the suppression pool to the sparger.
The HPCS system operating in the low pressure mode serves as a redundant core spray loop to the one LPCS loop. The LPCS system protects the core in the event of a large break in the nuclear system and when the HPCS in unable to maintain reactor vessel water level. Such protection extends to the small break in which the ADS or HPCS has operated to lower the reactor vessel pressure to the operating range of the LPCS. The pump is designed to deliver 6110 gpm at 122 psid and has a shutoff head of 289 psid.
s-
.k's 6-48
[ Since the number of fuel assemblies and the diameter of the core has changed compared to previous designs, spray distribution tests'will be performed on a simulation of the GESSAR reactor to assure lthat an adequate amount of spray reaches every assembly. ;
GE states that no significant differences are expected from other core geometries previously tested for spray distribution. The results of these tests will be available by the end of 197S. GE has agreed to provide this information. We conclude that this is acceptable for GESSAR.
The LPCI system consists of three motor-driven centrifugal pumps (that can be powered by either normal offsite power or the standby, onsite ac power system), associated piping, valves, controls and instrumentation. Each LPCI pump injects water from i ,
the suppression pool through a nozzle in the core shroud into the space between channel boxes over the active core. The suppression pool suction, vessel injection nozzle and connecting piping for each pump a,re separate and independent. Two of the pumps also function as RHR system pumps. These two pumps receive power from different ac power buses. One of these buses also
e
, 4
. .. .__ L I
['
i' 6
- l l
u
.. (. : .1 The LPCI system provides cooling water following all LOCAs i exceptlthose resulting from small breaks that can be controlled ~
by the E?CS system. The LPCI is redundant to the LPCS system.
j Each LPCS pump delivers 7100 gpm at 26 psid and has a shutoff i head of 225 psid. .
As in previous BWR designs, the GESSAR' reactor has the capability to use the LPCI pumps t'o spray water'into the containment. Presently, ,
a diversion of these pumps becomes possible through manual action i 10 minutes after a loss-of-coolant accident. .We will require that a diversion of these pumps be automatic'when required. In previous designs, an interlock prevented diversion'of the LPCI pumps if the vessel water level was below 2/3 the active core height. In the h proposed arrangement for this plant, this interlock would not be present but an interlock preventing LPCI pump diversion to contain-ment spray until 10 minutes after a LOCA would be present. In support of a similar change on another docket-(Grand Gulf) that proposed diversion of 'LPCI pumps at a specified time after a LOCA irrespective of vessel water level, GE has stated that diversion of LPCI pumps ten minutes af ter initiation of any loss-of-coolant accident will not significantly affect the performance of the ECC system. The Regulatory staff has requested that analyses of the performance of the ECC systems over the complete break spectrum be performed assuming that two LPCI pumps are diverted from the
. ,- -s e
1 1
o M4s> # #*e e "**=***d*em e s
-M i
, L- l 6-50 l
. (
ECC function. The Regulatory staff's position is that these analyses (or similar analyses using the appropriate number (s) of diverted LPCI pumps and assuming 10 minutes delay) are required to determine if the performance and acceptability of the ECC systems is affected.
The Regulatory staff addressed a concern regarding the overall role of manual actions required to mitigate the consequences of a i LOCA. The staff's position is that because GESSAR is intended to represent a large number of plants, the potential for undesirable consequences resulting from poiential improper manual actions
~ becomes significant and warrants a thorough assessment of all items addressed in a staff request for additional information
_C (6.125). Furthermore, this thorough assessment is required at the FDA stage to allow incorporation of any design changes found necessary before the plant is constructed (such as incorporation of any new automatic actions found necessary to replace previously assumed manual actions and changes in signals presented to the operator). The GE response was not adequate to allow such a review to be made, and therefore, an adequate response to' question 6.125 is still required. That response should consist of all requirements, such as guidelines, criteria and suggestions, that GE prov' ides to the utility for use in preparing the Emergency Operating Procedures. In addition, GE should provide the Regulatory
. i S
M%- 9mm -
u.
6-51 staff a copy of the Emergency Operating Procedure frem a recently submitted plant judged by GE to be appropriate for future GESSAR Emergency Operating Procedures, along with an explanation of what differences are judged likely to be present between the submitted document and the future GESSAR document.
The following additional outstanding concerns should be addressed prior to submittal of the first BWR/6 FSAR: 1) results of spray distribution tests on the BWR-6 must be submitted to the Regulatory staff. The results in the topical report (NED0-10846) are for BWR-4 and BWR-5 configurations, with results for BWR-6 promised during 1974,
- 4) final . copy of the proprietary version of the 8 x 8 tr spray cooling test must be submitted for staff review, and 3) there are outstanding questions given to GE in May 1974 concerned with or related to ECCS.
( These questions should be answered on the GESSAR docket prior to the start of our FDA review. The subjects covered were a) a list and description of the purpose of pre-operational and startup tests of certain l
(mostly ECC) systems; b) justification of applicability of Referenced Reports to BWR-6; c) description of methods used and results of 1
blowdown load calculations on reactor internals; d) a list of all J ECCS related valves operated by containment isolation signals; e) details of the calculational methods used to show NPSH require-ments of ECCS are met; and f) quantitative details of the main steam line radiation detector's ability to detect failed fuel. The resolution l l
i of these items will be discussed in a supplement to this SER. Resolution of these items is not required prior to issuance of a PDA. j k ..
(
,wa -
.-, me- r ** -- * - =
6-52 L($
Furthermore, there are outstanding. unacceptable items concerned with ECCS instrumentation and contro1' systems. These instrumentation and control systems items'are addressed in Section 7 of this SER.
6.3.2 Performance Evaluation
'}
In Section 6.3 of GESSAR, the applicant provided an analysis of the ECCS using the assumptions and calculational techniques
, described in the Commission's Interim Policy Statement dated i
' June 19, 1971-titled, "AEC Adopted Interim Acceptance Criteria for Performance of ECCS for. Light-Water Power Reactors."
On De'c ember 28, 1973 the Commission issued the Acceptance Criteria for' Emergency Core Cooling Systems for Light-Water-( Cooled Nuclear Power Reactors which is a new rule and replaces the Interim Policy Statement. When the requisite evaluations are' submitted to the Director of Regulation, as required by the .
1 implementation schedule contained in the rule, the staff will make its review and conclusions.
Fuel assemblies identical to the GESSAR (8 x 8) fuel have been previously reviewed by the Regulatory staf f. However, since the fuel assemblics in the GESSAR reactor are of a significantly-different design than the 7 x 7 type of assemblies considered in the General Electric Evaluation Model described in NEDO-10329 and s
referenced.in Part 2 of Appendix A to the Interim Policy Statement, the staff has l
(
L 6-53
(
reviewed the evaluation model to determine its applicability to the new fuel design. The features of the new fuel design which are different from the old design and are significant in determining ECCS performance, are: a) smaller diameter fuel rods; b) a larger number of rods in each assembly; c) an unfueled central rod; and d) a thicker wall of the fuel assembly channel box.
The features of the evaluation model which might be affected by these changes in the design of the fuel rod and which we reviewed include applicability of the transient critical heat flux correlation, a
the thermal radiation grey body view factors, and the spray cooling convective heat transfer coefficients.
We have reviewed the differences in the thermal and hydraulic
( characteristics between the 8 x 8 fuel assembly described in GESSAR and the 7 x 7 assembly described in the evaluation model, and as discussed in Section 4.4 have concluded that the critical heat flux correlation is equally applicable to both designs.
We have also reviewed the differences in thermal radiation and 1
spray cooling characteristics between S x 8 and 7 x 7 fuel assemblies l
and conclude that the procedure used to calculate the heatup of i
a fuel assembly following a loss-of-coolant accident is in conformance with the approved General Electric Evaluation Model. Our conclusions l I
are based on sensitivity studies which were presented for identical l
. , _ , k l
6-54
(
fuel assemblies in the GGNS PSAR, and independent calculations using a computer program developed for the staff, and the results of full-scale 8 x 8 rod array heater bundle tests. Spray cooling tests with both stainless steel and Zircaloy clad heaters have been completed and have been submitted with the exception noted above for the Zr test proprietary report. The staff will review the results of these tests and reexamine applicability of the tspray cooling portion of the evaluation model to an 8 x 8 assembly.
Various single failures, including failure of the HPCS diesel dl generator, were assumed to determine the situation that resulted in the maximum calculated fuel clad temperature. However, one out-standing single failure not considered is closure of the recirculation
( loop valve during a LOCA. Our concern is that closure of the valve during a LOCA would degrade heat removal capability during flow coastdown. The issue must be included on the GESSAR docket for l
proper resolution and will be discussed in an SER supplement. We require .
that: 1) demonstration be provided that flow control valve failures would not cause flow decay greater dian pump coastdown assumed in the LOCA analyses, or 2) Enginnered Safety Feature grad equipment must be added to the recirculation flow valves .to preclude valve closure in the unbroken line during a LOCA. This analysis should be submitted prior to the first BWR/6 FSAR submittal.
1
( .
. _ _ _ _ ____J
6.
6-55
( Previous analyses have shown that the complete severance of a recirculation pump suction line results in the highest fuel clad temperature and greatest amount of metal water reaction, and is therefore designated as the design basis loss-of-coolant accident.
Breaks in any other line in the reactor coolant system at any location, including a main steam line or any of the ECCS water injection lines, result in a lower fuel clad temperature and less metal-water reaction.
Analysis of the design basis loss-of-coolant accident results a in a calculated peak fuel clad temperature of 1430*F assuming failure of the LPCS-LPCI diesel-generator with a break area of 3.3 f t2 which is the sum of the area of a 24 inch recirculation line and the area of 10 jet pump nozzles. The calculated peak clad temperature decreases as the break area decreases to about 750*F at 0.5 ft2 area and then increases to about 1320*F at 0.1 ft , with failure of the HPCS, diesel-generator resulting in the highest calculated temperatures for small breaks. The calculated percent of metal-water reaction is 0.03% or less for the above cases.
The fuel heat-up following the design basis loss-6f-coolant accident is arrested within two minutes by reflooding the core with
, ECCS vater. Since fuel cladding is at elevated temperatures for only a short period, no significant amount of cladding will be embrittled. Therefore, fragmentation of fuel rods will not occur k e s
I
. ~
_ _ . _ L.
q L 6-56
(' i and the core geometry will be preserved. Swelling or ballooning of the clad is unlikely to. occur, but tests have demonstrated that even if ballooning occurs, it is limited, can be defined, and will not significantly affect the cooling of the core.
After Ehe core has been recovered, the ECCS valves are realigned to direct water through the RER heat exchangers. The core will continue to produce fission' product decay heat for an extended period of time. This heat will be removed by using the low pressure core cooling pumps to circulate water through the core and the.RER heat exchangers. The low pressure core cooling systems are designed so that even if a single component of the ECCS fails, !
i adequate long term cooling of the core will be maintained.
-( The analysis was done with no deviation from the evaluation model described in Appendix A, Part 2 of the AEC Interim Policy Statement. The design meets the requirements of the AEC Interim Acceptance Criteria. These criteria require that the consequences of a loss-of-coolant accident are such that (a) the calculated maximum fuel rod cladding does not exceed 23000F, (b) the amount of fuel rod cladding that reacts chemically with water or steam does not exceed one percent of the total amount of cladding in the reactor, (c) the clad temperature transient is terminated at a time when the core geometry is still amenable to cooling and before the cladding is so embrittled as to fail during or af ter quenching, l
and (d) the core temperature is reduced and decay heat is removed '
for an extended period of time.
k ,.
- M'M e e same h ,gg.
r-- l
L 6-57
(
It is required that GE incorporate by Amendment to GESSAR a 1
re-analysis of LOCA using methods approved by the Regulatory staff I as meeting requirements outlined in Appendix K, Acceptance Criteria for ECCS, published in the Federal Register January 4, 1974. This 1
amendment must be submitted and approved before submittal of the first BWR/6 FSAR (to which the new Acceptance Criteria apply), and our evaluation will be provided in a supplement to this SER.
6.4 Habitability Systems a
The applicant proposes to meet General Design Criterion No. 19, Control Room, of Appendix A to 10 CFR Part 50, by use of concrete shielding and by installing a dual fresh-air inlet system containing redundant six-inch-bed charconi filters. This represents a system
(
modification in response to a staff position that the initially proposed emergency ventilation system was inadequate.
The dual inlets are widely spaced to reduce the possibility of having both inlets exposed to contamination at the same time.
Under most meteorological conditions one of the inlets will be free of contamination resulting from a postulated activity release.
The free inlet will be used to supply sufficient make-up air to maintain the control room at a positive pressure.
The staff believes that properly positioned inlets can provide the necessary degree of assurence.against the possibility of both inlets being severely contaminated. The applicant has indicated the following locations:
k
u.
fi
.t.
6 ,
.('
Inlet 1: About 150 feet from the SGTS vent on the outermost wall of the control building.
Inlet 2:
About-200 feet from the SGTS vent near the outer edge of the auxiliary building (separated from Inlet 1 by about 240 feet).
This positioning is considered adequate for a single unit facility. There are several specific wind directions where con-tamination of both inlets night occur. However, the concentration at the operating inlet should be about an order of magnitude lower a
than comparable plants having only a single inlet.
o We conclude that, on a generic basis, the GESSAR design for control room habitability meets GDC-19 with the possibic exception
' of the dual inlet locations on multiple unit plants. Since GESSAR is for a single unit plant, approval of inlet locations will be done on a case-by-case basis for multiple unit plants.
The control room design does not specifically protect against a release of toxic gas such as chlorine. Thus, the hazardous material concerns as expressed in Regulatory Guide 1.78 will be reviewed on a case-by-case basis. The control room isolation system may, therefore, need to be modified for any site where gaseous chlorine is used for treatment of the circulating water or where substantial quantities of other hazardous gases are stored in the vicinity.
s
+1
6-59
.(.
6.5 ECCS Pump Suction Strainers Each ECCS pump draws water from the suppression pool through its own suction line and strainer assembly. The succion line ends in a tee arrangement in the pool, with each end of the tee capped with a 100% flow capacity strainer. GE has demonstrated that the potential for ECCS or containment heat removal system degradation due to plugging of the screens is minimal. The reasons for this include, 1)
GE has stated that all insulation in the drywell wfl1 be of such type that it minimizes ' the possibility of 'it breaking away from piping and being carried through the drywell vent system into the s .
1 suppression pool, 2) Since the suction inlets are located about midway between the pool surface and pool bottom and since the screen surface area is large, resulting in low approach velocities, !
there is little potential for drawing debris, either from the pool bottom or surface, to the vicinity of the inlet lines, and 3) due to the ramshead arrangement on each suction line, a 50% plugging of screen surface area can be tolerated without consequence to system performance. Therefore, their design is acceptable.
l e
e
- thetimiegp* sew ep
q l
t l
l l
l 7-1 1 .
( 7.0 INSTRUMENTATION AND CONTROLS k
This section will be provided in an SER Supplement.
l l
i 4
)
C 4
4 es t
1.
M
b 8-1 8.0 ELECTRIC POWER This section vill be provided in an SER Supplement.
i f
a r
e C
. I
(
l
. i
( .
O I
I
l 1
9-1 I
( 9.0 AUXILIARY SYSTEMS The auxiliary systems were evaluated for any safety function requirements and for interf ace data that would be used to design the balance of plant for any applicant referencing GESSAR.
The systems that would normally .be considered having a safety ,
1 function that are required for safe plant shutdown or prevention and mitigation of accidents are not all covered by the scope of GESSAR. As a result we have reviewed those systems where sufficient i information is available and our evaluations are set forth in the d following sections.
.? l l
Those systems that are described but do not have either adequate l P6I diagrams, interface information or safety evaluations incorporated are the Fuel Handling System, the Demineralized Makeup Water System, the Plant Chilled Water System, the Equipment and Floor Drainage I
System, and the Fire Protect' ion System. We cannot perform an evalu- ]
ation until this information is available. We will report further on these systems in a supplement to this SER. GESSAR also states that )
several diesel-generator auxiliary systems may or may not be within the scope of the Standard Plant Design. For these systems, we have not performed an evaluation. They will be reviewed on a case-by-case basis.
1 1
l
9-2
( 9.1 Fuel Storage and Handling 9.1.1 New Fuel Storage The description and evaluation of the new dry fuel storage vault, Section 9.1.1 of GESSAR, in conjunction with the general arrangement i drawings Figures 1.2-2, 1.2-3, 1.2-4, 1.2-5, 1.2-6, 1.2-7 and 1.2-8 were reviewed as the' basis for our evaluation. f The general arrangement drawings indicate the new fuel storage facility will be housed in the Fuel Building. The Fuel Building and the Auxiliary Building surround the Reactor Building and all form a portion of the Reactor Island Facility. The storage racks will be a
designed such that the spacing between fuel elements will assure K,ff will not exceed 0.90 when fuel is stored dry nor result in K,ff ex-ceeding 0.95 should the vault be flooded. To prevent the accumulation
('
of water in the vault, a drain will be provided to direct any entering water to the equipment drain subsystem. The storage capacity and number of storage racks will vary depending upon the design cycle time of the fuel. If the applicant selects an 18 month fuel cycle, the new fuel storage facility will be capable of containing 42 percent of ;
full core fuel or 308 fuel assemblies. If a 12 month fuel cycle is selected, storage capacity equal to 30 percent of a full core or 220 j fuel assemblies will be provided. 1 The fuel elements are loaded into the racks by lowering them through holes in the top of the rack using a general purpose grapple 1
1
~
( -
1
~
m
9-3
(
and the five ton general purpose building crane. The racks are designed to meet seismic Category I requirements. Handling of any other loads above the stored new fuel other than a new fuel assembly shall be pro-hibited by administrative controls. Procedural controls will limit the handling to one fuel assembly at a time and at a height not to exceed 2 feet above the top of the racks.
The height of the storage racks is such that the fuel bundles extend above the top of the racks thereby assuring that the grapple cannot engage the fuel rack and accidentally impose uplif t forces on the rack eastings ,during the handling of fuel. Hold down bolts are used to re-strain the racks and maintain their spacing during the SSE.
No considerations have been given to the concerns relating to sharing
( of the storage facility since only one unit is considered in this Safety Evaluation Report.
Based on the arrangement concept presented in Figures 1,2-2 through 1.2-8 and our evaluation of the design, we conclude that the criteria and bases for the new fOel storage facility design are acceptable.
9.1.2 Spent Fuel Storage The description and evaluation of the spent fuel storage facility, Section 9.1.2 of GESSAR and the general arrangement drawings, Figures
- 1. 2-2 , 1. 2-3 , 1. 2-4 , 1. 2-7, 1. 2-8 & Figure 9.1-2 were reviewed as the basis for our evaluation.
The general arrangement drawings indicate the spent fuel storage facility is housed within the fuel building.
I k l
l
6.
9-4
(
The top entry spent fuel storage racks will be designed to main-tain space geometry that precludes criticality under normal and abnor-mal conditions. A full fuel array of spent fuel in the storsge rackr will be maintained to less than a K,ff of 0.90 with a normal water level of 25 feet of water above the stored fuel. Under abnormal condi-tions, such as an accidental dropping of equipment or other evth.s causing horizontal movement of the fuel, K,ff will not exceed 0.95.
The racks are bolted to the rack support structure in a fashion to facilitate rack rearrt.gement or replacement without draining the
-pool. The holddown bolts maintain the minimum spacing of the racks for geometric reactivity control. Each rack is capable of storing 10 fuel assemblies. The design precludes the accidental insertion of fuel .
( bundles between tacks and will be designed to meet seismic Category I requirements.
The spent fuel storage pool will have sufficient storage rack capacity for 117 percent of one full core fuel load, if the applicant selects an 18 month fuc1 cycle and 105% of a full core if a 12 month fuel cycle is selected. In either case, temporary storage space for 25 percent of one full core fuel load will be provided in a storage pool in the containment. The number of fuel assemblies in one full core fuel load depends upon the choice of either an 18 month or 12 month fuel cycle (308 elements for an 18 month cycle and 220 elements for a 12 month cycle). The total storage space for fuel assemblies
- e~. . .
I 9-5 in the pool _is 857 (for an 18 month cycle) and 769 (for a 12 month cycle). The core consists of 732 fuel assemblies. It will not be possible to remove and store the entire core in the spent fuel storage pool while the previous refueling load is stored without using the storage racks in the containment pool.
Handling of any loads other than one fuel bundle or control rod assembly over the spent fuel storage array is prohibited by administra-tive control for the small, general purpose crane and by structural bar- I riers for the cask crane. The height of lif t of the general purpose crane will be limited to 2 feet above the top of the spent fuel storage racks.
a The fuel building and fuel storage facilities will be designed to seismic Category I requirements. The concrete sides and roof of the fuel building will be designed to prevent tornado borne missiles from
( damaging the stored fuel and permit maintaining a slight negative pres-sure in the building by the heating and ventilation system.
Stainless steel liner plates seal the interior pool surfaces.
Plate joints are fitted with leak chases that direct leakage to a sump as well as allowing testing and monitoring the leak tightness of )
the plant joints. l t
Based on the information provided, we conclude that the criteria and bases for the spent fuel storage facility form the basis for an acceptable design.
9.1.3 Syent Fuel Cooling and Cleanup System I The description and evaluation of the spent fuel pool cooling and I
1 l
I
~
l
_ _ __ _a
i a..
9-6
(
cleanup system, Section 9.1.3 of GESSAR and Figures 9.1.3a through 9.1.3c were reviewed as the basis for our evaluation.
The spent fuel pool cooling and cleanup system will be designed to remove the decay heat from the spent fuel assemblies and maintain pool water level in the containment pool and fuel building pools. Limiting the radioactivity concentration, the corrosion product buildup in the water and maintaining its clarity will also be accomplished by the cleanup system. '
The cooling system will be designed to be capable of maintaining
-the pool . water temperature below 125'F under normal operating condi-tions and maximum normal heat load. The maximum normal heat load has been defined as the sum of the decay heat released by the average
( spent fuel batch discharged from the equilibrium fuel cycle at the earliest refueling time plus the decay heat being released by the batch discharged at the previous refueling. In addition, the system is capable of being lined up to and removing the heat released to the containment pool through the drywell head, t The applicant proposed to supplement the above cooling system when larger ' heat loads would exist if more than the above anticipated l I
number of fuel elements are in storage by utilizing the RHR system. j i
If the pool water temperature exceeds the stated 125"F, the RHR sys-tems will be used to supplement the available spent fuel cooling system.
We will require, in the Technical Specifications of GESSAR users, for
( ;
-. ~
9-7
( this mode of operation that the reactor plant be placed and maintained in a cold shutdown condition and in the refueling mode as long as the RER system is used to assist in cooling the spent fuel pool water. GE has agreed to place the plant in a. shutdown condition when the RHR is used for supplemental pool cooling.
The heat from the spent fuel pool cooling system and the RHR system 3 will be removed by the station service water system.
Table 3.2.1 of GESSAR indicates that, with the exception of the makeup water supply and the filter / demineralized vessels, the spent
. duel pool cooling and cleanup system will be designed to the positions set forth in Regulatory Guides i.26, " Quality Group Classifications and Standards," and 1.29, " Seismic Design Classification."
The applicant proposes to use the RRR system interconnection to
-(
the fuel pool cooling system to provide a seismic Category I makeup source. We will require that some other system be provided to perform the seismic Category I backup makeup function to be in accordance with our position set forth*in Regulatory Guide 1.13, " Fuel Storage Facility Design Basis." The RHR system is not considered acceptable to meet fuel pool makeup requirements and perform its other functions.
Based on the information provided for normal heat removal and the indication that normal operation for the spent fuel pool cooling system
. is based on an interface inlet station cooling water temperature of 100*F, we conclude that the design criteri.i and bases are acceptable
(
s 9-8 for normal operation. Since GE has not indicated whether 100*F is the
. maximum station cooling water temperature for all plant conditions, even i
loss of the normal heat sink, we connot conclude that the design basis and criteria are acceptable for abnormal operation.
9.2 Water Systems The auxiliary water systems in GESSAR consist of portions of the service water system, the closed cooling water system, the demineralized water system, the potable and sanitary water system, the plant chilled water system and the condensate storage facilities. Many of these systems are designed entirely by the utility applicant and perform no safety rela'ted function and are therefore not appropriate to GESSAR.
Others like the essential service water system do perform a safety related function, but are not entirely within the scope of GESSAR. The GESSAR C. described parts of the demineralized makeup water system, the plant chilled water system and the service water system that do not have adequate P&I diagrams, safety evaluations or interface information provided to permit i
us to complete our evaluation of these systems at this time. We will )!
I report further on these systems in a supplement to this SER prior to j i
issuance of the PDA.
9.3 Process Auxiliaries The process auxiliaries in GESSAR include the compressed air system, the process sampling system, the equipment and floor drainage system, the liquid poison system, the failed fuel detection system, and the main l
l
(
l l
~
L 9-9 steam isolation valve leakage control system. Several of these systems are described in other sections of this SER. As discussed above, other systems do not have adequate information to permit us to perform our evaluation at this time. When further information is provided, we will review it and report the results of our review in a supplement to this SER.
9.3.1 Main Steam Isolation valve Leakage control System (MSIVLCS)
The description and design criteria and evaluation for the MSIVLCS, Section 9.3.6 and Section 3.2 and Figures 9.3-7a and 9.3-7b were reviewed as the basis for our evaluation.
The MSIVLCS information submitted is intended to meet the l
guidelines set forth in our request to the applicant. The MSIVLCS
- consists 'fo 2 independent systems to accomplish the leakage control function. The inboard system is powered from one electrical division and the outboard from the other division of the emergency power supply.
C The outboard system is connected to all steam lines between the outboard I
MSIV and the downstream block valve. All steam lines are connected in j l
series to the outboard bleed header. Suction is drawn on the space between '
the valves and exhausted to a building volume served by the standby gas treatment system. The inboard system is connected between the main steam isolation valves. An individual bleed line is provided for each j l
steam line. The exhaust is vented to a building volume served by the !
. SGTS. The systems will both be manually operated following a LOCA when the steam line pressure is below the pressure interlock setpoint. In the event of high MSIV leakage, the individual inboard system will automatically close.' In the event of excessive leakage into the outboard system, a high pressure alarm will be annunciated.
k (
F l l
k 9-10
( The system design will be based on the two main steam line isolation valves and block valve in each steam line being closed and that a nega-tive pressure can be maintained in this closed volume by fans that discharge ultimately to the standby gas treatment system.
The two isolation valves will be those provided for containment isolation (one inside and one outside containment) and are fart acting valves. The third valve is a block valve located at the end of the steam line pipe tunnel. The entire main steam system including piping, isolation valves and branch lines will be designed to seismic Category i requirements.
The HSLIVLCS will also be designed to seismic Category I require-ments and protected from internally generated missiles and environmental
( conditions consistent with a design basis loss of coolant accident.
The applicant has performed a single active failure analysis that l l
indicates that the stated design criteria for the system will meet the singic failure criteria. We have not completed our review of this I
system. However, several concerns, ,such as stem leakage, mixing of the I leakage in the annulus and actuation within 20 minutes have risen and are being pursued with GE. We will report on the r'esults of our completed l review in a supplement to this SER. '
9.4 Air Conditioning, Heating, Cooling and Ventilating Systems 1
The description of the control room HVAC, Section 6.4 of GESSAR, and l Figure 6.4-1 were reviewed as the basis for our evaluation.
( .
8 9-11 d
The control room HVAC consists of two subsystems which will be provided to assure normal outside air HVAC and minimum make-up air. j cleaning and control room pressurization. The functional operation of the normal HVAC provides filtered, heated or cooled outside air to the control room and the control equipment room and maintains these rooms at !
a positive pressure with respect to the ambient barometric pressure, j The functional operation of the minimum make-up air cleaning system 'is to discharge or recirculate normal ventilation air through the two rooms. During conditions of high radioactivity at the outside air i
'intakes, the system will recirculate the room air through a filter unit containing particulate and charcoal filters. A small amount of outside I air is taken from either of two air intakes and processed through the filter assemblies and air conditioner unit to maintain a positive ,
pressure in the rooms. The system can be completely isolated from the I
outside if required for a sufficient period of time to protect personnel l
~
during extreme adverse outside air contamination. I Another function of the system is to detect smoke in the outside 1 \
air intakes and in the system ducting to provide for isolation of the outside air intakes, open recirculation dampers or switch to a once through purge operation of a given area of the rooms or the equipment racks by automatic operation of the necessary dampers.
8 l
9-12
, {. ' The normal HVAC ' system' takes outside air from a missile protected
'i intake located on the far side of the control building located.near grade. 1
.1 p elevation (+ 1'0") where it is processed through one of two 100% capacity
~ unit air conditioning systems. This processed air ventilates the control room, control equipment room, and the HVAC equipment room. Two 100% !
capacity room return air fans are provided to discharge the air to the building exhaust or recirculate a portion back to the ' air inlet processing =
. system. A positive pressure is always maintained by mixing dampers in the recirculation ducting. Two 100% capacity equipment rack return air fans are provided to perform the same function.
a The minimum make-up air cleaning system consists of two 100% capacity.
air filtration units, a missile protected outside air intake located on the auxiliary building side at elevation (35' - 0"), the two unit air
(- conditioning units of the normal system and two sets of return air fans J
of the normal system. On a high radiation signal this system is placed
'in operation and the normal system outside air intakes and exhaust air .
l I
isolated.
The applicant has performed a single failure analysis assuming a failure of any individual component in this complete system. We have reviewed the results of this analysis and find the system can withstand any single active failure and perform its safety function.
Based on the above, we conclude the design criteria and bases meets the i
requirements of AEC General Design Criterion 19 and the applicable portions 1 s of the position set forth in Regulatory Guide 1.52 and are therefore acceptable .
k .'
m
. ~
b 9-13
(
9.4.1.1 Balance of Control Building.(BCBHVAC)
The description and evaluation of the BCBHVAC, Section 9.4.1 of the GESSAR and Figure 9.4-1 were reviewed as the basis for our evaluation.
The BCBHVAC consists of the main control building outside air in-take, three inlet filter fan assemblies with two being provided with heating coils and the other a cooling coil, three exhaust fans and assoc-lated ducting. The system serves all areas of the control building except the control room and the. control equipment room.
The applicant states that the system is not required to perform a a
safety function. Our review ofJthe system indicates this is a correct statement.except for the isolation dampers that isolate the system"frc=
thecontrolroomoutsideairinbakesandthedampersinthecommon
( ,
discharge vent from the control building. Due to the interaction be-tween the control room HVAC and the BCBHVAC we will require these iso-lation dampers to be designed to seismic Category I requirements.
Based on the above, and considering our requirement will be incor-porated in the design criteria, we conclude the design criteria and bases for the BCBHVAC are acceptable.
9.4.2 Auxiliary Building (ABHVAC)
- The description and evaluation of the ABHVAC, Section 9.4.2 of GESSAR and Figure 9.4-2 were reviewed as the basis for our evaluation.
The ABHVAC consists of a normal HVAC system and an emergency oper-ating condition system. The functional operation of the normal HVAC 2
i 9-14 i i'
system is to provide ventilation air to the reactor water cleanup pump I rooms, the ECCS pump rooms (RHR pumps A, B,. & C, LPCS pump, HPCS pump and the RCIC pump). Single zone HVAC systems will be provided to pro-vide ventilation air to the electrical area and corridors and battery room. The steam tunnel is provided with a cooling system to maintain the tunnel space at 125"F during normal plant operation. This temperature is 10*F below that at which the 3000 cycle test of the FSLIV control valves was conducted (APED-5750 supplement 1) . l i
The normal HVAC system consists of two outside air filter fan units supplied by a single outside air intake, series isolation dampers in the ducting of the two main areas it serves, two variable vane ex-a haust fans and series isolation dampers in the main exhaust duct. The system normally operates with one outside air filter fan unit and one
{ exhaus t f an unit . The other fans are provided for standby service.
All of the above areas are maintained at a negative pressure by the variable vane exhaust fan.
Each of the two single zone HVAC systems consists of a missile pro-tected outside air intake, particulate filter, heating coil, two cooling coils, an intake fan, ducting and an exhaust fan, located down-stream of the battery rooms. Each system is capable of ventilating the three areas it serves.
Our review of the normal ABHVAC system, including an independent failure mcde and effects analysis, indicates that the system can per-form its normal function with the exception of single damper failures which could require the system to be placed in the emergency mode of operation for the areas where equipment is operating. The single zone i
(
l .
9-15
. /
l tion valves at theinstrumentsair inlet or i HVAC system does'not contain iso awe he will' require that detec applicant exhaust. !
the battery room vided for these areas unless he areas fromt and isolation dampers be pronot required for protecting t can demonstrate that they ared ogen from the battery room.
radioactivity, smoke or hy r l system is isolated of coolant accident the norma l main-During the loss Gas Treatment System (SGTS) wil y
by series dampers and the Standby tive pressure and filter an The ent.
tain the essential pump rooms at a negato discharge SER. Each pump to the en
' radioactivity from the air prior Sections 6.2.3 and 11.3 ource of this SGTS evaluation is presented in it which is powered from the sa room has its own fan cooling un SGTS in-as the pump it cools.
~
mode of operation using the i p to the con-
- Our review of the emergencyfailure mode and effe dicate that a ciuded an independent results of this analysis in The water either reactor nection to the SGTS. h exhaust duct from rooms from being single failure of a valve in t eld prevent onebeofmade the i
cleanup system pump We rooms kouwill require that will not a modificat connected to the SGTS, assure that a single failure tem to to this portion of the sys in emergency ventilation funct o . f the required dampers preclude the tabic.
Based on the above and the incorporation oi n criteria
. system, we conclude the des g to the
. pC
(
u 9-16
(
9.4.3 Radwaste Building (RWBHVAC)
The description and evaluation of the RWBHVAC, Section 9.4.3 and Figure 9.4-3 were reviewed as the basis for our evaluation. .,
The RWBHVAC system will be provided to ventilate the radwaste con-trol station, the radwaste storage cells and non-contaminated areas. The system will maintain the storage cells and the non-contaminated area at a negative pressure and the control station at a positive pressure.
The system will be a once-through type consisting of a roughing i 1
filter, heating and cooling coils, two 100% supply fans and associated j
d uctwork.- The system will be exhausted to the plant vent by two 100%
capacity exhaust fans. The control station will be exhausted separ-ately to permit maintaining the required positive pressure. l
( Our review of the system shows that the dual fan system will be 1
capable of ventilating the radwaste area and single active failures will not preclude its function. The storage cell area will be isolated l by series dampers if activity is detected.
Based on the above, we conclude the system design criteria and bases are acceptable.
9.4.4 Fuel Building (FBHVACl The description and evaluation of the FBHVAC, Section 9.4.6 of GESSAR and Figure 9.4-6a were reviewed as the basis for our evaluation.
The FBHVAC system consists of four subsystems to control the fuel building and standby gas treatment system equipment rooms at a negative pressure and to process any radioactivity that could be released due to
( .
l 9-17 9
a fuel handling accident. The subsystems are (1) the dual outside air pressure supply fan units, (2) two 50% capacity fuel building recir- l l
culation heating and cooling fan units; (3) the dual variable vane exhaust fans which normally exhaust the above mentioned areas and con-trol them at a negative pressure, and (4) the standby gas treatment system which will provide for filtration of radioactivity from the above areas and maintain them at a negative pressure during accident conditions.
Our review of the integrated system, including an independent failure mode and effects analysis, indicates that a single failure of a, valve will preclude the SGTS from performing its function for the fuel building. In addition, the radiation monitors. provided in the
[c system should be , located in the fuel building instead of in the exhaust ducting. The ducting and dampers associated with the performance of the SGTS function for all the areas should be designed to seismic Cate-gory I requirements as set forth in the positions of Regulatory Guides 1.13 and 1.52. We will require that modifications to the design cri-teria be made for each of the above mentioned deficiencies.
The SGTS evaluation is given in Sections 6.2.3 and 11.3 of this SER.
Based on the above and considering that our requirements will be incorporated in the design criteria, we conclude the design criteria and bases for the IBHVAC meet the positions set forth in Regulatory Guides 1.13 and 1.52 and are therefore acceptable.
s
s -
, s 9-18
( ,
9.5 Fire Protection ,
The GESSAR information provided is not adequate to perform an evaluation of _ fire detection, protection and design data. The application states (Table 1.10-2) that those systems within the scope of GESSAR will be designed to Regulatory Guf.de 1.70.4. This guide only 'requesto the information that should be contained in an application and does not set forth any Regulatory Positions. !
l The information contained in GESSAR does give some understanding.
of the possible design approaches that may be taken but is inadequate for us to make a conclusion.,
.( .
e e
4 e-O
-..m_.
l s i
l .
l l
10-1 l
l
('t l 10.0 POWER CONVERSION SYSTEM l
This section of the plant design will be reviewed for each appli-cation that references GESSAR.
i e
a
~
, .~
(
e t .
"7 l 11-1 r,g C CiV E D l' .' C ',.S ' 1.00.
11.0 RADI0ACTIVC WASTE 11ANAGEMENT -$ -
bVI \gy" PP 11.1 Summary,_ Dese ription [ ] T T ,i 3.1t h 10 L The radioactive waste management system for the General Electric ~
Standard Plant (CESSAR) will be designed to provide for the controlled handling and treatment of liquid, gaseous and solid radioactive wastes, and for monitoring of all major release points of radioactive materials.
The liquid waste system will process liquid wastes from such sources as equipecut drains, system leakage, condensate domineralizer regen-erant solutions, laboratory and decontamination liquids, and detergent auastos, ,The liquid waste vill be procepsed.and recycled for reesc if '
the plant water balance requires makeup and if the water quality is h na q ud.e .
Caseous wastes will consist of offgases from the main condenser I- .
air ejector, vents from equipment containing radioactive materials, i
and leahoge from systems and components containing radioactive material that is released via the building ventilation systems. The offgases from the main condenser air ejector will be treated by catalytic recon-bination to reduce the volume of offgases and by charcoal adsorption l
to selectively delay fission product noble gases before relcase to the environment. Certain equipment vents and building ventilation exhausts a will be treated by high efficiency particulate air (HEPA) filters and charcoal adsorbers to remove radioactive particulate and radiolodine i prior to release to the atmosphere.
, c.nyg q -
re g g ar P
- .. v a , -a 27
!D 3
T kLW/WDL 'J J A U w
, Jo No: Remove" rom ACRS05ce
__ _______________ __ O
11-2 Solid wastes generated during plant operation will consist of spent demineralized resins, evaporator bottoms, discarded radioactive components and tools, and miscellaneous dry solid wastes. Wet solid
. wastes will be solidified and drummed. Dry, compressible materials will be compacted and drummed. Drummed wastes will be shipped to a licensed burial site.
The CESSAR design objective, which is similar to other systems pre-viously approved, is to restrict the amount of radioactivity released from. normal plant operations to unrestricted areas to within.the design dobjectives, set forth in Appendix I to 10,.CFR Part 50.
The following sections present our evaluation of the liquid, i
b ;
gaseous, and solid radincetive wante trcctment cyctemc. Tha liquid, gaseous, and solid waste systems will be designed to accommodate the waste produced during operation of a singic unit at a maximum thermal power level of 3758 FWt. The radwaste treatment system, as discussed in GESSAR, includes provisions to process wastes produced by plants employing either regendrable deep bed condensate demineralized units (with optional ultrasonic cleaning of resins), or by plants employing l
powdered (Powdex) type filter-demineralizers for condensate polishing. )
l Our evaluation and calculation of annual releases of radioactive
! materials are based on the parameters and calculational models given in Appendices B and C to WASil-1258. j l
l
(
1
_ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ - )
i
~ ;
j l
1 l
11-3 1
l .
11.2 Liquid Waste Treatment Systems 11.2.1 System Description l .
Treatment of the liquid waste is dependent on the source, activity l and composition of the particular. liquid waste and on the intended dis-posal procedure. The liquid waste system is divided into three sub-l systems: the waste collector subsystem (low conductivity), the floor drain neutralizer subsystem (high conductivity), and the detergent waste subsystem.
The liquid raduaste treatment systems are designed to permit complete
" recycle of processed liquids during normal operation. Processed liquids will be handled on a batch basis to permit optimum control and release I
+
of radioactive materiais. In (the event of the release of treated
, liquid vastes, samples will be analyzed to determine the type and amount of radioactivity in each batch. Based on the analytical resultc, these wastes will either be released through the circulating water discharge or processed through the detergent evaporator and released as vapor.
In our evaluation, we considered that processed high purity and low purity liquid wastes would be released as liquids, and that processed detergent wastec would be released as vapor.
The waste collector (low conductivity) subsystem receives liquids from t!'e drywell, containment building, auxiliary building, fuel build-ing, radwaste building and turbine building equipment drains and the decantate from the cleanup phase separator. Depending on whether the
(
l 4
! 11-4 I
design of the condensate polishing system uses either the deep bed or !
e Powdex demineralizers, the waste collector subsystem will also receive I
f the decantate from the condensate phase separators (Powdex systems),
l condensate domineralizer wash water, and ultra-sonic resin cleaning l rinse water (deep bed systems). These wastes will be collected in one i
of two 57,000 gallon low conductivity tanks, processed in a batch through a traveling bed filter, and collected in a 3,000 gallon filtrate i
tank, l
i The filtered wastes vill then be processed through two 350 gpm mixed
"' bed demineralizers in series and will normally be routed to the con-f densate storage tank for reuse in the reactor. If storage capacity is l l
jt - not available, the processed liquids will be manually diverted to one !
of two 50,000 gallon excess water tanks. From the excesc water storage ]
i j tanks, the liquids may be sent to condensate storage, released at a con-l I
- trolled rate to the discharge canal, or routed to the detergent evapora- !
f tor and t. hen relea sed as vapor, General Electric expects that all waste l l
! treated through this system will be recycled for use within the plant,
{
i but has evaluated the releases based on 10% discharge. In our evaluation,
)
< I i
l we considered that 34,000 gpd of low conductivity liquids will be l i
processed through the system and that 10% of the treated liquids will be I discharged to the environment.
The floor drain neutralizer (high conductivity) subsystem collects liquids from the drywell, containment, auxiliary building, fuel building, I
(
11-5 t
l l
radwaste building and turbine building floor sumps, demineralized regenerant solutions, laboratory drains and non-detergent decontamina-tion solutions. The wastes will be collected in one of three 18,000 gallon high conductivity tanks, neutralized, and processed through a 40 gpm waste evaporator. The evaporator condensate will be processed through a 350 gpm distillate demineralized, a 350 gpm backup demineralized, i
and recycled to condensate storage. If the condensate storage tanks
.! are full, the processed liquids will be routed to the excess water tanks.
I From the excess water tanks, wat,er may be either recycled to condensate ,
l -
\
l dstorage,. released to the discharge canal, or routed to the detergent i evaporator and released as vapor. The applicant estimates that 9300 I.
,g upd of high conductivity liquids will be processed through this subsystem, ,
I I
and that 10% of the processed 11gui,ds will be released to the discharge canal. Our analysis is based on an input rate of 8200 gpd and discharge of 10% of the treated liquids to 'the environment. The applicant's i
estimates of liquids are based on his projections of the design j capability . Our estimates are based on WASH-1258.
l Laboratory wash water and wastes containing detergents (laundry,'
1 personnel and equipment decontamination wastes) will be collected in a i
1500 gallon detergent waste tank. These wastes will be filtered and evaporated in a 5 gpm detergent evaporator, wh1ch is vented to the I
1 atmosphere. The applicant considers that 1100 gpd of detergent waste will be processed through this system, with 100% of the evaporator distillate released to the atmocphere. Our analysis is based on 450
'm_ .-
e
i 4
~
11-6
(
gpd of detergent wastes being processed, with 100% release of the l processed-liquids to the atmosphere, j Bottoms from the waste evaporator and the detergent evaporator will
, be collected in a 25,000 gallon concentrated waste' tank and transferred to the solid uaste system for solidification, packaging and shipment offsite. Spent demineralized resins will be collected in a 10,000 gallon spent resin tank and transferred to the solid waste system for 1
packaging and shipment offsite. Resin sludges from the phase separator tanks will be dewatered and transferred with the filter sludges to the a solid wa.s.te system for loading into shi,pping containers.
- 11. '2 . 2 - Liquid Waste System Evaluation b Using Lhe mulhuds and parisotcrs of RASU-1253, we have calculeted the liquid waste releases. GE has also calculated what the liquid L
i l releases will be and their values are presented in parentheses following ours. In general, the liquid releases will be less than 1 Ci/yr/ reactor (GE - 4.6 Ci/yr), excluding tritium and dissolved gases. We have normalized these relca'ses to 2 Ci/yr to compensate for equipment down-time and anticipated operational occurrences. We also calculate that C1/yr (CE 3.6 x 10 -6
~
6 x 10 Ci/yr) of particulate material, will be released to the atmosphere due to evaporation of detergent wastes.
Based on reported releases at operating boiling water reactors, we calculate that 20 Ci/yr/ reactor (CE 12 to 20 Ci/yr) of tritium will be released in the liquid eff'luents. The principal reason for the
(
11-7
(
applicant 's higher estimate of the quantity of material to be released in liquid effluents is the applicant's use of lower decontamination factors for the demineralizers in the low conductivity system. The ~
bases for our decontamination factors are found in WASH-1258.
All major processing components are redundant. There are spare collector tanks in all three subsystems of the liquid radwaste treatment system.
There is a spare travelling belt filter, a backup demineralized in the ,
waste collector subsystem, and a spare waste evaporator in the floor drain-neutralizer subsystem. In the event of equipment f ailure in the s low conductivity system, it will be pos,sibic to divert the wastes to
, the high conductivity tank. The only equipment items that are not l redum' ant in t! ic cyctu.T are the detergent vacte filter and the deter-gent vaste evaporator. If, in the event one of these equipment items i -
is unavailabic, detergent vastes will be released without treatment, the liquid effluents would be increased by 0.06 Ci/yr, which is a small fraction of our calculated total release from the plant.
Overflows from the low conductivity tank, the filtrate tanks, the spent resin tank and the cleanup phase separator tanks are piped to the radwaste building equipment drain sump. Overflows from the detergent drain tank and the high conductivity tank are piped to the radwaste bu'lding i floor drain sump. These provisions are acceptable.
t The applicant has committed to release the processed liquids to the discharge canal through a line that is equipped with a radiation
(
-_. - , , _ m % Tv * '.neg anstgunse.- __ - n_ __
_ ammspee
_.-_____,._____a
11-8 monitor that will alarm and automatically terminate the release if the .
1
, concentration of radioactive material in the effluent exceeds a pre-determined level. We consider that this design provides adequate control ,
over releases of radioactive materials, in accordance with General Design Criterion 60 of Appendix A to 10 CFR Part 50, and find it acceptabic.
The applicant has indicated that all equipment in the liquid rad-waste treatment systems will be designed to Quality Group D and non-Seismic Category I classification. The equipment will be contained in the radwaste building, which has a seismic Category I substructure.
The seismic design classification of the liquid waste treatment system a .. :
components and structures is acceptable. The system should be designed to the " quality Group D (Augmented)" classification described in Appendix A entitled 'tesign Guidance for Radioactive Waste Management Systems Installed in Light-Water-Cooled Nuclear Power Reactor Plants."
The liquid radioactive vaste system includes the equipment and instrumentation to control the release of radioactive materials in I
liquid effluents. Our review of the liquid radwaste system included '
normal operation, anticipated operational occurrences, design provisions incorporated to preclude uncontrolled releases of radioactive materials in liquids due to leakage or overflows, and the quality group classifi-cation and seismic design criteria. We have reviewed the applicant's system descriptions, process flow diagrans, piping and instrumentation diagrams, 1
l l
, . _ , , , , , _ .m,.,,,_ , . . .
~
i 11-9 l and design criteria for the components of the liquid radwaste treat-l l ment system. We have performed an independent calculation of the
\ '
i releases of radioactive materials in liquid effluents based on the calculational methods of WASH-1258. I 11.2.3 Conclusions Based on our evaluation, we cbnclude that there is sufficient processing capability and flexibi.1'ity in the liquid radwaste system to provide reasonabic assurence that the design objective of limiting annual releases of radioactive materials, excluding tritium and dissolved gases, in liquid effluents resulting from normal operation, including anticipated operational occurrences, to 5 Curies, can be met.
a . .
] There ic 'also reasonable assurance that the concentration of radioactive materials in liquid effluents will be a fraction of the limits in 10 CFR l
I 20, Appendix B, Table II, Colu=n 2, for the expected and design releases, for plants uhece discharge canal flow equals or exceeds the m;nimum discharge canal flow of 1500 gpm specified by the applicant. Adequate control of relcaces of radioactive materials in liquid effluents is provided in accordance with General Design Criterion 60 of Appendix A of 10 CFR Part 50. ;
Compliance with the design objective doses to individuals at or beyond the site boundary is site dependent, and will be reviewed for individual license applications. Houever, based on our review of previous facilities, we expect that the system, as described, will meet ALAP dose considerations
, for the majority of sites within the U.S. The liquid waste treatment system should be cesignea to "q.:ality Group u (Augnentec)" as per the attached Appendix A. The applicant has designed the system to Quality Group D, which is unacceptable. We will report on the recolution of this item in a I I i anpplement to thia SER. I
{
l
11-10 11.3 Gascous Waste Treatment Systen.s t
~11.3.1 System Description e
The gaseous radioactive waste treatment systems for the GESSAR plant
' will consist of a charcoal delay system for treating the offgas from the main condenser air ejector, and iodine and particulate ' control sys-tems for certain building ventilation systems. The release of radio-active materials in the offgas from the turbine gland sealing system ~
will be negligibic, since non-radioactive steam from an auxiliary boiler will be used to seal the turbine glands. During startups, a mechanical vacuum pump will be used to evacuate the main condenser. The, discharge from the vacuum a ,, .
pump will contain radioa,ctive gases that will be released
. without treatment through a roof vent on the turbine building to the atmocphere, along .ith the offnases froa the charcoal delay system, the I
standby gas treatment system and the containment purge. Building ventilation system exhausts are normally released via roof vent. The turbine building ventilation system is outside the scope of GESSAR, and will be reviewed for individual Ifcense applications.
Offgases from the main condenser eir ejector will be treated through a low temperature charcoal delay system to provide for decay of the radio-active noble gases before release to the atmosphere. The offgases from the main condenser will contain principally, hydrogen and oxygen from decomposition of water, air from condenser inleakage, fission and I
(
1 I
I t
11-11 i
activation gases, and water vapor. The offgases from the last stage 1
of the condenser air ejector will be diluted with steam to maintain hydrogen concentrations telow 4%. The pressure boundary of the system is designed to maintain its integrity in the event of a hydrogen explosdon. The gas mixture will be heated, l l
pacsed through a catalytic recombiner to react hydrogen and oxygen and passed through a condenser and moisture separator. The condensate will be returned by gravity to the main condenser hotwell, and gases will be passed through a ten minute delay line to provide for radioactive decay a
of activation and short lived fission product gases. The gases will be cooled to 45 P in a glycol cooler, filtered through a high efficienc-;
particulate air (Hr.PA) filter, and dried to a dcupcint of approximately I
-90 F by a desiccant drier. The gases will be further cooled to approximately 0*F and passed througli a' train of eight three-ton charcoal beds in series. The gases exiting the charcoc1 .<tay train will be filtered to recove charcoal fines and particulate matter, and released to the atmosphere via a roof vent.
In passing through the charcoal delay beds, the higher molecular weight xenon and krypton will be preferentially adsorbed on the charcoal surf ace and delayed with respect to the flow of the carrier gas (air).
The delay time will be affected by a number of variables, such as moisture content of the carrier gas, charcoal type, carrier gas flow rate, poisoning of the bed by impuritics, and by temperature gradient 9 i
I 1 I
l l :
l 11-12 i
in the bed due to fission product decay heating. Based on an air in-leakage rate of 30 scfm into the condenser, with the air dried to a devpoint of -90 F, and beds filled with 8 x 16 mesh coconut base char-l coal at 0*F, GE has estimated delay times of 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> for krypton and 42 days for xenen. The applicant has based his evaluation on the results of small scale experiments that he has performed. The applicant is currently gathering verifying data on a large scale system at the KRB reactor. We have accepted the applicant's adsorption data (dynan.ic adsorption coefficients) pending confirmation by the large j scale experiments', and have based our evaluation on the above delay
. times. Topical Report MEDE-10731-1P, which provides the results of large scale tests, has recently been submitted for our review.
I .
( During startup, a acchan:ecal vacuum pump will be used to evacuate t
the main condencers and the Tf-gases from the mechanical vacuum pump will be released directly to :he environment without treatment through a roof vent.
1 Leakage frca components and systems containing radioactive materialc
$ will be released to the atnosphere via the building ventilation systems.
Volatile radioactive materials will be released to the containment t
building atrnosphere as a result of relief valve actuations and exhausted 5
without triatmenL Lhrough the containment purge system. Building ventilatiorh systems within the scope of GESSAPJ that are sources of radioactive! gaseous ef fluents are the containment purge, the drywell purte, the shield building annulus, the auxiliary building, the fuel
.~ . _ .
11-13 i
building and the radwaste buildings. The turbine building ventilation system is also a source of radioactive gaseous effluents; however, it is not included in the scope of the application and will be reviewed for individual license applications that reference GESSAR.
The containment building air is cooled and recirculated at a rate of 42,500 scfm. In addition, the air in the containment dome is recircu-lated by two 6900 sefm fans. The applicant's design provides for the containment to be at a slight negative pressure by means of a 4300 scfm low volume purge. Fresh makeup air is supplied by a 4300 scfm supply fan. In ,.the event it is required to purge the containment rapidly, the a :
applicant's design also provides 25,000 scfm purge and supply fans.
During norm 1 operation, the drywell is a closed systera with air being recirculated and cooled at a rate of 92,000 sefm. When necessary, the dryuell may be purged via a 4300 scfm drywell purge vent fan. The shield building annulus will be maintained at a negative pressure by redundant 4200 scfm capacity exhaust and recirculating fans. Areas in the auxiliary building ,containing ECCS pumps and heat exchangers will be maintained at a negative pressure via redundant 4200 scfm capacity ,
supply and exhaust fans. The fuel handling areas of the fuel building will be maintained at a slight negative pressure by means of redundant 2000 scfm supply and exhaust fans. The ventilation exhaust air from each of the above areas will be monitored and released to the atmosphere without treatment. In the event radioactivity levels in any area
( l i
i f
~ - - - . - - _ . - .- _. =
11-14 I are above a predeterrained level the exhaust from that area will be diverted to the SCT3.
The radwaste building will be maintained at a slight negative pressure j by means of redundant 17,600 scfm capacity supply and exhaust fans.
The exhaust air vill be filtered thrcugh IIEPA filters and monitored for radioactivity content before release to the atmosphere. In the event of a high radioactivity level in the exhaust stream, an alarm vill sound in the centrol room and the cells containing radwaste equip-ment will be isolated. The exhaust can be terminated by a remote manual u suitch. General Electric does not provide radioiodinc control systems
', for the redwacte building ventilation systen, but proposes to c.mploy l , c':: r eml +.:" ers on vents from selected t e@c. and .*quipnent.
11.3.2 Cascous Ucnte Treatmrt Scstem Evaluation We calculate that the noble gas releases from the CESSAR plant will be 5700 Ci/yr, and that 0.28 Ci/yr of I-131 will be released through the building ventilation systers. The applicant cctinates 5000 Ci/yr of noble gases and 0.16 C1/yr of 1-131 will be released. The applicant 's lowcr estimate of 1-131 releases is based on a smaller rate of steam leakage to the turbine building. Our value for this parameter is 1340 pounds per hour. The basis for our parameter is given in Appendix j B to WASil-1258.
, 3 All major equipment in the off-gas system is redundant. The glycol i
cooler equipment is not redundant, but will be located in a non-radioactive l
l
(
- i 11-15 1-area so that it will be accessible for raaintenance in the event of 1 equipment malfunction.
We find that the off-gas system design has sufficient redundancy to provide reasonable assurance that the system will have the capability to perform its intended function.
l The applicant's design provides that, in the event of radioactivity above a predetermined level in the exhausts from the containment l purge (4300 scfm), the drywell purge (4300 scfm), the ECCS pump rooms (4200 scfm) and the fuel building (2000 scfm), the exhausts will be l directed to the SCTS. The combined capacity of these ventilation systems i a f ar exceeds the treatment capability of the SGTS (redundant 5000 scfm trains),
j It is our position that additiokal means of reducing the re1 cases of l radivaetive materlulu in gasenn:j offlunnra from r'nn plant venriinrion systems need to be provided. he also find that continuous purging of the containment directly to the environment without treatment is unacceptable.
Either an internal recirculation iodine cleanup system for containment attacaphere cleanup before purging or a charcoal filtration system for continuous operation during purging is required.
All equipment in the main condenser offgas system is proposed to be designed to Quality Group D and non-seismic Category I standards. The equipment will be contained in the turbine building, a non-seismic Category I structure. We consider these design criteria unacceptabic.
_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ - _ - _ _ ~
11-16 -
l The system should be designed to " Quality Group D (Augnented)"
classification cnd portions of the charcoal delay system downstream of the delay line should be designed to seismic Category I and should be contained in a scismic Category I structure. Our position is given in Appendix A attached to this report. We will report on the resolution I
j of this item in a supplement to the SER.
i Offgases from the charcoal delay system, the SGTS, the containment purge and the mechamical vacuum pump will be released to the atmosphere via a roof vent on the turbine building. The design of this vent is outcide the scope of GESSAR and will be revieued on a case-by-case "basiswhen'applicationsreferencingGESdARarereceived. The major inputs to this roof vent will be monitored and controlled 1 individually I .
I and the release of radioactive materials from the vent will be i monitored. The offgaces from the charcoal delay beds vill be monitored and the release automatically terminated if the radioactivity exceeds a predetermined level. The radioactivity in the SGTS exhaust will be monitored und will annunciate radioactivity in excess of predetermined l
levels in the discharge from the SGIS.
We find that the applicant's design provides adequate control of releases of radioactive materials in gaseous effluents from the condenser offgas system, in accordance with General Design Criterion 60 of Appendix .
t
, A to 10 CFR Part 50. However, provisions to control releases from building ventilation systems are unacceptable, as discussed above.
The provisions for monitoring and control of releases of radioactive materials in gaseous effluents f rom
(
- - . - - - - - _ _ _ _ - _ ~
11-17 I
the turbine building is outside the scope of GESSAR, and will be evalu-ated on a case-by-case basis when applications referencing GESSAR are received. -
11.3.3 Conclusions The Gascous Waste Treatment Systems include process equipment and associated instrumentation to collect, store, handic, process and con-trol releases of radioactive materials in gaseous effluents produced as the result of operation of the GESSAR standard plant. The gaseous waste systems include all plant systems that have a potential to relsase a radioactive materials in gaseous effluent to the environment, including i
building ventilation systems. Our scope of review of the gaseous waste )
l l trea t: e::t eyt t-: - has included the app 12 cant's system deccriptionn, I
schematic flow diagrar.:s and piping and instrumentation diagrams (P&ID's), )
l the applicar t's design objectives for relcaces of radioactive materials during norcil operations, including anticipated operational occurrences, l
and the capability of the applicant 's proposed system to meet the con-centration limits of 10 CFR Part 20, Appendix B, Table II, Column 1, for the design conditiens and during periods of equipment downtime. We have reviewed the applicant's analysis of the expected releases of radioactive materials in gaseous effluents and have performed an inde-pendent calculation of these releases based on the methods and parameters given in Appendix B and C of WASU-1258. We have reviewed the Quality Group and Seiscic Design Classification of the proposed treatment
(
- - - - - - ~ . . _ _ . _ _ _ . . _ _- _-
1 l 11-18 I
systems and the provisions.to prevent and withstand hydrogen explosions.
] We have reviewed the capability of the system to prevent uncontrolled i
releases of radioactive materials to the environment in gaseous effluents. l l
The basis for acceptance in our review of releases of radioactive materials has been the capability of the systems provided to meet the !
design objectives of Appendix I to 10 CFR Part 50 during normal opera-tion, including anticipated operational occurrences, and the concentration lir.its of 10 CFR Part 20 under design conditions and during periods of a equipmen.t downtirae. Our acceptance cr%ccria for Quality Group and Seismic Design classifications are set forth in Appendix A.
Our .
!' accepLancu etiLetia fu: iraLicaentption pluvided tu coultul teleades of radioactive materials is based on the requirements of General Design Criterion 60. I We find that the systems provided by the applicant have the capabil-ity to limit relences of I-131 to less than 1 Ci/yr/rcactor during normal operation and anticipated operational occurrences. We find that the applicant has provided adequate control of releases of radioactive materials in effluents from the condenser offgas systen in conformance with the requirements of General Design Criterion 60. (
l We find the following items do not conform to our acceptance criteria and must be resolved prior to issuance of the PDA.
(
6
s 11-19 t
(1) The coupone.nte of the gaseous waste treatment system are not -
designed to " Quality Group D (Augmented)" criteria and the charcoal delay tanks, interconnecting piping and valves that isolate these components from the rest of the, system are not designed to seismic Category I and housed in a seismic Category I structure, in conformance uith Appendix A of this report.
(2) Adequate iodine removal systems are not provided for the purge and ventilation exhaust systems. Additional means of removing radioactive caterials from the gaseous effluents, other than diversion to the present SC'TS, need to be provided, a .
(3) Cont'inuous purging of the containment directly to the environment without treatnent is unacceptable, cnd a eleced containment with an internal iodine removal system or a continuous purge system with a charcoal filtration sys'emt should be provided.
Additional information is needed in the following arcas to verify the applicant 's design. This information is confirmatory and may be submitted during the TDA review. We will report cn this in a suppletent to our SER.
(1) The applicant's values for the dynamic adsorption coefficients (KD '# **" " #" ' ypton in charcoal delay systems at low temperatures (0"C) will be confirmed by large scale tests.
, (2) Following an in-plant measurement program, the applicant will identify for our review, tanks and components that require charcoal adsorbers on vent lines.
(
11-20 The following items arc. not included in the scope of GESSAR or are site dependent and will be reviewed for individual license appli-cations:
(1) The capability of the system to . meet the design dose objectives of Appendix I to 10 CFR Part 50 for noble gaces and iodine; (2) The capability of the system to limit concentrations and radioac-
- tive materials in gaseous effluents to those given in Table II, Colunn 1 of Appendix B to 10 CFR Part 20 at points at or beyond the site boundary; (3) The treatment systems, monitoring pnd control systems, and release
, points for the turbine building ventilation system.
j n . <: so; H i&s te un a,ement su te The solid vaste management syctem consists of subsystems for handling, storing, solidifying, drumming and shipping wet solid wastas, and for compacting and packaging dry, compressible wastes gcnerated as a by-product of reactor operation. Wet solid wastcs consist of spent filter-decineralizer sludges, . spent decineralizer resin beads, diatomaceous earth filter media, and evaporator bottoms. Filter-demineralized sludges and spent domineralizer resin beads will be collected, dewatered and transferred to a 170 ft shipping container, j l
Bottoms from the waste evaporator and the detergent evaporator will be collected and pumped to the shipping containers where a measured l
amount of cement will be added and mixed by a disposable mixing blade, j l
)
(
- . . ~ . . - _ . - _
)
1
)
L i
4 j 11-21 1
i f The container vill be closed and placed in a storage area. No pro-1 visions have been made to ensure the absence of free liquids in the 3
3 scaled containers. Storage will be provided for twelve 170-ft '
i l containers. ;
l l Dry, compressible westes consisting of spent air filters, rags, clothing, paper, small contaminated tools and solid laboratory waste l will be compacted into 55-gallon drums, capped, and stored prior to shipment offsite.
Prior to shipment, drums and containers will be smeared to detect
, surface contamination and cleaned if reguired. The drummed wastes will j be shipped to a licensed burial' site in conformance with DOT and AEC t
^l j regnictfe w.
4s .
j GE estimates that for a plant employing deep bed condensate demin-eralizer resius, approximately 140 containers will be generated l annually containing about 1700 curies.,
}
Based cn experience at operatins EUR's, we estimate that 110 con-t taincra per year containing approximately 2400 curies will be generated at plants using either Powdex or deep-bed condensate demineralization systems, with essentially the same isotopic content as calculated by g the applicent. We estimate that 450 drums of compressed dry wastes will be shippad from the site annually, containing a total of less than 5 curies of radioactive materials.
I I
i i
11-22 1
j Our estimate of the isotopic content of the drummed solid waste containa several short-lived isotopes, Mo-99, I-131, and Ba-140, not i
,! normally expected in solid waste shipments. Relatively short-lived j i
isotopcc will be present in the wastes at the time. of shipping because
, the applicant has not provided adequate storage space for vaste packages.
The applicant's design provides storage for only 12 solid waste con-i taincrc. This is sufficient capacity to store only one month's pro-t duction of drums under average operating conditions, and the available space will be filled more rapidly during periods of condenser cooling I
water inleakage. It is our position that the amount of storage space a ,.
provided is inadequate and that additional storage capacity is required.
- k'c feel that sterace for 13' days w'll resul.t. .in decey of mat of t ho
'. , short-lived isotepas. Also, provisions should be t'ade to verify the
- absence of free water in drunrad solid waste (i.e. , by electrical conductivity measurements or by ultrasonic methods).sothattherewould not be a potential for leakage from any of the drums or containers in shipment or storacc.
k'e find the seismic classification of the components and the struc-i j ture acceptabic; however, the components should be designed to " Quality 1
! Group D (Augmented)" classification, in conformance with Appendix A.
i
{ The system has sufficient ccpacity and redundancy to perform its
! intended function during periods of normal operation, including anti-I j cipated operational occurrences.
3 I
4 l,
- '***~-*--++=.,--..w.,%..,,,m_ _ , . . ,_,; _ __ _ _ _, . . , .
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __. _ _ _ _ h
11-23 Except for the lack of adequate storage space for drummed i containers, the seismic and quality classifications of the components, and the lack of provisions to verify free water, we find the applicant's l j design acceptable, 11.5 Process and Effluent Radiological Monitoring Systems 11.5.1 System Description The process and offluent radiological monitoring systems will be designed to provide information to operations personnel on radiation levcis in plant process streams, to initiate operation of emergency sycteca, to provide inputs to the reactor protection system, and to ,
a record the rate of relcase of radioactive materials in plant effluents.
The applicent has provided radiation monitors to monitor and control j the rcicesca of rcdicactive matericia in gascous cifluents frem the I
offgas system vent, the containment, dryuell and shield building annulus purge exhausts, and the ventilation exhausts from the radwaste, auxiliary and fuel buildings.
The applicant provides the capability to obtain liquid saeples from the effluents from the ,cxcess water storage tank and the waste deminera-lizer. Provisions have also been made te obtain gaseous samples upstreau and downstream of the offgas system, upstream of the steam jet air ejector and f rom points within the offgas system. The appli-cant has provided an area monitor in the charcoal bed vault to detect I leakage from the charcoal delay beds.
(
---.. ~._ _ _ . _ . _.m. ... .. . . . _ , _ . . _ . . _ . . . - _ _ . . . . , ....._ _ , . _ _ . _ . _ , _
- 11-24
'g 11.5.2 System r.vu3u., tion 4
3 The provisions for process and effluent radiological monitoring include the instrumentation and controls for monitoring and controlling the relcaces of radioactive materials in plant effluents and monitoring i
l the icvel of radioactivity in process streams. The scope of our review included the provisions for monitoring and control 11ng'the release of radioactive meterials in plant effluents in accordance uith General Design Criteria 60 and 64 and Regulatory Guide 1.21, and for monitoring radioactivity levels uithin the plcnt in process streams in accordance
! with General Design Criterion 13.
[ a .
i The basis for acceptance in our revicu has been conformance of the eppUr.4 ,( N dm lg. , 4.~ nn n ito.!< , m1 eint.a ha m fn v tho pro m m l
I 4
and effluent monitoring systems to the Commission's Regulations as set j' i l fort.h in the Gancral Lcsign Criteria and to applicable Regulatory t
t Guides, es referenced above, as wel] as staff technical positions and industry standards.
i Ue fit.d the radioactive effluent control syctets actuated by the i
building ventilation monitoring systems inadequate (as discussed in Section 11.3). The radiation monitors for the turbine building ventila-tion enhaust are outside the scope of GESSAR, and will be reviewed for i
individual liccace applications.
' 8 The typc of instrument, range, set point, sensitivity, calibration frequency, and provisions for maintenance and testing are outside the
(
s
~ ~ , - - - - . . . . _ . . . . . . . _ . . , .
i a
.l 11-25 l<
t l
scope of GESSAR and vill be reviewed for individual license applications at the FSAR stacc.
i
) !
0 P
4 O
M * ,.
I F
?
l I
i 4
b 9
h l,
e O
i l
t i
s e
f.
a 1
i k
!(
.I
}
e
- M sur e ,.,
- **'N*'M ersg m,ggs, ,, , , gp
- # -Ide' 9 #4-g+ , 4 , ,. _
r 1
12-1 I
l1 12.0 RADIATION pKOTF.CT10N The staff has evaluated the proposed radiation protection program presented in CESSAR. The review was conducted to determine that the .
! progran satisfics the following objectives: (1) to ensure that radia-tion exposurco to operating personnel and to the general public will
~
I meet the requirements of 10 CFR Parts 20 and 50, and (2) to assure that occupational radiation c):posures (ORE) to operating and construction personnel durJn3 nornal operation and anticipated operational occurrences (including refueling, purging, fuel handling and storage, radioactive matcrial handl$ng, processing, use, storage and disposal, maintenance, a .
l rout $no operational surveillance, and inservice inspection and calibra-t.wn) 111 he an 1:/.. x: pr acticable ( A1.AP) .
12.1 Shiciding .
The shiciding for GESSAR is designed primarily to protect operating personnel and the general public from radiction cuanating from the reactor, power conversien, process and auxiliary systems while maintain-ing suitablu access for operation and maintenance. To meet the design I i
objectives, the facility design classifies all areas of the plant into i radiation zones based on the access requirements of the area. Zone i
dose values are based on operating experience from the large BWR plants that have been in operation for many years. Both operating conditions and (in some cases) shutdown conditions have been considered in designing the radiation shielding to meet the zone dose rate criteria
(
..._.-_..,.y._.,...,._,..
L _ - - _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _
~
e 12-2 in specific creas. Our evaluation of the radiation dose rates in the various areas of the plant where shielding separates the sources from norraally occupied arcas enables us to conclude that shielding is appropriately utilized and will be conservatively designed. It was our concern in the early review tiiat icyout and other design features for radioactive fluid processing, transporting and storage equipment did not indicate that adequate consideration had been given to assuring .
that ORE will be ALAP.
Equipment design and layout recommendations of AEC's Reguletory Guide 8.8 "Information Relevant to Maintaining
, Occupational Radiation Exposures As Low As Practicable (Nuclear Reactors)"
, a :
were foraulated primarily with iogard to areas within shiciding where t,v p, a : ~r , . .. w 3he.es that the njer ORE cccur. In recpenne to our
~
concerns relative to Reg. Guid6 8.8, CE has supplied sufficient additional
, informatica to assure the staff that such equipment design and layout
- r.casures vill be implemented for the proposed plants although Reg. Guide 8.8 is not specifically refarred to. Although there is not an indica-tion that che design has been under cor tinuous review by a competent health physicist, as recommended by Regulatory Guide 8.8, the extensive experience with INR plant operation acquired by the applicant through interaction with utility EWR users provides an adequate substitute in this case.
General Elcetric refers in a number of instances to the use of proper personnel practices and procedures to assure that ORE will be
_____._____---_____.___________-_a
l l
l.
12-3 i
I ALAP; the utility purchasing the standard plant will ultimately be responsible for assuring that ORE will be ALAP.
While some of the detailed design practices that the applicant says will be followed are related to maintenance, GE has provided examples i
of specific plant dorign features for minimizing personnel radiation exposure during maintenance. Even though detailed plant arrangements and equipment designs are not available at this stage of the design, CE has censidered further design activit'les for lowering dose during maintenance. These include selecting lou maintenance, highly reliable a cquip'c :.nt , decigning for fast access cnd egresc to traintenance arcas, choosing quich disconnect and replacement parts and equipment, pro-I i
viulw.c for Legaary t,hicidi:rg, adegante wrLing cpaca and provicions for re:aote viewing, and other such measures. In addition, the utility 6
applicant proposing to use the E'lR/6 system and plant will have to specify in detail the restrictions and controls that vill be impleacnted to assure that ORE vill bc ALAP.
The area radiation monitoring system has the objective of indicating and recording abnormal gamma radiation levels in areas where radioactiv-ity is present (or may be inadvertently introduced) and 'to monitor the radiation 1cvels in areas where personnel may need to be (not just work) for whatever reason. For example, in access or traffic corridors where radiation icvels might unexpectedly change significantly there should be area monitoring. The monitor stations of Table 12.1.3 of
(
m,____.---__-__-- - - - - - - - - - - - - - - - - -- - - - - -
l l
l l
i l
12-4 l
ss .
GESSAR cover ir.est nuch areas. h'c will review c.he specific area radia-i tion monitoring systera details for specific uti'.ity applications.
'l l The feature of having the area radiation monitors be able to -
i-receive power from an Engineered Safety Feature power supply is important l' for many of the crea radiation raonitors and its incorporation is t commendabic. It is not clear how this feature is implemented, however, I
I :
and whether it includen all area monitoring systems.
Oper tJw prxedurer, are bricfJy discussed in section 12.1.5. It is I
I not clect wh. t the applictnc intcuds to do with information on "high i
l a
levels in creca not previously ccusidered", or on " doses received by ;
l plant perscnnel." Operating procedures will be the responsibility of the utilit, a +11cc;.t. a/d if it is intcuded tht.t. tLe Oe.w al Electriu 4 i Ccrpany will supply advica to the utility on the bacis of the noted in-l.
- formatien, it should be cc indiented, and the manner of use specified.
The an .lycis that c'etermiacc whether the pl.mt design assures that OBE vill be /1AP is the estimate of exposure, at.d ic coveled in GESSMt o
Scction 12.1.0. GE hap cctiented ORE on the basis of e:cperience gained in operatin;., P2.7 plcnts , k" nile thln experience should place an upper limit on c::pected ORE, the many improved design . features incorporated since the earlier design should result in ORE that are lower than those experienced in the older operating plants. It is the staff's position that a quantitative entinate of the expected doses for the new design should be ende by January 1975, in order that the utility purchaser of
(
l l
12-5 1
't l I
a DWR/6 plant may have an idea of what he can expect in overall ORE.
-) It should be possibic to anticipate and project ' maintenance doses I
for many activities, since a great body of data already exists at ]
l operating BWR's. These doses should be presented in terms of
? l accumulated man rem for typical maintenance activities in the proposed 1 plant configuration. We will report on this in a supplement to our SER. 1 1
12.2 Ventilation The propoced ventilation system has the objective of providing effective protection for operating personnel against possible airborne radioactive contamination. Appropriate design features have been incor-porated to make certain that the objectives are met and that airborne j
~
radivw Lisity levels fut auttal upetulluu, inej mlir.g imi inipared opera-l ticnal occurrences, are eithin the limits of- 10 CFR Part 20, Appendix i
E, Tabic I for areas eithin plant structures and on the plant site where construction vorhers and visitors are permitted. The staff is currently re-evaluating the inholation dose that could be obtained by a worker within the containment in the event of a pressure relief valve venting during occupancy, The airborne radioactivity monitoring system consists of monitors in various building exhaust control systems. It is stated that this system will provide a_ clear indication to operations personnel when 4
abnormal amounts of radioactivity exist in the exhaust from the 4
buildings involved. Sensitivity of these monitoring systems, and other
(
1 l
l ;
i l i
12-6 {
i.
system details are left to final design stage. It is the staff's i
position that the systems should be sufff ciently sensitive and extensive to be able to detect a level of airborna radioactivity in any room in the pertinent buf1 Jing at HFC level. The utility applicant will have to i revicu the need for additional instrumentation, since it has been our f
! l experience that t.here are several areas within the various main
{
1
, buildingo where it is desirable to monitor the air continuously. An !
)
j excr?le is the spent fuci pool area. Tha utility applicant will have i
j t
,1 to sr.tisfy the staf f that cufficient fixed airborne radioactivity non-1 ditoring will be provided to meet the requirements of 10 CFR Part 20, !
20.103 and 10.201.
n ',
12.3 Her1Lh Phvvies i
I This scetica vill not he addresned, cince the utility applicant f supplies the entire health phycies program.
l l
i 8
, s 3
(
f
-- ~ . . ...- ,... -- . . . . . . - . . ~ . ..m,-... .. , ,
~
. . . ,c...~.-n .. . . . .
15-1 15.0 f.:CCIDE!!T ANALYSIS
, Two bacic r,roups of events pertinent to safety are separately evaluated in this section: abnornal operational transients, and accidents. In order for the anclysis of events in either group to be acecptabic, it ic required that an accurate model of the reactor core be used, and that all appropriate systers whose operation (or postulated misoperation) uould effect the event be included. For transients, the analysis results must shou no fuel damage and no reactor coolant pressure boundary danage. For accidents, which are f ar less probable, i
analysis resulto arc alloucd to show fuel damage, but no other reactor I coolant pressure bourJary datcge as a result of the asstm.ed accident
, in alloued, other than that caused ini,tially by the accident. I The acceptability criteria of analysis results for transients are j that no ! w.! !. a r r 1< , (cind) dct :.e occurs (e coff].cient, but not n e c e s an.ry ;
condition to meet thin requirement in that UCH.7R remain above 1.0) Ond i
that penh nuclent yearel pressure not execed 110% of the design pressure (ASME Codes,Section III, C1cas I are act if nuclear system pressure remains be:nv 1375 psi;;, which is 110% of the 1250 psig dcaign pressure).
These two requircnents dernnstrate, respectively, that the first radioactive material barrier (the chd) and the second barrier (tha pressure vessel) are protected for abnormal operational transients.
For desir,n basis accident analyses, which evaluate situations that require functioning of the engineered safety features (including containment), it is necessary only to demonstrate that the second ,
barrier (the pressure vessel) is protected. This is done by incuring that peak fuel enthalpy remains below 200 cal / gram. This limit
(
1 1
{
- k. _. _ _ _ _ _ . _ _____..________________m
- e 1
15-2 1
1 conservatively demonstrates the absence of any destructive pressure I
pulses due to fuel vaporization, thereby demonstrating that coolable I
geometry is maintained, which requires that cladding remain essentially. {
i intact (even though some perforations are alloued). GE will meet thislimit.-f For postulated accidents for which fuel damage is calculated, the i
extent of damage is determined by correla ting fuel energy content, .
cladding ten:perature, fuel rod internal pressure, and cladding mechanical q
character 1ctics. These correlations are substantiated by fuel rod failure tects and are presented in subsection 3.5 and Secticn 6 of (
CESSAR. l l
15.1 Abnormal Oneg tdenal Transiente Ab:.orar.1 operational transients dre the result of single ecuip-ment failure or single operator errors that can re::sonably be expected i
during any mode of operation (~ The applica'nt has provided analyses of i various cbaormal operational transients in CESS!.R. These analyses l
include such eveats as process system centrol malfunctions, inadvertent
]
control rod withdraval, turbine trip, loss of electrical lead, and vorfations in oport. ting parameters.
Eight nuclear system parameter variations are listed as potential initiating causcs of threats to the fuel and reactor coolant pressure
(
boundary. These parameter variations in the analyzed transients are ;
l as follous:
1 i
a), Nuclear Svntem Pressure Increase. Transients analyzed in this j 7 )
e' group included loss of load events such as generator trip, turbine trip, loss of condenser vacuum, closure of one or all of the main steam line isolation valves, and malfunction of the reactor primary
(
system pressure regulator.
~ - . . . . . . , . _ _ _ . . . . . _ . . . . . _ . . - _ , . . . _ . .
l 15-3 b) L actor Watcr Tc ger_a_ture Decrensa. Thesc' transients included events that might cause a pouer surge by reduction of the reactor ,
l primary coolant water temperature. They included malfunction of the I foeduater control in a direction to increase feedwater flow, loss of a feedvarer heater, shutdown cooling malfunction, and inadvertent activation of an auxiliary cold unter system.
c) Renetivity ins er t i er t. . These transients included rod withdrawal transients fran zero reactor power, hot critical condition, and from full pcwer, fuel assembly insertion errors and control red renoval errors during refueling.
d) Resetor Wo er Invertore Decrease. Thecc transients included events leading te a decreas e in the inven' tory of reactor primary coolant such as Joer of auxiliary povar, loss of feedwater, pressure regulator failure in a directaen te ca,';c decreating reactnr systen pressure,
, inadvert<.nt openin3 of a taft .y/rellef valve, nr.d openinb c.f condanser
' bypass valves.
c) Primary Con 3nnt T10w Decr"ase. These transicats included failure of one or more recirculatica pumps or malfunction of the recircula-tion flor control'in a direction to cause decreasing flou.
f) Reactor Coolant Flou Increase. Those transients included events that might increase the rceirculation flow and thus induce a positive reactivity increment. They included a malfunction of the recirculallon flon controlicr in a manner to cause increasing prir.ary coolant flow and the start-up of a recirculation pump that had been on standby.
g) Core Coolant Temperature Increase. The transient analyzed in this I
category was lose of shutdeun cooling.
_ _ _ - _ _ _ _ _ _ _ _ _ ___ ___- w
15-4 i
h) D:cens Coolant _1nventory. The tranelent analyzed in this group ~
was feedwater centroller failure to maximum demand.
We have reviewed the applicant's transient analysis methods and find that we have unresolved concerns in two areas.
First, the transient analyscs do not use the scram reactivity curve which GE currently considers to be the boundary in defining the worst case curve cnpceted in the life of the plant based on analyses of core parameters at maxiraum core average exposures. In response to Question 15.51 concerning effects of the worst case scram curve on the turbine trip and neueretor tiip transients, CE states that these two transients "have been reanalyzed using this curve, and the analyses have been provided. Other tranaient analyses affected by the new scram a
l ren.ctivity curve are in the process of being redonc c'nd ware recently sub-
. f
!. 1 mi'eccd. *.c vill leitt c che rasults c ouc review or tnese analyses in a i
supplemant t.o the SGR. A large fraction of the trans!:nta analyzed in Chaptuc 15 invoJve reacLcr scram, and GE is revising those analyses to include the latca. "uorat case" scram reactivity curve. In additien, the primary scram signals used for reactor trip during the re-analyzed turbine and generator' trip transients (stor valve clocure and control valve fast closure) do not meet applicabic criteria (see Section 7.2 of this SER).
The second staff concern with the app]icant's transient analyses is lach of inclusion of offects due to new systems and planned operational modes. New systems include proposed addition of a Prompt Relief Trip syster. to open relief valves (before the pressure setpoint is reached) during turbine and generator trip transients.
I Approval of the overal] concept of using this type of system on CESSAR
15-5 pInutc is concidered an open item by the Regulatory staff since GE has net edequately described why PRT is needed nor have they satisfactorily.
discussed alternates to PRT. In addition, the PRT system as currently designed is not acceptabic. However, if such a system is eventually redesigned and approved for une on GESSAR plants, the appropriate channes to Chcpter 15 ana)yses vill be required. Other changes whose ef fectc r:ust be considered include a new solid state 2-out-of-4 logic protection system replacing the 1-out-of-2 twice logic used in current plantc, and changes which involvc use of ganged control rods and include systen changes to make thic possible, including deletion of the Rod Ucrth Miniminer (REI) and Rod Bloch Monitor (RDM) systems and addition a of neu , interlocks (a revised rod pattern control system). Also included is a revised control rod position detection and indicating s
1 of w e... .
I The Regularcry et.aff believes that the Chapter 15 trnraient anclyscs vill be more representative of the planned GESSAR type i
l rcacLar ence the prenited July 1974 reenalysen arc incorporated. GE her: iccently subnitted the nev cerc:n recctivity curves at: well as GLT/.D t.nd /.ppendin K *analyncs. They plan to submit the effects of the
'I new des.igns by January 1975. Because of the above anticipstod changes and the present uncertainty as to which of'the analyzed transients GE j assumes will be affected, a supplenent will be issued to the SER evaluating all Chapter 15 transients after the revisions 4
' )
I have been revicued and accepted by the Regulatory staff. l l
15.2 Desien Bacis /,ccidents I The applicant has evaluated a broad spectrura of accidents that
( might result from postulated failures of equipment, or their
~ - . . . . - _ _ . . _ , . . - . _ , _ _ _ _ . _ _ - . . . . . .
15-6 maloperation. Four highly unlikely accidents (design basis accidents)
I that are ri.present tive of the spectrum of types and physical locations of postulated causcs and that involve tha various enginected safety feature syrtems_have Lcen analy;cd in detail. The calculated conse-quences of these design basis accidents (DEAs) cxecod those of all other accidcuts considered and are the same as those analyzed for previously licensed BWR plants. The DEAs analyzed were: (1) control rod drop, (2) refueling, (3) steam-line break, and (4) loss-of-coolant accideat (LOCA).
The Cantrol Rod Drop accident results in the rapid removal of a high reactivity vorth rod from a localized region of the core. The a .
DLA f orthf e event is chosen as the highest worth rod that can be deve'oped et any tine in core life under any opernting conditions O.
( being dropped from fully inserted to fully withdraun position. This
! i resulto fr the max; 1 . credibic locci reactivity addition to the core.
Refueling accidenta considered are chocen to represent the vorst et edible ci entr. possibic with the pressure vessel open and therefeta not previdirg a barrier to eccare of radioactive material.
Accideate censidered in thic category are: 1) Cash drcp accident in uhich a spent fuel containing cask is dropped back into the cash storage pcol (it is physically impossible for the cash to be suspended over the spent fuel storage pool, so that event is not considered),
l
- 2) Spent fuel cask accident in which the cash is dropped from a trans-port vehicle to the ground (representing the worst credible event outside the plant) and 3) Fuel handling accident, in which a raised fuel cic= cat ic dropped onto ctored fuel bundice (representing the a
worst credible accident to spent fuel within the plant).
L - _ _ _ _ _ . _ _ _ . - - - - - - - - - - - - - - - - - - - - - - - - - - - _ - -
15-7 The steam line break accident is chosen to represent the maximum credible break outside containment, resulting in maximum release of i
primary coolant outside contaimncnt.
The loss-of-coolant accident (LOCA) is chosen as the instantaneous
. 1 brc.:h cf a recircu).ation flow pipe with unimpeded discharge from both severed ends. This accident reprocents the most rapid credible coolant lose and depressurization to the core and therefore is the most sevure challenge to the Dacrgency Core Cooling System's ability to raaintain core cooling.
Analysee of !.nticipated Transients Uithout Scram (ATUS) must be .
added to the GESSAR document by amendment before final Regu'latory staffa'p'provalofanyPSARreferencindGESSAR.
The Rep.ulatory staff believes that the same concerns stated i
above for. transients also apply to some accidents considered in Chap:er 15. Concequeat:ly the accidents analyses uill be evaluated after revision by CE to include effects of the " worst case" scram curve, and effects of any new syrtens upon the accidents. Results of the eenlutica vill be includcd in an SE" supplement.
15.3 Loss-of-Coolant Accident (Radiological Considerations)
A design basis loss of coolant accident has been postulated for the GESSAR design. The fission product removal and control systems
, providc3 by CE are not adequately described to allow sufficient credit to reduce potential dose consequences at typical site boundaries to within Regulatory Guide 1.3 guidelines. The primary containment leakage which hypcanca the chield building annulus has i
not been adequately defined nor has sufficient detail on the shield
~
15-0 building annu]us recirculation and exhauct cystem design been provided.
I The staff has performed a LOC /. doce analysin to determine the exclusion ,
i boundary value for the relative concentration which would limit the dose consequences to Regulatory Guide 1.3 guidelines. This analysis was performed for various bypcss leahage fractions assuming the balance of the primary containment Icshage is filtered by the standby gas i trea t:ac nt systca. No credit for mixing in any secondary containment volume uas assumed. Tabic 1 lists the assumptions made, and Figure 1 presente the short tern relative concentration required to liuit the dose conce,uences to P,cculatory Guide 1.3 guidelines as a function of direct bypass leahcce frccticn. As indicated in Figure 1, even with the aseumption that all the primary containment leakage is filtered (0% bypasc), the limiting rol tive concentration is lower than any i
- vcluc previously liccured and corresponds to an exclusion bounda'ry i of greater t .an 1700 netcrs for aqt acrolocical conditions of stability category "F" and udn! rpccd of 1 meter /second (Regulctory Guide 1.3).
(See Saction 2.3 for a further discussion of matcorology.)
Although a filter officiency of 99% was used in the dose computations, there are scrae outstanding items relating to the filter decign which arc still under revicu and vill be covered in the Safety Evaluation supplemant (See Section 6.2.3).
i l
(
l l
1 i
)
15-0 a
I r, , ', Pu j " ! aJ '. ..a . /, e c id- a t In this acedjunt, it is eseu'ad that e fuel cnn.mbly in dropled duf.'.ng ref uelln;(
opernt.icas, and th:it an a result of t.he fal] ,
- 97. fuol roda vre d.).:. -d. Our ca. reptiens f.or this accident c ce ct.1.:
43 atent u.i rh tha conservative asstaptica0 of Pegulatory ,
C.:!.h 1.25, Activity released to the environa in assun J to be rr.leau:.. throup.h the SGIS deep -Lcd, cber.ccal fil:ers uJd.:n a 2-hndr peri.d.
1
- 1 Un:'.n3 an cs'.. u] ealee fer X/Q cf 1.0 ;: 10-3 for calculational pur pr..er ,
th : . recultlne, doan t.ould ba about 2 rem th:~oid
. .r.:. .. . , n - ve.m y :4 , i; . : ;.. T.n ;c, t,a._ u;; e , .. .. .::2 ci. tuu
, len-of-coa 3 cn t accid o.it are um n liriting.
e 3 .r . 0 .C.o. :.r.c._re.1 :1 D rn.e The Iostulci.<.d coanol r.rl drop cec dent assim.e; that a bottora .
entrj c utrol red han .Lau: full; 1:'serted and becoms ocuch in this pc.sition, unbr. n to ti.e racerer opcrator. The drive
$:, then assun ed to becer.2 uneoup3 ad and fullj- uithdraun.
The rod subsequently falla from the core, in.certing an amount of reaer:.vi.t/ corresponding to its reactivity worth.
In evaluati.g, the rdiolo;tical consequences of thic accident, we made
.t. :ur;ptions that are based u;>on the applicant's analytical inndel as presented in the PSAR.
(
IS
- 0 '
~(
In thc. aan1/ i;;, t.ia: rod la ecuaaed to drop out of the core- !
I durin;*, c.t cr:op i ; teh occutu ')D :;lnut es t.!.ter chut.Jovn from I
full p:nec op.cn Les. Thia ic scnuend to cause 770 fuel reds Lo p; D ' a t a, rele .Nia'; .100., of 'their ' co7 tained nobla. ;/ mas l 1
cnd Sm' of their conenined hnlor,en,, to the. rcactor coelant t
cy s t e. .. Th. parforated rods are ensurued to have operated at !
- m. l n d 5'.";, creater than the a/ar'sa red (1.5 pcching facto *).
.0T tl.e Mo p s m _aared, it. re ceu::ed to be retvined in the p -~rv nynen ar.d hali of the re- cf. Mar J r removed by plat: aut. rs, all.o r Lbc neb 3 a gr aec and 2.5."; of the !
1;: 10:<. -
t*'n a#Tc'atOJ rodr r:0 c.c:'.n d to be available In I
i-for ;;el:;rse. T . tac tlea of a hig' radiation si:;ntl in the noin t-nt an li..ee mitsmel.::nily cle:;ar .tl.e ta:'.n c' eam line J aci a tion v a l n s , + ' m t.c bre.1 ti.e mechanical vacuu i ptc.p and closes the
$ v1,t.h n 'ecivr da :nntrann of t he pu .,p. Tiu activit" entrained in t h; c; cdens. ' in apumed to t e released am ground la"ol from tha tu:blu bu11Jin;; by leaku_;e ficm the cendanner at the rate
, of 0.5% o ? t'ic condenser vo]uma per day for c duration of one day.
TL cchulate. t.co-hour d ms , assuming, u ;/,Q of 1.0 , !P-3 gcr '
calcu ht hnal Tucpor.es, are 22 rem thyroid and about 1 rem whole body. 'I; ' dass for the course of the accident are calculated to ba 4. bout 13 ren to the thyroid and 41.0 rem whole body, nsnnntn; n y/q nr 1.0 x 10 int enicolationt1 purposes.
(- l
e y
15-J1 TAsiE 1 '
-A S W.? T I O_'... _F_L. _R___G_F._E S A._P
.L..OC.A is iSi A" u'i m "
J e . ., . . y n. ,., u. ..
- 3 f ,e, ,,. I.,. . . .,. ,
c.,, 3,.h. va, ,
L c . . . . . .,
a .ve:xn Iw ' Cr;u! !.n~ ,L T.cu.. T.;L. LTjany ,
a 1IM : 1:Tri .:. ;cy (cl1 icdi e fer;c)
- .. 9;;,
j e, ., . .
e' &
y a 8" a e e 8O W e.d k g' e e g 'e J %
'g...- .1 3- s . .
s ss e e
f n
i
i 15-12 .FICURE 1
~
~
'Ah: ~**
'u L I i^ ; ~s* TNo f lo u l;. Cy. C:. . ; I o u I
S; y II D A P. ,'i D o C (NI, "To /( 66 M L frTo y
( l GuIb2 (, 5 G U ID E LINEC g .
} . _. .
l t
.\, . . . . .. -
\
\
\
. en
\
i
.. . . . ... . . . . . . W dut C Co l. y. D e s e ...
vs . L 1 N. tT G t> To 2.:) l!'m
__ _ ~ ._. ~ ~ **-- . ,,, _ j c --- ~
y, gc s
. . . ~'-.-~.--- ..
q, ..
g . . . . - . .. . . . . . . . ._ ,
. . . . t .... _ ..... . ... . _ . . . _ . _ . . . . . . _ . .
. s \
r s . . . . . ._. . . _ . . _ . . . . . _ . . . . . ...
n
\
i, - .. . _ . , .
....I.._ _ . . . . . . .
c.
.. a .
d /
. '\ . . . . . .. .. .
c . . .. .
... ..\ .
i .,.
r ,.
s,.
N , . .. . . -
\
(,,' . , ,
c tn K -
.7'n ; t O t 3 ;> cs : C . l. t t.ilic
- ra tso R ti1
(.r ::
y . . . . .
' \ %..
. - e, i s N
- n - .
N a5 '
- a J'
- %
- Ur . .
N.
o
}-
s .
., .. . ... ..N. . . . .
y(% ==
E' .
l
.. I l
1 z /10 6 o IO 20 30 no So
< c) To SO 70 1:. 0
(
B1 r* C" L i t, c M E ) i i n C E H o f inic s.U T ch l
1 Ch 11 Yir:1:t 1:-r 1. E' <*r K. R f. T&
.am
15-13 3...... . . . ..a..
- uv. 's ,
'rir a 24. hours 7..,,... , .,i 7. e. , ts 1 *2 ,.0. J .,. .Lra W C; : /,0,116
- 1. ' .: 31 +- t . n;.- Ir e ~ v J
- ** c, o J u s.t ,. ,, ..ww... * - .
- ' 3,i n y.. .j. , .,... L .. . ;t
+"
- 1 *
'.t18 t / . T . 7. . . ' D w . . h. 4 * ' . A ? C .s -
.... p..,v ..
I,.,,4. ,
7.r> * ,
<e , . . . .....a.. s .3 a.
- v. -
. . . r-. .. . . . -
. ., 7 r ;.. ..,.. -
0s 0.4< *
' M .',
i
-o
.s .. _. . . . . v ,, . 1 - ; 3
_.u. t . . . . . ..
e .t D - 2 1 " :t 1 >: 10 ~
i I
L i
l '
l ,
+ 1
s 1.',- 14 I
- s, ,,,, ..s,. ..,,. .
_O .s .c.. . . . . . .......J...... ._ +. . . . . . ..
Ih.: a. :. of F: sd t' '. Inui.v.d 770 l' r .' c:1 c.. ' ir ;ie F *.,J p :. s ventery I L.>l ; - . c .'. . . o l..: c
,n , o ..ee(- m a, ,.v J, /,.
} U ; . d. r'. .' $ ') C ).
- 3. L . , . . . c.
. .. .. ... L. .
.. . .. .w.,
r . ,. .o t .. . . . . . , , ,
a,
- 7. s..
. A. . i... .. .. . ..
.1. .3 ciii, .:. r.
Cu: ; .a* 50.M a C r ,,. , .
y.~. ~ n - + .., . C3 . ,e, , , c._. .. ,,
yen . ,, ,.
, J,
- r. .
.d .
1 x 10 -3 C - 2 h; ra C - : ' . . . . - .- ..
- 3. , . 1,..,-
- e. *,s y o - _c , .n y, 7 g, b
0 1
I
d
.17-1
- ]7.a ..y ._d _i ;
- _ m. . .a_.c_.2
]).1 .C.a. ..ci_a.1
'/ht: ur r.crip tion of t ho Qunlity .'.ci.ureu c c (0'.) Pr ogr.m for the C p::r"1 1.le 't. ic . (G) In P./G J a coni.c3 nd in Ecction 17 of the GE
- .l . r J .rd !b f J r. /;.ta1.ysJ n Rc.; or t (Gr.Ml *.d . Our evaluation of the QA T. ;y,re n i f, 1- ' er a revi . of t his J ufc n!.h tic.r. NJ discuri,lons and w.e t;n;m 4;; th Ci. to c'ercer., t as bm ti cir Q!. l'rn:;ru:. co6plieu uft.h the r'* , w a o f /.',.erdi- E u 10 C. ' P r t 50 ,nui . the apn]icsMa . ,!
IU a . < . y nn ? .
3 */ . .' W . . . n j G: ' r Lud - '
Eccr:;,s . n/
- u L i . l c.:'.w (? d.).1.
recp6c.r il 1c for du d :. r i g . n ac i r.c'.t.p i 1;,, . !'.. ; s t!on c. ' rc e.L it.g cl i.oi ini, d u er P.c.. ' t or v
s s .s e
. . . . , ..:. 7., ..
.4 , . .. ; : ,. , , , .
Opa:cciv.i, (L ,) Ulvj .; Ibn 'T , l. ' tisic" h ~ e fcur n::j t - d e cer tL c.c t ."
i 1
1
(;. '
"izura 17.1-1 of C. %?- t) : '-runc t & Ovalit f /ccucance Oceretiene i (P /'.0;, Mue'- . Fue3 1!..rv u ut (:'.0),'L'J.: Sye.s..e Dcrert n et (K 't.SD) , j i
l l
nni 2 frcd.et- % c rt et (U.7PD). Each ni th nt deptrtu.-r.te in headed l l
Ly a Ccat. ral Mar. ,:r vbo ren:irte dir ect:y to the 1' 'fr Dq uty D.i eisieu l
C: r,0 cc i Ma r; g. r . The rejer f aaet.icns c# theue fou-. departments cre ru ;nrizcd ac.d de..cribed in rigurc.17.3-8 of GESSAR.
The. D<pu t-; Div3 sf on Cc.ncra' .an. ger ha:, ent 'b31thed c UI:RO Qun]1 ty Cruncil chcired by t,ha Mar.c.ncr of PLQ.t.0 anc concicting of thnaccrr, fron1 th e ::.ajcr org .nire tiors in the Division. Tha Qt. lity I 4
Council ic resp.nsible f or assuring tuts 1 c;ucif t ) unifet. alt.y and I I
censir t.cncy throughout the design and manufactiir!.nc of Li.?. cnd for ,
1 4
keeping the reputy Division General Manaccr abrens: of quality-related i .a t t cr s . 2 1
1 I
l 4
I I
l 1
1 17-2
.4 'b (; '. orpoitstf atu,1:ithin 13.*RO end th: Er reportinr, levels are as fo] 2 o . (::.'e FJgurc 2 7.2 -3 ) : a (7' 'i nc ::n.'arcr: of Quality Assurance and Product Assurance repo:" to Ll;c General ihnagcr of P0Q.'.0.
(~) .JD G .nm. ,cr reporr:s to the Gcuerr.1 Manascr of the ::rD.
(3) 'l h c- :. rs g:; cf Contra 2 und Inctnmientation ::anufacturin;;
(:. 7:I: ), Re:ietor Equip.yat ::nnufccturing (RE :), ar.d Design D::U n. ' 4 r 1, ecch uT ':ho ' han a Q' :::ngcr report.ing to hin.
n cn. , to th C c i : ..) . : .n.e.t.cv cf tb a Li,'.SD.
(4 ) lu: un:.f c' of T:ncin ccing 12quir:".pt Prccurencut luctcllatien
, (Ia '. :) 1.1. h;n n QA :amgcr reporti:.;; to hin reports to the G= n? : .m . ~ o E Y,t.'i:1).'
! y; ve, ,.m M $ 'eic- cf t.,r- i q;_ :1:n gcra are as follo,r:
I'
.M..' .' O. .: . . _..- ert-blishcr c.ar.litv-cclated 14.;G and Divicion i
~
F;' - l cr e ' lw ' ." u '.. . ; nudii c the ::LD f tmetion.21 organinetient and th.ir p i N :( . 2nd Jr. clices to ace ur e cer' ort;.nce rath DI.'R0 and L' i '. ; ; ' " l '.lJ c. ires nud ? t. ' : t.c .j or 9 ; verifics co.r.plianca of the overall Q'. Tronc > s i t h ci.p. 'h?.. codcL, strndards, rogu'.c.tions, and recp ctive cont ' . .: c oTc ' '. 1. .h: c'wdutl. selected ety-'ncordna design revimte inJccer ' u of th:; C , ', verif? cation and review groups,
.c. f :::. .r .mr o f '._; '.l - n.rr.vidcc quality assurance plannin~ oand QA Procrne is.:03 < m ntat r.n for equipment (fue.1 elements, channels, and 8 .
Iec1 el. "
cc:':ca .;t s) ranuf a turr.:d 1.y :70 in Uilmington, liorth C:ro.'in , r c *. ell an p:.u Laced :.it:. rial cnd equipacut used in the inanufc e t ut e of I:TD produc t s.
( tu lcgy,y of Lci:1 - provides quality assurance planning and Q/.
A 1 u, p. ..:. i a .plei..;m a t 2.ou i vi equ y ...n. . i t (::.cu ,u t 6, . ,
.Lu z,i t u. il pane.2u unu
17-3 i
rac' ;, g riph.<<1 e2cetr$.J! m:.o. d r;: a r. cont rol c .:vipt.wnt , and I
ccn.ir b t: 01 nr? equip- nt) r.und ,w ture) l'y C;1M in S.in Joce, C .1.ti m: .. .t ,d1 au pur e! actl m .tcr ial cud equipr.mnt e.<cd in the l':. ! . '.; ' $ . P ' pi (.' "< I .' a t ir D[in i .
.n.
r -
,x a
ai r. . p: ;.vf d,. ip. l.i t"J c sr.urc nco plcnnin'. rud O.A Prc..cic i: dc ar: tion for.equiprent (e :>r trol rod:', contrc1 rod drives,
- v. n .- un . c.m '. C R hn.;tu'ic control r'odu':; a) mnufac tured by PrM .
h, ' f r ; -- , en' Cr.; ;i Ji ., r + uc.11 as pozcL:.cr.d r terial anj
. . ,p ,
- 2. . t 't . .o - ' '.."...."c- ". , o'r'. .
.C. . .' . ,
,'n -
" *. . c .; : ; .~ - ce r ine. the desQn '> recess; a
.o.,,, .. .. r .i. s
. c. > F.- (J....-*.
..,1 . , ' < . ' . . . . . -, c m. .^u .1 .'.. 4. r. . c ~i . ." .".. . i .t. '. t r, ,
a 4 n; c o. r. .: .r . i . : :c y.. . '; u.re:.tive 6ccien.
0- . # r. - ' - . : 0, r< ;u;i c.t tr vv.iders of
?
. L, . s -
u
- 4. t - ...
- i. .
i .1. r , , , .
. \ . 4.. ,.1 * ) p .. .q .
,.. t, s . .s,.. . , .
i.........
..4 . ... . . . ..:s 3. . ...
. s ., a . .,t., .., . . .. .... . ,. .. t a t .< , i . .e . . . . n..
c.t 1 . , . . , . ..
, , . .. . t. . . < a- r.g. 3. .. *. - r- ., .-.-,e...'i..t., ,, .e- r .r ,.,
._.e .. .. - s .:..
- c.,
.,..*;,.. o.r r 1..
.. ,, ,., .. ., ., , . . .... . .tu .e. ,. ., zi J, ,
. . . . .. . 1.1 . . .:. v. .
,n J.,, . ,
yi . ..
1 3 .. .... t. , ,. , ,
- s. ,. ... . . .
.......a . . 2 ,s....
. ..-e ..
1
, ., s + 1.i.~i..-..u.,
., a e 1,. 5 s .
4,...
.. , , . , t. c. i .
, m. .
c .. .. n. .: . ... .. < n ....
s.s...o rO,; *.12; .i. - .~. TI:C ( '.' O r[i.
-, . t.;J ii ; l C L it.!; L U C C *
..U0 .f UU;. s !.OP31 rerpr .rit '.'s " nor igrcu.r t: are cuch th vcrificotje.n of co-foruance to cataL.' ' e, t / i t,t r: , ccle;.' .! : c .c ;. li; '. :J b"s cliose .m do not 4 1,. 2..\.,,s. i:.!... . . 4.<.4,...
3, 2.. s r
.t fu, . b ,.,s. . . . r . , ,
. ,. . O . .u. .
u.+. ., O &, c. .. .. ....s . 4.n 6 proJuctt. I.,
. .m 1o .e) ~. .
i t .. eccc : functica.. have di r t.: accccc t, t c. -l evc 1 . ' 1 it.cne.;; . cent . The PA Managers have the neth.rit , 's.. .: i .~ , . c r > .m.' n . :-. ; c.r. ~ 1 Frc .m t o ' .? r r. t i. .f v e.uai ' t .
1 s
l l
17-4 p ro'il e ,;. verif y ' i ; l e; .ro t a tioa s,i tht sol utions, and prevent further 4
procera19 dell er:', inntelle tien, or ut <11:atien of nonconfor:aing ' .
1 itens until ptop;r i..punitioninr, has occurced.
Im u . ri cue r ot i, ' ar d evalua tien ci the repm: tiny. Icvel, duties a'r. .v J J 11tica' o f pcr.wn;t re. ;murJ ble for QA fm.etions, ne find tk: it' is . e ca.i fit J c it d.:le: c t eJ av'.harit f to pi ct.luda undae ca.st- -
- m. s n :. .,x . ':" o n : ,,ic , c a t ., v a. t a..c., and hr.w c.u.4 :r. ,cien t
.c r. c u<< .
r term- r.:1 ' .: '. tc. pr * "rly < : :.c ' lir h 01.d iv. l rent an cf f ec tive QA-1 -[. s y: , ','. . T .. r ; , i ', je t1 - t f f 's cc :" ' o ; .aa tt,r ' ' t!.o QA or ~.;.n c. t ic.t a _
ntrursu c jtra m..-1 .; n G !. .. ',"' i', ace:-;,i M.;x
. r. ' tw t c tbe require :n t:
c: ,pt ac co la C. i :. .. .u.
d .,
-- - i.
. :. : . . : ':- ..,n
! - c. . pL '.q ., . . .(.. . . . =
Al] m: et t . . ' ' "' ' m ;. . . ve r s. 0 . . d c e' ec widered te j ,
t-i -
d e t er. ' t '!. e: :;. a , "
t ' ' ew - .a c.!cMr s'escribcl in Sae tion 17 i
l . c f C.53 8,' r;; ; o. tr eh uti.icr r.n? f c.rr' , t;.e tot 41 hod:: ef :).c i v.
)
l Q i l i c .,t. T' , c? p!.cc t penJ'c.- e!. t.hc QA Pre;iam Jn GTSSA; nrc 1
Ce.:cr.ilt in t ' c f o' r .d r4 p.r. .. u:;h 1.b:'ch irr06 llc br ais of cur ev al v : t i o r :..r. ) 1.;i .t : 1-l Le ep con:.lmir:n 11;c t it riect" tm i et.'
d
,1. e.: c n '.s 0: A i, cis ' Eto1D C'.' .% .
The Q Frogra.a of CE spp15 s to .11 T'a safety reloted structures, cyntc=c, c:.J cc: pencnts, includhm fuc], throucheut all phnses of desi;;n, manuf ac: . ring, inup r tica end Ler;.dn3 GE carr.ite to ec.7%ing teith the reythe;.cnts of /c.endia D tc 10 CIM Pcrt 50 a'.d uit:i those ANSI Stant'c ed , t.nd rou:;h draf t s.,tc n:k d.a a j ircr.a,u i in tne 'EC Crc; Loch "Cuidc.nce ort Quality Arsurance During Design end Procure.cr.t Phase of Naclent Po cr Pla..t.c" dated Junc 7,1973. C1 uill give !.EC tir.c.ly notifien! .i.en of any p':opwed sier:ificant changes in t he QA Pretran.
s 17 5 *
.hs o',, J q , 1: . .:<o 4 .;r.c ur:tirm, cuch ar QA Manuals and Preen J r 1, a n i c racteu! f rer the GE Cor;rcato PreJunt Quality Po lj .; .. r < r.c is.11! y :er ti.a fir: .1 revh and imuance of' the e: . ' . 1. 'r:
- F : . ,. c. n
- m 1.' ' b t he r.71 Deputy T.'>1sion Ocneral .
M...' ,.. CT -* ' . ' ' '" I' s ' I. 0 I : p.1' D I . Inl O.'. Cc:Nf.'.1 2! ace t er ' a o !' ) m , Phmi: n 1 : ; ' .c. i r . . ' rc ' .1. Q.: .'.: ;.y reli.'y end ~ Diviciot, 1; r . i n:. , r: t" Li a Ou: '
- L, l'alic! c:.,, :bn nis , 'st,d Proc cdnr.2r.
.m.. t. 4 . .. , .s . .. .. ,_.. ..
. ..,3...,. , .... . . , . . . o.
. ,,.,s...
.. .. t e.
i . .
u.. c r,. et . i. . u. .
( ..t' i co n - " ^E
' .. L'i ~;.~.! ].* .- D i l'U f J:J C .: :: O f f. i e d 3 0 t r f. [,U!i d te ;' -' .: .1 :. . c. ' ,
pe. .. t . d . T. avis ic. ' t o tin. :s d:: . uncn ts
. u. ,
.. i. .
. , . . . .,.s... . ., 1,:f,i, s.., . .. ....m. ,. .r .
. . ter ,t., . c....,.
4 .. .. .
.c i:-
e' c.h Oct: a . . u ;,
.' ' , ,3 rc .. .a m 4 f ,. r: .s..i
. er: :, !.y. .-. . e J.,
2,.5, .- p-
_- .J'
. ..n.... ,,.. . r. n~ ...f ..3. , .. .. .1 . m~. 4 .ea, ,, q u_..t .;.,, i e 9 . ; t 3.sr .t a.
l
- f. C t f.' *. L - . i (, r; . .t o II." C LI j r/ $ 1.C w r' '. t Cd in Lhi. *rCper intel [.rC- .
- t. - t r. . . , n: .1 :a l o.
- . : . (;t N 1:eles, n at : 1:, , P r c,c ad u:c c.;, and
'S'. r .v , !. n . ' . t . . '; . . ? t n. .' . p r n ar travid : . ' cr tre.inin". of
.i n t i . !c.. L e t L . c. , t' m; ;, t . t ' c .1 r ci .. _.: n er e: .
E: 0 P;c .
p re'. i l . fu au c-ec.; table desic:1. control cyr:en foc c:.ru".ovt- :.t. v.s :.nd ccr..;w..cn'. s s:Y.ch ic do_,=cnt ed and controlled 4
ley prc cet: trco. ar: f1 ,:Lruci; ion: Ti.e.ce pro:cdur w m.d inetr'.:c tions deceril ; the rt ..onsih it i c.- c c, J j nt. -faces of each orcard n t'onci unit '. .
i .r. - e m.,1c?:e:. 4.: . , re'r. rcdbs11ty. Tht.; also include t:Ca s t ' c :. 'C '. :. c. L " ' thLL; (3 ) The deed n rcruJreacm.s arc c'cCined and that desi;;n ciL L .1 V J t 1.'b Wil LJ6 G U E i i LN OUt i L1 0 } ),, L.111 '.. , CalltrO1J 4OG , 3DO or& rly .cnner.
l i
______-_ _ _- _ . __~
17-c.
p.) / Terr :: .; ya.;lJ ty re.;uit enant s and stact:ards are specificd
(
in de:Q::'1 d O C'.r r. L f, ,
CT SaitM T e ' t cr!.cilc , . im r a, empoornti, and processen are ;
- c. } } ' l ~. ,
(! ) t c ' u . i ra' i f.i -r' + i n u : L t.hr Js .': e p;operl: F. elected.
(.3 ) 'f . J s .i y , c. c c .e.-15i..~ for a:u:, . i c y b) indIvi..'v..1: or
- . u,. n;.t hr.vir; . ;.p ra';Miity for t! " or tcina] d r.a. i cn ,
(c) P. , c r';er ce ':u:t:oU .d to tha er' J c e c 3 c c c.1 c a: -
'te L'.c c , J J e < ;n inc.1 :: 1 . ,. r e, n and 3 c-roval o: :Le . , n, i ita th- per corne.': a c r2 cs .o 1 r ev v.:
.m' '
- a. 1 tr 1 . nc Lt- 3. : pe:,albl. d e'.p,n or;nni .tl.,
a -
1 s c u. . 01 L: G. v a;;- t. t .
n -
.i m. , . . , , , -, .. .r .y ,. , ,..t..
.:. . .L g. .1 .. (, ,.. : ,s.
i
. 8 n :n 24 ~17.c c in ti. . J; :1<;-
tn - t: -- cu ;:.:tivco: ' L: >
th J -
't z e a r '. t f ., f re- 1.:.o st er n.c J n;' der. v i:dl: funcMen as
,3-' D <.q. r d c>c m 1. r o . a f ter n . J to cl.c enclicc-t anJ/ r Lic
- r, p.v' fc.r ir.ecr' :-. corcy t Ibilicj r er.:c . and caordin it ica,
( n. . ' J c. L . .i t i u lis t.c o f d e aiu n .l oc u: ;t - c e c:tL ainaJ to a c .i r m up . t. j ti::aly dir: ribution te r: crorribl ..ndividu ]t. cad .
8u - ~ . ..:,,.
s........
..n.nres c.re rrt.ablirl:J and documented frr the prepcretion, r es l c. , a pp: ovc ' en3 mntrc2 ci procurecient doeurient.c to prcv.ie:e
. ccmr i L "e;' ] .tcry requircrent:, c' e ci n bns es , are: quality rei.iin Trt n: so.1cet or re!crcnced la tht. precorci..ent docu r.ents.
Qual .i t;. rt.cuire:.;;.nts are p'mpared , rc.vi enc.d , cpproved , and issued by 01: Q.; oc;,.n. : <, t is o f or ina urson in the pro;ur. : -nt acco:.
.t . Purc hen.c
(
....4..
3
. . . .. ..,......,--,,u.s...,s
. .. ... % ... . .,,. . a... . .,
4 , . . ,a. . ....<3... .j c,. . L..,
. . .,, r. .,g,4. _. . . 4. , .,
.g
1i'*, s n l '
- s. A' . pr ' *r l a 3,v , '
.s., }h>pietr of pro,;gretmpt docy;,;cnts 4.
17es ' i G .i ,
. . J.u,' and th ; a~ Pu: pro' tid. un ar nce that
. qual.. , . rt is. ze co:,; h ! . m J cer' celly arr.t u', v.9 t. hat they C. 11 b. (s .* lli .oC ! .'.y ,' t '
- '.Pd Y s'" I f j ed hf 9.TO Q3,pcryN1nc],
(p J ; t, .
'. c. ir m:: , .
. nr r! icu u are cisjoct to th.
t .l v ! c f .' ec . i ; ' >
1: v. s i ; .W .19 thJ prop .il.r t 3 m' C2 the O' a[j n ' ,
- ' .U P ;; ^.) ;. IN c7 :.h *
+4 ). , ( ,4, ., e-0 O.c .
s
, , . .i. , , -
-3. u. s.-.
4 , i.is
. C s., .1 .
{.t i.L,3. .< . . . , . , .-.,...,2,
.a.
'O ' "
1, ir( .
j'; ' - ' f,y
. ..s" .
- M. .. i ' T 'J f. s ' s; '[.s -*
. .; / ." <
- c ' , 2 ' "5
- ; . !
n, t y . .' ,. ,. - ,,- '.
~. . . . . , . .... 2..,,
, . ., . .9 . . ,
.w . .
3.c s O ,4-
- s. ., s t..,.. u ,..t.,
.s ,
I' 31 )
t
!j l' ' F .l ! e , '!.,
t' ? '.1 "* '; * [ ( ' P dr , -[I {. h, ( 7h3 33C .;C,' ,
,I
,c, c ,
n . . v. c n 6 .t . ... . . . t .1 r, ,. r . .
C, ._.
IIO I
- l. . $. f ,' f ., f $. *
- y e .. , 1 g
{*I--
4 jO; UU)/ L * ?'] ' ' '
l t '. f. "d 3 3. . C'?por;;*tJ, ,,;!rl) at ,,,,f P t ap Of
- p. @ c fvc:. <. : 't. .i : 0 ::c u"h fi . J c.e tion sn i sh! 9n:,
tv . . q p .u;- .
. : a. c r o.. x . . . <. a . I
- $ .. .> k* , " ', - 4 4 E ,h b ' - W %O e g .
i cau @ .. ~s r.c L: . ' uj r h e.~ :, ; ? I,. ncsc a are quali!ica to the bk- k h.
, ' , i 6 S Ie'. 4 +.%- t S 4 .
- I - 4 () h . a o f L N s ,. .! , d i.r) ef r,':'h qunifi ' rienc is tc f Pt.aincJ f c.r all cuch 1 i
pcr t.or , ' c c :: .h - - .m. c . . .
Gi ,c. nP <
...o x .: i n 'i r , eq u f ': , r. t . pr.v;crea end
{
4 l
serviece 1,e in :.:ccor i n:'c
'.'5. h i .c t abl.ished q'. planc, procedures, or i
I 1
1 4
3 7- P.
la : n ;t f - t r. m im!Je w,vran::'.: t:c: t Ll., ite.:in conform to applicabic d r: - , c;'c.:.Jrr..iua , c ad et:, istanJa rd s, ano . rN ula tloam The in t.;- : tec; c a _ definc ec refercr e cypropriate qunatitative criteric
. (J P: ,. ;. , t- race: , .. ..! cperatin;; J J::ite) oc qualitative criteria
(:n .
, it h:p ': -1 :. or '. J ur./.1 standcrds) f or de.terrining th::: . ;r: .:i : . .. '. e bc , e l een saticfc.cteri.ly nece:rpliclut' . The t , , ,~ :c leza <
N (tri; val <.r) } rc*. % spcce fer rc or.lin;, or st:vping i he ', , f.ina: of the in'L :t or a ;' cnables c nc ;o. dei.creia-' the c';ci. ' p:fr $ r. t . ci the its- ( i I d :v+ :. :i .
.:c t ct' Inimcction. are ui . "A h? d _ .4 :. . 1, <
,n;.31 ;
- ri.iti.1
- i E ceh .d.r>D cre r ! .., . i. '. l ; , -j* Tr0 Ll:0 1 :' *i [v.C th:*l r ;l p0. F( ;1*h 3 L1.0 a ,. c fin : .' t.". c.c h ' c. , in.gcm.cJ. T.':w; p en.e ..n .s 1 ) <..p er t to the v+ ,, 6 I ' ' ) f; ' , ca . . *: ; li. -l y iry .:.; cor s , t e r.:t t.cchn f ci: . 7, nnd
,, 1 . . , . ., t .u ..; ;s e ,i ! . . . .
- t. t . ..., 4.....>f,s... u,. . s, . +...s
. ,. . C a i
.,. ...s .
- s. .. s :. 6 .s . ,y. e.c.....m
.. 3 , . . . . .. .,
~.,
. .+.9,.*,
.1 . . .. . .
c : 4; . ,
.4
. L,.,,,-, ,
6 . . *n. s ...
C+..s. u. ..c.,, c I
in ;;' ; .b 'i -i cm _ t .'. ? c '- l'
" . !"'ip ."*.i. 2 Iccults are
... ,. . .....i .. ,
.a, 4
.i4 .. j.
.. ,.,..,.....s,
. . . . a.. .. , . . . ,., ,a. c, r .'. ' t o L v' r i i: . Y u -, : ' '
4 ',, '-'
to 71-
- i ' .: :s > r M .r.c a LM t- ir.E pt r. L lon i e . g. . -:. .. ],.,.,,, q. ,. .,., i ,,
. a u' ,,.,. , . , , , , ...>.,. , 3. . c, L c .. 4
,, y3 ...%,.. , al.c u .
. c.
. ... . . . ~ ..
.. .. c.f tie ! O C'. or- ' n t '. .e m h '. e p rc v - l c. - f' e i h<
- g. <. r. . ,., , s. . . u. , t i , .,, , , ,. .. , , t. , ;.-.o..,.c,.....,2..,...
1 . . . . . . . _ . . . . , , .. . ..a.:. e .,
,u. .. d e '_ u' r. ' .' '.'..'. a ', i t e c . . .,.n.a. ., i ..
- i. ,, i...,,.t.>..<a..t'..>,-
... . . 7F .,, .. . .c. 4m .' , a> u'
_.s" or to "ICpiu ' " . 0 DO.iCN f D er.ir; I t O' . r Of;' U2 Y O i a r s '? .:. ' aJ a;rO"35 I 'a
' l l
u ... c ,. ,i .
- r. .s.
. . . . c. , . , . .. $ , , c ,e .w m. t ......,i. ,.
- _.c.., . 4.r.,,
and Q.\. i:r; hu.13,v tbere mre not adequc,1t precirJ , J. c;nf-for re: rr t irn nonconf er.r n'er c f re f ety re3a t ed .at '- Lore: . v icmo, f
4:er L .s (; c. . ;w .i l .g uite s. n a m.,, . e, t . .i to o' cept c" a f; c:l e,? ( , ;r to tJC 3
17-9 e 1 utility. The staff discussed this conc'ern with GE, and they have amended GESSAR to describe and report nonconformances of safety related equipment to the utility which could affect end use.
Provisions are established for conducting a comprehensive system of planned and documented audits to verify product quality and compliance with the QA Program. The audits are designed to assure compliance with all aspects of 10 CFR 50, Appendix B, including the quality-related aspects of design, procurement, manufacturing, storage, shipment, and reactor site acitivites. The P&QAO, by delegation from the NED General
, Msnager through the Division Product Quality Policy, has the responsibility for the conduct of QA audits of each'of the departments in BWRO. In
^
E addition each BWR0 department is required, by Division Policy, to conduct internal QA audits of its ;roducts and all elements of the BWR QA Program.
\
BWR0 suppliers are subject to audit by cognizant BWR0 QA organizations.
The construction site is audited by a resident Site QC representative from QAEE&I who performs surve111ance of applicant and AE conformance with BWR0 supplied installation and test documents.
These audits are performed to determine the adequacy of QA-related practices, procedures and instructions; compliance with procedures, instructions, and policy directives of the QA Program; the effectiveness of the implementation of the QA Program, procedures, instructions, and i
policy directives; the adequacy of work areas, activities, processes, documents and records; product compliance with applicable engineering drawings and specifications; and implementation of corrective action
( in accordance with applicable procedures.
d 17-10 i
(
17.4 Conclusion I
Based on our review and evaluation of Section 17 of GESSAR we find that the QA Program description provides for a comprehensive system of planned and systematic controls which adequately demonstrates GE's ability to comply with each of the eighteen criteria of Appendix B to 10 CFR Part 50. GE QA personnel are required to be actively involved ( )
in all quality related activities throughout the design, procurement, I f abrication, inspection, testing, shipping, preoperational testing and auditing phases of the BWR. We find that the QA Division has sufficient
" delegated independence and authority to effectively establish and execute their -QA Prograin without undue ' influences from those organization L
elements directly responsible for costs and schedule.
I We conclude that the Quality Assurance Program as described in GESSAR, as amended, complies with the requirements of Appendix B to 10 CFR
. Part 50 and is acceptable.
)
0 l
l l
l
)
l
i
, SER APPENDICES
. 1 Appendix A Chronology 1-Appendix B Effluent Treatment Systems Branch Position No. 1 Appendix C llendrie To Hinds Letter Dated April 19, 1974 Appendix D Technical Report On The General Electric Cgmpany d
8x8 Fuel Assembly, dated 5 February 1974 Appendix'E Review And Evaluation'Of GETAB For Bh'Rs, dated p . ', September, 1974 i
e l
l I
L i
4 I
______________L
APPENDIX A CHRONOLOGY - REGULATORY REVIEW 0F GENERAL ELECTRIC COMPANY'S STANDARD -
SAFETY ANALYSIS REPORT
__. (GESSAR) .
~
Docket No. STN 50-447 April 30, 1973 General Electric Company (GE) letter submitting report entitled " General Electric Standard Safety Analysis Report" (GESSAR), Volumes 1-7 for an acceptance review, pursuant to AEC Policy Statement issued in the form of a press release on 3/5/73.
GESSAR describes a standard 3579 MWt boiling water reactor and consists of safety information for a complete BWR-6/ Mark III containment system. (Project No. 484 assigned).
July 20, 1973 ' '
AEC letter advising that GESSAR is sufficiently complete for us to initiate our detailed. review, l upon receipt of the appropriate number of copies.
I The letter also encloses a list of deficiencies,
, the response for which we should receive as an i amendment within 30 days. )
July 26, 1973 GE letter transmitting additional copies of GESSAR for docketing.
July 30, 1973 GESSAR docketed July 31, 1973 Meeting held in San Jose, California, between AEC and GE representatives.
July 31, 1973 AEC letter advising that GESSAR has been docketed and transmitting a related Federal Register Notice i
July 31, 1973 GE letter transmitting proprietary information (Figures 4.3-22and11.3-2andTable11.3.2)in support of the non-proprietary descriptions contained in GESSAR.
August 6, 1973 GE letter transmitting a notary page insert for GESSAR.
August 8. 1973 Notice of Receipt of GESSAR published in the Federal Register (38 F.R. 21444) a l
t August 10, 1973 GE letter transmitting Amendment No. 1, which (notarized 8/8/73) submits.part of the information requested by AEC letter dated 7/20/73.
August 17, 1973 GE letter submitting Amendment No. 2, which furnishes answers to questions contained in -
AEC's 7/20/73 letter, and other additional information. -
August 24, 1973 GE letter transmitting Amendment No. 3, which .
submits further information in connection with AEC's 7/20/73 letter.
August 31, 1973 - CE letter transmitting Amendment No. 4, which contains additional information requested by letter dated 7/20/73.
August 31, 1973 Meeting between AEC and GE representatives to a
discuss GESSAR p .
j August 31, 1973 AEC letter transmitting the staff's review schedule for GESSAR.
August 31, 1973 GE letter submitting proprietary information to be
, included in GESSAR -- Tables 4.2.1 and 4.2.la, which consist of fuel data and fuel cladding properties, ;
respectively.
September 28, 1973 GE letter transmitting Amendment No. 5, which provides information related to the core power distribution study (Appendix 4A), requested in questions 4.1.1, 4.1.4, and 4.1.6 in AEC's letter dated 7/20/73.
September 28, 1973 GE letter transmitting the proprietary portion of Amendment No. 5 (Table 2 of Appendix 4A and several figures).
October 5, 1973 GE letter transmitting Amendment No. 6, which consists of updated figures of building design and equipment arrangements, corrections of typographical errors and clarification of portions of the text 1 where obvious discrepancies exist. I October 12, 1973 GE letter transmitting Amendment No. 7, which consists !
of revised and new pages, tables and figures. 1
( l 1
1 l
i
I
( I October 24, 1973 GE letter providing a tabulation that itemizes questions posed in other BWR/6 projects (i.e.,
Grand Gdif, Perry, River Bend, Douglas Point.
Allens Creek, Clinton) for which answers will be provided in future GESSAR amendments.
November 1,1973 AEC letter transmitting (Q-l's)a request for additional information.
)
i November 6, 1973 AEC letter requesting information in connection with !
the staff's review of anticipated t.ransients without '
scram (ATWS) in water-cooled reactor power plants.
November 8, 1973 AEC letter granting the withholding of proprietary I information submitted by letters dated April 30, 1973 (Figure 4.3-22); August 31,1973(Tables 4.2.1 {>
and 4.2.la); and September 28,1973 (Table 2 of Appendix 4A and several, figures), pursuant to
~
Section 2.790(b) of 10 CFR Part 2.
November 21, 1973 ..-
AEC letter requesting additional information concerning safety-related and control systems; integrated leak
[i '
rate; site characteristics; General Design Criterion 4; engineered safety features; ciectric power; auxiliary i systems, et al. l i
j November 27, 1973 Meeting held between GE and AEC representatives to {
discuss GESSAR schedule, technical matters to be resolved, and the future use and expected benefits ]
of GESSAR.
I December 7, 1973 GE letter reaffirming its position that proprietary )
Figure 11.3-2 and Table 11.3.2 should be treated in l accordance with 10 CFR 2.790. '
. December 11, 1973 Meeting between AEC and GE representatives to discuss
. the seismic design of GESSAR. l l
December 12, 1973 AEC letter requesting additional information. {
December 13, 1973 Summary of meeting held on 11/27/73.
l December 14, 1973 GE letter transmitting Amendment No. 8, which responds i to questions forwarded by AEC's letter dated 11/1/73, '
including 16 pages of proprietary information (submitted by letter dated 12/17/73). Amendment No. 8 identifies
(
i I q
i I
all interfaces between the nuclear island and the-balance-of-plant.
December 28,.1973 GE lettbr s'ubmitting information relating to .
interface and electrical' areas requested in the 11/27/73 meeting.
January 11, 1974 GE letter transmitting Amendment No. 9, which' responds partially.to AEC letter dated 11/21/73-(part2ofQ-1 list).
Janua ry 18, 1974 GE letter transmitting Amendment No.10, which is in reference to AEC letter dated 12/12/73.
January 22, 1974 . Meeting held between AEC and GE representatives to discuss LOCA flooding of the containment drywell AP test, suppression pool swell, and liner corrosion allowance.
February 8,1974 GE letter tr'ansmitting Amendment No.11, which
. ,, addresses all quest}ons raised by AEC in the first ,
round of questions and identifies those questions
, for which residual information will be provided.
- h. , Amendment 11 includes proprietary information.
February 12, 1974 AEC letter expressing concerns about delays that have occurred in the review to date of GESSAR and transmitting a revised review schedule.
February 19, 1974 GE letter con.sisting of a rebuttal to AEC's letter dated 2/12/74.
February 19, 1974 GE letter responding to AEC's letter dated 11/6/73 regarding ATWS.
February 27, 1974 Meeting held between GE and AEC representatives to discuss the GE scram system including the control rod drive position detection and indicating system; the rod pattern control system; and the use of ganged control rods.
February 28, 1974 Meeting held between AEC and GE representatives to discuss the adequacy of GE's responses to the interface questions.
March 15, 1974 Meeting held between AEC and GE representatives.
Areas of discussion included design details beyond
(
3H q, .
. f
. the scope of the standard format, new regulatory positions, outstanding items and resolved items.
March 20, 1974 AEC letter requesting' additional information (Part 1 of Q-2).
March 26,1974 AEC letter transmitting request for additional *
~ informatin (Part 2 of Q-2).
April 3,1974 Meeting held between AEC and GE representatives to discuss site parameters.
April 11, 1974 : AEC letter requesting ' additional information (QA program included) and discussing concerns related-to the availability of information needed on preliminary instrumentation' design, as discussed in meeting of 2/27/74.
April 11, 1974 AEC letter granting the withholding of proprietary (1) Figure 11.3-2 and Table 11.3.2 on the basis' of a
reasons contained in letters dated 4/30/73 and ~
, ", 12/7/73; (2) pages (14) and responses to AEC questions 4.71 and 4.72 submitted by letter dated:
'. '12/17/73 (amendment No. 8); and,(3) page 6.2-116 h' , responding to question 6.86 submitted by letter dated 2/8/74 (Amendment No'. 11).
! April 17,1974 Meeting between AEC and GE representatives to discuss-site parameters:
April 19, 1974 GE letter transmitting Amendment No.12, which clarifies inconsistencies in Chapter 7.
April 19, 1974 GE letter commenting on use of GESSAR questions on projects.
April 25,1974 GE letter stating its position in connection with
, the information requested by AEC on 4/11/74.
May 1, 1974 Issued summary of meeting held on 4/3/74.
May 2, 1974 Issued summary of meetings held on 2/27 .2/28/74; and 3/15/74 (summary relating to 2/27 - 2/28/74 meeting, dated 5/1/74; summary for 3/15/74 meeting dated 4/30/74).
May 7, 1974 AEC letter requesting additional information relating l to Sections 7.22 and 5.39.
May 10,1974 GE letter transmitting Amendment No.13, which answers I
Q-2's.
b l
l l .
i May 14, 1974 Issued summary of meeting held on 4/17/74.
May 17, 1974 GE letter submitting Amendment No. 14, which consists of proprietary GESSAR pages 4.2-10ld, 4.3-42c, and a 14 page report entitled " Fuel Rod Heat Transfer Model." Amendment 14 is in reference
, to AEC letters dated 3/20, 3/26-and 4/11/74. .
May 30, 1974 AEC letter advising that AEC review of the new i
Section 1.11 for Chapter 1, which was included in Amendment 13, will not commence until the fall of this year.
1 June 17, 1974 GE letter transmitting Amendment No. 15, which responds to AEC letter dated S/7/74.
June 11, 1974 Meeting between GE and AEC representatives for the purpose of discussions relating to drywell structural proof test and high pressure leak test. AEC staff l reaffirmed its position with respect to both tests.
June 27, 1974 . . Issued summary of meeting held on 6/11/74.
June 28, 1974 GEletters(2)submittingthenon-proprietaryand g-1 , proprietary (Question 4.15) portions of Amendment No.16, ;
which is in reference to AEC letter datea 11/1/73.
July 1, 1974 ACRS subcommittee,e meeting held.
July 2, 1974 AEC and GE representatives met to discuss Reg. Guides 1.31 (testing for weld delta ferrite) and 1.44 (testing for non-sensitization of welds).
July 12, 1974 GE letter transmitting Amendment No. 17, which clarifies and updates portions of the text; and provides information relating to main steam line leakage control system.
July 29, 1974 AEC letter making proprietary findings on May 17 &
June 28 submittals.
July 30,1974 AEC letter requesting additional information on Chapter 7 of GESSAR.
August 2,1974 Amendment 18 filed.
August 16, 1974 Meeting .to discuss schedule and procedural considerations.
August 23, 1974 Amendment 19 filed.
(
_ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _a
'*w 7--
, August 30, 1974 Amendment 20 filed.
September 12, 1974 Letter to GE notifying them of a one month delay in publication of our SER due to the number of outstanding items.
September 27, 1974 Amendment 21 filed. This amendment addresses the outstanding items listed in the September 12, 1974 *
. staff letter to GE.
e e
h l
l
- ~ '
- N m
e e
(
1 i
1 L ,
L .,
. Appendix B -
f 8/28/74 l
EFFLUENT TREATMENT. SYSTEMS BRANCH POSITION NO.1 Directorate of Licensing Design Guidance for Radioactive Waste Management Systems Installed in Light-Water-Cooled Nuclear Power Reactor Plants .
, I. Liquid Radioactive Waste (Radwas te) Sys tems, Inc,1uding Steam Generator Blowdown Systems l a. The portions of the steam generator blowdown system ~
extending from the steam' generator to the outermost containment isolation valve should be' designed to' s a
. Quality Group B and to seismic Category I.
- b. The, liquid radwas te treatment sys temb including the
( .
steam generator blowdown system downstream of the second
'I containment isolation valve should be designed to the following:
(1) " Quality Group 'D (Augmented)" should be applied-to all, equipment, interconnecting piping and com-ponents except those specified in I.b.2 and 1.b.3 below.
(2) Quality Group D classification may be applied to equipment and interconnecting piping and components i
1/
For the purpose of this guide the liquid radwaste treatment I system does not include sumps and floor drains provided for collecting liquid waste. Discharge of liquid waste directly to the environment is unacceptable.
(
B-2 ,
g containing radioactive materials of sufficiently low concentrations that release of their contents !
to the environment would be permitted if the inventory limits are within those specified
, in the plant Technical Specifications (e.g., waste I sample (test) tanks containing liquid wastes which have completed radwaste processing). .
The maximum radioactivity to be contained in any liquid radwaste tank that can be discharged directly to the environment should not exceed 10 C1,
, excluding tritium and dissolved gases.
(3)"' Quality Group D classif"ication may be applied to
['
i co11cetion tanks and components used for detergent waste (laundry, personnel, and equipment decon-tamination). If detergent wastes are co11ceted separately but processed in the floor drain or l miscellaneous dastes subsystems, the necessary equipment, interconnecting piping and components I
should be designed to the criteria given in I.b.1 above. !
i (4) Materials for ' pressure retaining components should conform to the requirements of one of the specifications for materials lis ted in Tables I-7 and I-8 of Appendix I of Section III of the ASME Boiler .
and Pressure Vessel Code, 1971. Other materials should be limited to those permitted in ND-2121(b),
(
I O
8
4
-+ .
.B-3 I
'(d) and (c) of Section III of the ASME Boiler ,
1 i
. and Pressure Vessel Code of 1971, except that malicabic wrought or cast iron materials and l;
, plastic pipe should not be used. Materials-Manuf acturer's - Certificates of Compliance with l this material specification may be provided in lieu of Certified Materials Test Reports.
(5) Foundations and adjacent walls of structures that house the liquid radwaste system whose failure a may affect the co11cetion, processing and storage
, ,. t equipment and components should be designed to l' ;
seismic Category I.
(6) Equipment and components used to collect, process,
.t and store liquid radioactive waste, except as noted in I.a above, need not be designed to seismic Category I.
(7) The liquid radwaste treatment system, including the s team generator blowdown system, should be designed to prevent uncontrolled releases of radioactive materials due to spillage in buildings or
- from ou'tdoor storage tanks which may contain radioactive material. Rctention capabilities should be provided for:
(
4
B-4 1
(i) potential spills from outdoor tanks by installing dikes or grading to form retention ponds. Dikes and ponds should have provisions for routing spillages to the liquid radwaste treatment system. Dikes should be designed as seismic Category I structures.
(ii) potential spills from indoor tanks by connecting
~
floor drains to the liquid radwaste treatment system and installing curbs around equipment
, and tanks er elevating thresholds.
"* (iii) for all tanks con [aining potentially radioactive l . materials both inside and outside the plant including the condensate storage tank (s) provisions should be made to monitor liquid
, levels, to alarm potential overflow conditions, and to collect'and sampic liquid overflows.
II. Gaseous Radioactive Maste (Radwaste) System The gaseous radwas te treatment sys tem, including the treatment of normal of fgas releases f rom the main condenser vacuum system for a BWR and the treatment of gases stripped from the primary coolant for a PWR should be designated to the following:
- a. " Quality Group D (Augmented)" should be applied to all equipment, interconnecting piping and components used to to process gases containing radioactive materials.
(
6
B-5
\ -
- b. Materials for pressure retaining components should conform to the requirements of one of the specifications
+
for materials listed in Tables I-7 and I-8 of Appendi):
I of Section III of the ASME Boiler and Pressure Vessel Code,19 71. Other materials should be limited I
to those permitted in ND-2121(b), (d) and (e) of Section III of the ASME Boiler and Pressure Vessel Code of 1971, execpt that malleabic wrought or cast
. iron materials and plastic pipe should not be used.
Fbterials Manuf acture,r's Certificates of Compliance a
,witl.this 3 material specification may be provided in
'. Ifeu of Certified Materials Test Reports. t l
I
- c. Equipment and. components used to collect, process or store gaseous radioactiv,e vaste, and the structures that house gaseous radwaste collection, processing and storage equipment and components. need not be designed to seismic Category 1 except as noted in II.d below.
- d. Those portions of the gaseous radwaste treatment system which by design are intended to store or delay the release of gascous radioactive waste, including portions of structures housing these systems should be designed to scismic Category I classification. This should include isolation valves, equipment, interconnecting
(
piping, and components located between the upstream f
k
B-6
( .
and downstrcam valves used to isolate these components .
from the rest of the system (e.g., charcoal delay tanks in a BWR and waste gas storage tanks in a PWR). ,
III. Solid Radioactive Waste (Radwaste) System ,
The solid radwaste system consists of slurry waste co11cetion and settling tanks, spent resin storage tanks, phase separators, and tanks equipment and components used to solidify aqueous and non-aqueous liquid waste prior to offsite shipment. The solid radwaste handling and treatment system should be designed a to the following:
.. .. /
, a. " Quality Group D (Augmented)" classification should be
[ applied to equipment col.lection and settling tanks, evaporn-tor concentrate tanks , spent resin storage tanks, and I
equipment and components used to transfer recovered liquids from these tanks to the 1.iquid raduaste system, except for equipment and components specified in III.b, below.
- b. Quality Croup D' classification may be applied to equipment and components used to solidify aqueous and non-aqueous liquid waste.
- c. Materials for pressure retaining components should conform to the requirencnts of one of the specifications for materials listed in Tabica I-7 and I-8 of Appendix I of Section III of the ASME Boiler and Pressure Vessel Code, 1971. Other j
. 1 I
I e
i.u-- .-_- - _m._-
l l
l B-7 l
l materials should be limited to those permitted in ND-2121(b),
]
(d) and (e) of Section III of the ASME Boiler and Pressure Vessel Code of 1971, except that malleabic .
i wrought or cast iron materials and plastic pipe should I
not be used. Materials Ibnuf acturer's Certificates of Compliance with this material specification may be provided in Ifeu of certified materials test reports.
- d. Equipment and components used to collect, process or store solid radioactive waste need not be designed
, to scismic Category I.
- c. Foundations and adjacent walls of structures that house l ,
the solid radwaste system whose failure could affect i !
the collection, processing and storage equipment and i
components should be designed to seismic Category I.
IV. Definiation of " Quality Group .D (Augmented)" !
In addition to the requirencnts inherent in the codes and standards listed in ' Regulatory Guide 1.26 for Quality Group D, the following criteria, as minimum, should be impicmented for components and systems designated as " Quality Group D ( Augmented)"
in this guide.
- a. The Quality Assurance provisions described in V of this guide should be applied.
- b. Pressure retaining components of proccss systems should
~
i utilize welded construction to the maximum practicable extent.
(
4
_ _ _ _ _ _ - ---________--.w
B-8 f/
Flanged joints o' r suitabic rapid dis connect. fittings should be used only where maintenance or operational requirements cicarly indicate that such construction is preferable.
Screwed connections in which threads provide the only seal
,, should not be used except for instrumentation. connections where welded connections arc'not suitable. Process lines should not be icss than 3/4-inch. Screwed connections ba'cked up by seal welding, socket welding or mechanical joints may be used on lines greater than 3/4-inch but 1 css than 2-1/2-inch nominal size. For lines of 2-1/2-indi nominal . pipe size and above, pipe welds should be of the butt-joint type. Backing rings should not'bc used in l' '
I lines carrying resins or other particulate material.
. All welding constituting the pressure boundary of pressure retai.iing components should be performed by qualified-welding procedures in accordance with ASME Pressure and Vessel Code Section IX.
9
- c. Completed process systees should be pressure tes ted to. the maximum practicabic extent. Piping systems should be hydrostatically tested in their entirety utilizing temporary plugs at atmospheric tank connections. Testing of piping systems should
, be performed in accordance with ANSI B31.1, ASME NH-6111.1 and NB-6111.2, but in no case icss than 75 psig. The test pressure should be held for a minimum of 30' minutes with no leakage indicated.
- __.__-___.___.-___.___-__._._.______-._-__a
I B-9 V. Quality Assurance for Radioactive Waste Fbnagement Systems A program shall be established that is sufficient to assure that the design, construction, and testing requirements are met. .
The following area should be included in the program:
- a. Design and Procurement Document Control - Measures should be established to insure that the requirements of this design guide are specified and included in design and pro-curement documents and that deviations theref rom are controlled.
- b. Control of Purchased !!aterial, Equipment and Services -
"casures should be established to assure that purchased Y
mat eri dl , equipmen t and ennstruction sorvices conforn to ,
i the procurement documents.
- c. Inspection - A program for inspection of activities affecting quality should be established and executed by, or for, the organization perf orming the activity to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity,
- d. Ilandling, Storage and Shipping - Measures should be established to control the handling, storage, shipping, c1 caning and preservation of material and equipment in accordance with work and inspection instructions to prevent damage or deterioration.
( l l
4
r i
1 I
l i
I B-10 t
]
- c. Inspection, Test and Operating Status - Measures should be established to provide for the identification of items which have satisfactorily passed required inspections and tests. .
- f. Corrective Action - Measures should be established to assurc that conditions adverse to quality, such as failurcs, mal-functions, deficiencies, , deviations, defective material and equipment and nonconformances are promptly identified and corrected.
4 e j
- l. '
1 i
e 9
e t
0
(
1 j
4/ i i h1.
,.v- / "Y ATOMIC ENERGY COMMISSION l l' I % .
WASHmG'rON, o:C.* 20m Q) ,lJ,*. ~.C. ';[ t '
hN:4#b_W f-Appendix AFH 1 UC../4li I .
\
. )
1 Mr. John A. Hinds, Manager Safety and Licensing ,
' Ceneral Electric Company 175 Curtner Avenue San Jose, California 95114
Dear Mr. Hinds:
At various times the AEC staff has discussed with the General Electric Company the subject of appropriate classification requirements in boiling water reactor (BWR) plants for main stcam system components. These discussions have included consideration of (a) structurcs, cocponents and systems that are not casssified as saf e ty-relhted items but are located downstream of the isolation valves; (b) those not specifically designed to seistic Category I standards; and (c) those not housed in s'eismic Category I structurcs.
[
To date, BWR plant reviews have resulted in various ap p ro ach e s I
for different individual applications. While these different i approaches have resulted in. acceptable levels of saf ety in each case, they 'have required time-consuming customized revicus.
The GESS.4R BUR /6 ap' plication, under review as part of our standardization program, includes this portion of the BWR plant.
In the course of the GESSAR review, we have identified a systematic basis for classification of such components and structures that will result in an acceptable and uniform design basis for the main steam-lines (MSL) and main feedwater lines (MFL) in the standardized plant. Although it is recognized that a significant portion of the equipment involved in this classification scheme may include equipment outside the normal scope of supply of the General Electric Company, specifically the shutoff valves in the main secao and feedwater lines and the equipment beyond those valves,
' the impicmentation of these requirements defines acceptable standardi=cd re q ui re =en ts with respect to quality and seismic design for the BITR /6 nuclear s team supply cys tem and the power ceaversion systeu.
(
__.______-_U
~
4 APR1 d w/4 C-2' - * *
- J. A. Hinds .
This approach' involves specification of app rop ria te s afe ty .
requirements for those portions of the'MSL and ((FL that I are' housed in Seismic Category I structures (e.g., the auxiliary b uilding) , and includes suitable restraints near a shutoff valve outside the containment-isolation' valves. .
' The portions of the:MSL and MFL.on the turbine side of the shutoff valves would be designed:in accordance with quality group and certification procedures as outlined in Attachment A of this letter.
To' implement this approach, Attachment 1 of'our letter of Movember 19, 1973, would be amended.to provide acceptable seismic and quality requirements f or BWR/6 s team sys tem' components as shown in the attachments to this letter.
Uc believe1this provides an appropriate standardized approach'to MSL and MFL classificati.on and is acce.ptable as an. alternate to the guidelines currently specified in Regulatory Guides 1.26 (March 23, 1972) and 1.29 (August 1973). .
a .
- As we have discussed with #
you, a suitable intarface~ restraint should be provided at the point of departure from the C1hss
_i I structure where the in(crface c::is ts be tween the safeCy t ,
and nonsafety-related po tions of the MSL and MFL.
t idccrely, k
Joseph M. Hendrie, Deputy Director for Technical Review Directorate of Licensing
Enclosures:
Attachment A (Classification Requirements for Main Steam Sys' tem Components Other than the Reactor Coolant Pressure Boundary).
Attachment B (Ske tch - AEC Quality Group and Seismic Category Classifications Appliccble to Main Steam System Components in BWR/6 Plants) cc: L. S. Gifford -
i
_ _ _ _ _ . _ _ _ _ _ - _ . _ - - - - - - - - - - - - - - - - - - ' - - ~ "
a z i m u n c.n i x
-Classification Requirements for EURf6 Main Steam and Feedwate'r !
Sysrco Couponents Other than the Reactor Coolant Pressure Boundary. f
(
ITEM QUALITY CRj SYSTEM OR COMPONENT
_CLASSIFICA!
1.
' Main' Steam Line (MSL) from second 1 solation i valve to and including shutoff valve. B 2..
Branch lines of MSL between the second B
isolation valve and the MSL shutoff valve,
' from branch point at MSL to and including the first valve.in the b ranch line. !
3.
Main feedwater line (MFL) from second isolation B valve and and including shutoff valve.
4.
Branch lines of MFL betueen'the second B isolation valve and the MFL shutoff valve, from the branch point at MFL to and in.cluding the first valve in the branch line. '
5.
Main steam line piping between the MSL shutoff D (1; valve and the turbine main st'op valve. l
'6. Turbine bypass piping ,; .D I '
7.
Branch lines of the MSL between the MSL shutoff D valve and the turbine main stop valve.
t .
- 8. T.urbine valve, turbine control valve, turbine D (1, bypass' valves and the main steam leads from the dr turbine control valve to the turbine casing. Certificate.
- 9. Feeduater sys tem components beyond the MFL D shutoff valve.
(1) All inspection records shall be maintained for the life of thi plant. These records shall include data pertaining to qualification of inspection personnel, examination procedures and e::acina tion res ults . .
(2) All cast pressure-retaining parts of a size and configuration for which volumetric methods are effective shall be examined by radiographic methods by qualified personnel. Ultrasonic examination to equivalent standards may be used as an alternate to radiographic methods. Examination procedures an acceptance standards shall be at least equivalent to those defined in Paragraph 136.4, Examination Methods of Welds -
Mon-Boiler External Piping, A t!SI B31.1-1973.
(
..e . . .
O
> y e
- l 1
C-4 2..
1 I
ATTACHt1ENT A
( ,
(3) The following qualification shall be met with respect to th.
certification requirements:
- 1. The manufacturer of the turbine stop valves, turbine control valves, turbine. bypass valves and main steam leads from turbine control valve to turbine casing
- shall utilize quality control procedures equivalent to those defined in General Electric Publication GEZ-4982A, " General Electric Large Steam Turbine- 1 Generator Quality Cont rol Program". l
- 2. A certification shall be obtained from the manufacturer of these valves and steam loads that the quali ty con t r o."!
program so defines.has been a, accomplished.
J . .
/'
O b e es *
' m e 4 v I ,
~
l l
4 e
l
C-5 l at' r' g c-
. p D *
' O
[
r
)
.(
J r
t M
h d
O
> >nA I Z .d' .
a p P A 2 -
c d %. dc g >
g 2 -
0 0 et 2* O - e ., ~ o -
F ,
p Q W O Y 2 % & y 1 op r ir e U E c u l
. yc tu
{ vg.
- 3 j t) Z g
l I - .t er m^
, c-
> .S V W 6
a g
- Z T to U
V
- g. .
$ c ci -Z .. a < ~. D L .
n i 2s ? ~ 2, 8 o 2 ?
e i
1 c
t 2 8
" "v.
ko g m, o e e
O' El -
r 2
n .-
e, o, --- j
.I g- d, u '
c.
Y o '>- i.:
l J i . p 4
%$ U" - 3 %- h" f 0
>- . 2 i, u
. o g 0 t a l .
{ a y s v =u-c 2 , w .
. e e.e o
o t,-
c .
T z v Ve
.a. i, H *>
of C 0 F W U '
d . 3 * . - .
\
l 3 *$' U O b I
a ," 4 g ..
,6
--- P t--
0 C
3 CI' Q d x ,, :
, ,, 0 - p.
W'" r
\ Q @ -
t M 'h b
A- .. ., ' '
O e/ .&
a '$ :
1-- --
?. V 4 , i.
.> D. .
l ', Z 7' -
Di g, .. ~E Z E
- g
. i *> f g, + >rl o
a w o4 3
J u d-Lc g::
' " hl ~8 f
m
/
(
O,.
U V -
. 3 2, l 7 4 G C$
f & s
$ 3 3 ~ > U$
b y -
'h 1 d h .~1' C-
, l.d co v zu i . o .> e k *J d y. = y} E:': !! ~~, d 0-v I # -
O# s d z E!
a 3l 2 f .--
a a
4 .O
.- C-z e -
' U t- u.< 2 o .E g :: g c wl U ; c. . n . .
- c. o x I d o o
il c. lj . . . . . - . . . - 3 f' 1I g- -
n Y L. -
7 .. _. e e ...0 5.Z O r. d w F.
3 o u .;
5< '
i }
$ C E c-
.m
.22 x-.__I _ _
- U s y U dO
" h -
'3G .? U <
- d "F- .. OU M > _d d G
0-D-
'2 1 1 .J ;- '3 u < -
p a 4
> 0
. E 0 g r com en t -- U 3 C o & d O e - r 2 t- o W Q C 4 J
g, 0
m e
it
1 1
Appendix D j
( ,
l 1
I
' l
< i e
4 j
k Technical Report On
, the {
Ceneral Electric Conpany a . ,
8 x 8 ruel Asse,.mbly I ,
1 5 February 1974
. Regulatory Staff -
U. S. Atomic Energy Cennission l I
$ e 0
4
- -- __--__._-x.-___
D-i ERRATA Technical D.eport on the General Electric Company 8 x 8 Fuel Asser.bly .
, , Dated February 5,1974 Pg.1 - last line change " evaluation" to " evaluations" Pg. 2 - line 8 add "and" af ter word " diameter;"
Pg. 4 - line 3 change "0.038" to "0.060." , ,
~
Pg. 5 - line 6 change "two" to "one" -
i .
. Pg. 6 - Line 2 delete "(7)" i Line 4 delete "(7)" .
Pg. 7 - Line 16 change "one-half" to "up to three quarters" Pg. 9 - line 6 change "effect" to " affect" Pg.11 - line 1 change "2.3" to "2.10" '
Pgs 20 - line 14.".less severe" to "similar"; line 15 change "than" to "and" to read "... the consequences of these events are similar for S x S assemblies and for 7 x 7 assemblies."
I Pg. 25 - last line change "not" to "no" l .
+
0 8
e 9
4 1
1
( -
0 8
e=
1 .
, ,_ . D-li' t
Table of Contents Page 0
/
1.0 Introduction ................................... 1 2.0 Mechanical Design .............................. 2 3.0 H uci c a r De s i gn . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 4.0 Th e rina l - Hy d ra ul i c . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 .
5.0 Abnorca l Ope ra ti onal . . . . . . . . . . . . . . . . . . . . . . . . . . 19 6.0 Accidents ............../...................... 21 6.1 Rod' Drop Acci dent . . . . . . . . .$ . . . . . . . . . . . . . . 21
'. 6.2 Refueling Accident'f. . . . . . . . . . . . . . . . . . . . . 21 6.3 L os s- o f- C o ol a n t ". . . . . . . . . . . . . . . . . . . . . . . . . . 22 l 6.4 S team Li n e B rea k . . . . . . . . . . . . . . . . . . . . . . . . . 22 7.0 References .................................... 29 t
Table I Mechanical Des i gn Compari son . . . . . . . . . . . . . . 3 Table II Hacl ear Des i gn Compari s on ~ . . . . . . . . . . . . . . . . 10 C h r o n ol o gy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 I
a Y
1 D, .
I 1.0 Introduction ,
The current fuel in General E1cetric Company boiling vater reactors is sintered, slightly enriched uranium dioxide pellets . sealed in Zircaloy tubes. Bundics of these fuel rods are contained within a squarc open-ended Zircaloy channel box to form fuel assemblics. The General Elcetric Co. has recently modified the design of these fuel 1
assemblics and licensecs propose to reload assenblics of this new (1, 2, 3, 4, design as replacements for depicted assemblics of the old type; This report presents the results of the Regulatory Staff's gencric review of 8 x 8 fuel assemblies as used both in partial and full core l l
J < , l reloads. As' part of the stcff's review of the General Electric Company MTR-6 c1mu4 of icactors, which are currently under consideration for l
[
construction permits, the Staff is continuing its review of the 8 x 8 fuc1 l
t assemblics used in those new reactor designs. The Staff's review of reload assemblics considered the ef f ects t hat the changes in the fuci design have on normal operation, abnormal operational transients and accidents. IIoucver,. the Staf f review considered only generic aspects of the fuci design .such as the adequacy of design methods, the comparative performance of'the old and new fuel designs, and the applicability of accident analysis methods. The plant specific aspects of the review, J
such as compliance with the interim Acceptance Criteria, including the effects of fuel pellet densification, any necessary revisions to Technical Specification requirements, and the radiological consequenecs of postulated accidents will be addressed in separate evaluation for the individual plants.
i 1
, . l D 1 2.0 Mechanical Design
'Ihc reload fuel assemblies consist of 63 fuel rods and one unfueled, l
capture-spacer rod in a squarc 8 x 8 array within a square channel
, l box. The rods are spaced and supported at the top and bottom by stainicss steel tic plates. The rods are also hcid in alignment by spaccr-guides located along the assembly. As shown in Tabic I the S x 8 f uel assembly is similar to the current 7 x 7 design. The major ecchanicci changes are the larger number of rods; the reduct'.on in the rod dinmeter; the introduction of the asymmetrically located' t nfueled a
spaccr-c$ptur'c' rod; and the use of fuliy anncaled, rather than cold
~
worked, Zircaloy cladding. Other changes, which have also been in-corporated in the most recent 7 x 7 designs include shorter, chamfered and undished pellets and a hydrogen getter. However, the designs of b6th asncablics have the same objective, that is maintainance of clad integrity during normal operation and abnormal transients. The designs of both are also based on the same stress criteria, that is, the ASME Eoiler and Pressure Vessel Code,Section III. In evaluating the performance of the fuel, the design analysco considered strecscs due to external coolant precsure, internal gas pressure, thermal ef fects, spacer contact, and flow induced vibration. Other effects which were considered included pellet-cladding rechanical interaction, stress corrosion cracking, fretting, .
and densification. Verification of the adequacy of the design of the 8 x 8 assemblics is based on analysis, mechanical tests, operating experience of previous designs, in-pile tests of a prototypical f uel rod and similar f
fuel rods, and an out-of-pile test of att assernly of similar design.
i D .3 l
)
TAP;LE I i MECilANICAL DESIGN COMPARISON . ,
ASSD!BLY Rod Arrcy 7x7 8x8 ,
1? umber ci Fueled Rods 49 63 Rod Pitch, Zn. 0.738 0.640 FUEL ROD Active Fuel Length, 10 144 144 Cas Plenu:n Length, In. 11.25 11.25 -
Fill Gas lle lic ,
rdEL ,
Material UO U0 2 2 a "
Pc11c t' Dic=c't cr, In. .
/*0.477 0.416 Pelict In:acrsion Density, % TD 95.0 95.0 l CL/.DD1: C licterf cl Zr-2 Zr-2 Thichness, In. 0.037 0.034
! Outaide Diameter, In.
0.563 0.493 Clush'i EL ,
Material Zr-4 Zr-4 Thickness, In. 0.080 0.080 Outside Dicension, In. 5.438 5.438 Length, In. 162 1/8 162 1/8 SPACERS Nu .b er .
7 7 Material Grid Zr-4 Zr-4 Springs Inconel. Inconel.
l '
Iluch of the previous experience with fuel rods and assemblics is g
(6)
These rods ranged in diancter applicabic to the 8 x 8 fuel assemblics.
' from 0.344 to 0.593 inches, in clad thickness from 0.022 to 0.088 inches, Rods have and in pellet-clad diametral gap f rom 0.002 to 0.016 inches.
been irradiated for up to 6 years and had peak exposure of 30,000 l
Although rods identical to the S x 8 design have not been tested IfWD/T.
by GE, the background of experience is sufficient to enable GE to design rods of niu design with confidence in their durability.
Confidence that the vibration and fretting characteristics of the (7) 8 x 8 assemblies are knovg is based on rod vibration c'xperiments and the
" operating' cxperience with other types of fuel assemblies in general and the 7 x 7 design in particular. The 7 x 7 and 8 x 8 assemblies are very I
I similar in this regard. The fuel rods in both are of similar design, 5
are made of the same material and have nearly the same natural frequency.
The fuel rod spacer grids in both types of assembly also are of similar
~
design, are made of the same materials and exert the same spring force.
l Both operate at the same pressure and temperature with nearly identical fluidvelociticsandkuality. >
' Further verification of the adequacy of the design has been provided by the testing of an assembly of similar design for 7000 hours0.081 days <br />1.944 hours <br />0.0116 weeks <br />0.00266 months <br /> in h'igh (6) pressure, two-phase flow loop. This test was perforced by ASEA-Atom, a Swedish gWR manufacturer and a General Elcetric Company licensee, as part of a fuc1' development program.
D 4 A compe.rison.of the significant parameters of this test assembly (8) and the GE 8 x 8 assemblies indicate that the wear and fretting characteristics would be similar. The most significant differences are that the test assembly had no unfueled spacer-capture rod, and had, four latern springs supporting a fuel rod, where the GE asserrSlies
.have only two. However, the vibration and fretting in this test would be expected to be at least as severe as in a GE 8 x 8 assembly since the axial pitch of the spacers was larger and the rods thinner walled and smaller in dicmeter. Inspection at 1.5 month intervals and the conclusion of the test revealed no significant fretting' ucar.
' Although the design of the unfueled spacer-capture rod is new, it is I
based on experience with similar designs. Five G x 6 fuel assemblies with i eccentrically located fueled spacer-capture rods which have a locking tab design identical to the 8 x 8 design have operated in the Humboldt I
Bay reactor. Visual examination of these assemblies has revealed no 1
deficiencies. Asse:Slies with cccentricclly located fuel spacer rods with a different iceking tab design have eperated in the Dresden-1, KP3, !
. 1 Tarapur and Garigliano reactors. Twenty four assen61ies with unfueled 1
rod have operated in the Big P.ock reactor, j
(
A number of mechanicci tests have been performed on 8 x 8 fuel
' assemblics and conponents in order to demonstrate their integrity. Dead-weight loading of the 7 x 7 type assembly spacer grids has dennnstrated g
that they are adequate to withstand all expected loads. Although, as GE l 1
has stated, the C x 8 assembly spaccr-grids are stronSer than 7 x 7 l l
D -6 spaccr-grido this vill be verified by dead wcf ght crushing tests of (7) the new spacer design. Testing of the spacer locking tab has shown that it can satisfactorily resist a shcar load, and further verification (7) tests which more closely simulates the actual loading are to be performed.
Other tests have been made to determine the bending stiffness of the acccmbly and the force excrted"by the iuol rod expansion springs. In addition,a channel box removal.and replacement test'and an instrumented sipping test have been performed. The methods used by CE to calculate the effects of fuel pelict (9, 10) dcncification end cladding creep have been previously sub:nitted and reviewc l (11) a by the Staff.' These methods and the/ Staff conclusions apply equally to both the 8 x 8 and 7 x 7 fuci designs. I Performance of the fuel!.during operation will be indirectly monitore l' n during operation by measurement of the activity of the primary coolant and condenser steam air-cjector off -gac. During refueling fuel assemblics will l be tested for radioactivity lechage, selected assembif es will be examined vicually for bowing and some rods will be given standard non-destructive and destructive examinations as part of the norm.21 fuci surveillance l (C)
- program. ' Additional surveillance of ten rods in each of two 8 x 8 fuel assemblics will also be performed. This program will consist of din,cncf onally characterizing the rods prior to irradiation and then the
l visual dimensional, ultrasonic and cddy current examination of them during j (8) l during each refueling.
t' l i . _ _ - _ _ - _ _ _ _ _ _ - _ _
D
- Accident induced loads and strecscs have been calculated for both the 7 x 7 and 8 x 8 assemblies using the same methods. The limi' ting accident loads result from a steam line break. The pressure differences following a,cteam line break are 1 css than 10% greater than normal
- operating pressure differences. As in normal op'c ration, the pressure differenecc in an 8 x 8 assembly following a steam line break are 5 to 10% greater than in a 7 x 7 assembly. The loads following a steam line breal; are well below the allowable loads.
The behavior of the tuo fuci designs under scismic loading is nearl'y identical. This is so because the stif fnens of the fuel channel and the a . . ; ucight of the fuel assembly are the same for both desi ,ns. t Only thesc l rwn pat u..uten, need tc tc con,idered since the ctiffness of t he bundle of fuel roda is small compared to the char.nc1, and the cicarance between the channci and the rod bundle is small compared to the limiting de-flection of the channels. The predicted loads from the postulated safe chutdown carthquake are one-half the allo abic loads. Uc conclude that based on operating experience with similar fuel, the results of an out-of-pile test of a acccmbly of similar design, the increased thornal margins which the 8 x 8 fuel has, the Technical Specifi-I catica requirements to monitor and limit off-gas and coolant activity,
. and the exictence of a continuing fuel rod surveillance program which includes destructive and ncn-destructive port irradiation examinations, the cladding integrity of the C x 8 fuel will be maintained during normal operation and abnormal operational transients and significant n=ounts of
'l
. l
. - _ --_-_-________---_L
i l~ e p f radioactivity will not be relcaced. Furtlicrmore,we conclude. that accidents or carthquake induced loads will not result inan inability to cool the fuel and safely shutdown the reactor. l f 8 4 5 J * . ,e t *
. l 0
i e t 7 _ _ _ _ _ _ _ . _ _ - _ _ - - _ _ _ - _ _ _ _ _ - _ _ _ _ _ - __=_-
D ' 3.0 , lluclear Design The nuclear design of the 8x8 reload assemblies is simila'r l to that of the equivalent 7 x 7 reload assemblies as shown in Table II . The U-235 enrichments Yor the individual fuel I rods, the nunter and distribution of fuel rods containing gadolinia, mid the water-to-fuel ratio are similar in the two designs. Ilowever, two features which might effect the nuclear characteristics differ in the proposed 8 x 8 reload , assemblics and the equivalent 7 x 7 reload assemblies.
, Fit:st, there are 64 rods in the 8 x 8 assembly, compared to !
. 49 in the 7 x 7 assembly. Second, the 8 x 8 assembly has a l ,
water filled red near the center of the assembly end the 7 x 7 does not. . , The tr.ajor items of interest from the standpoint of nuclear design of the 8 x 8 reload fuel assembly are the uncontrolled and controlled (.all control rods in) reactivity, the change in reactivity of the assembly with burnup, the local peaking in the assem* ly,.the Doppler reactivity coefficient, the delayed neutron fraction, and th'e void reactivity coefficient. Values of these parameters as a function of burnup for an infinite lattice of 8 x 8 reload assemblies were presentegl,2,3,4,0
. and compared with values for an infinite lattice of 7 x 7 4
9
D TABLE II t . Nuclear Design Comparison 8x8 7x7 Pellet Outside Diancter, in 0.416 0.487 Rod Outside Diancter,_in. 0.493 0.563 -l Rod-to-Rod Pitch, in. 0.640 ' O.738 Uater-to-Fuel Ratio 2.60 2.43 U Bundle Ucight, lb t, , 404.6 427.8 Cladding Thichnce: .7e _1. 34 32 r,,, cold uncontrolled 1.166 1.163 k , cold-controlled 0.981 0.988 , Max. Local Peaking Factor 1.22 1.24 . Average U-235 content, % 2.62 2.63 ,
, , Number Gadolinia containing pins
,4 4 Relative gcdolinia content of
.gadolinia containing pins 2 1
[ , Number of water rods 1 0 2.59 u/o U-235 8 8 Assembiv: 2.50 v/o U-235 7 x 7 Assembly
. 8x8 7x7 Pelict Outsfde Dfaucter, in. 0.416 0.477 Rod outcide Diameter, in. 0.493 0.563 Rod-to-Rod ritch, in. 0.640 0.738 Witer to Fuel Ratio 2.60 2.53 U Eundle Ucinht, lbc. 404.6 412.8 Cladding Thickness,. mils. 34 37 kg , cold uncontrolled 1.148 1.129 R g, cold centro 11ed 0.966 0.960 Max. Locck Peaking Factor 1.22 1.30 Average U-235 content, % 2.50 2.50 Number gado11nfa containing pins 4 4 Relative gadolinia Content of Gadolinia containing pins 1 1 Number of uatur rods 1 0 -(
l D . ( asserblies of similar enrichment. In general, the values for the 8 x 8 lattice differed by less than 10% from those of the 7 x 7 lattice. .
, The same calculational techniques were usdd in calculating the lattice paramaters for the 8 x 8 reload assemblies and those equivalent 7 x 7 assemblics. The particulars of the I
design of the ussembly do not directly enter reactor calcu.- lations since homogenized parameters for.the assembly (e.g., - few group cross-sections, diffusion coefficients) are used a .as input. The 8 x 8 reload asscmblies are neutronically similar to the 7 x 7 assemblies (i.e., similar enrichment, i water-to-fuel ratio and gadolinia content), and we believe i the calculations 1 techniques are of equivalent accuracy for an 8 x 8 cssembly as for a 7 x 7 assembly. The local peaking factor for the 8 x 8 reload assemblies is reported to decrease monotonically with exposure, while that of the equivalent 7 x 7 asseinblies is reported to decresse with an exposure of about 10 Ci!D/t, then increase sicwly. This behavior was explained, in response to a staff question, in terms of l differences in the shift in the position of the peak local
, puser rod within the bundle as a function of exposure.
l l i
~.
)
p.- i2 * { i
.The effect of the water rod .is to increase n:oderation in {
i the interior of the bundle and reduce the rod to rod power ' peaking. Voiding of the water rod would decrease the reactivity of the bundle and would depress the flux in the center of the bundle. (Voiding of the water rod is equiva-lent to increasing the void fraction in the assembly of about1%). l We have reviewed the nuclear design of the 8 x 8 reload fuel assemblies by comparing their properties with equivalent
" {
7 5 7 assemblies and conclude that the nuclear design of the 8 x 8 reloed asenblie: is cccepicble. I { ._ t
D ' ( 4.0 ' Thermal-Hydrculic Desien During normal operation and abnormal operational transients, the design objective for both types of assembly is to main-tain clad integrity and prevent the rel, ease of significant
. amounts of radioactivity. The fuel damage limits and thermal-hydraulic criteria used to evaluate the performance of the fuel is the same for both designs. During normal steady state operation the Minimum Critical Heat Flux Ratio, (MCHfR)isheldabove1.9. For abnormal operational transients, -
. the clad strain is limited to less than 1% and the MCHFR is a maintained greater than 1.0. THese design bases are the same as the dcsign bases for fuel previously reviewed and l'
! acccpted for boiling water reactors.
i In general, the 8 x 8 fuel has. greater thermal margins to these design limits than'7 x 7 fuel. The design value of linear heat generation rate for nor. mal operation is 13.4 kw/ft for an 8 'x 8 fuel and 17.5 to 18.5 kw/ft for 7 x 7 fuel. Based on' previous experience, this lower thermal duty combined with the other design changes is expected to result in fewer clad perforations. During norcal operation, the hot channel MCHFR in the 8 x 8 assemblies is expected i e a 9
D , I to be greater than 2.3 which is 11% greater than the hot channel MCHFR expected for 7 x 7 assemblies. The LHGR which is calculated to produce 1% strain in the cladding , is 1.8 times the design value for 8 P. b fJel and only 1.5 times the design value for 7 x 7 fuel. 'Similarly, the LHGR which produces fuel pellet center-line molting is 1.4 times the design value for 8 x0 fuel as compared to 1.2 times the design value for 7 x 7 fuel. . Since the 8 x 8 assemblics are different than the 7 x 7 a .. .. : assemblies, we reviewed the thermal-hydraulic design methods to doterminn their applicability to the new fuel design. The
, differences are the modified flow geometry and the introduction of an unfueled rod. The portions of the thermal-hydraulic 4
design methods which might be affected by these differences and which we reviewed are the techniques used to calculate flow rate and critical heat flux in the 8 x 8 assemblies. I l The methods used to calculate flow and pressure drop in the 8 x 8 assemblies are the same as that used for the 7 x 7 ; assemblies. However, empirical constants are varied to adjust
' the results to the specific fuel design. Tests have been rade to determine these empirical constants for an 8 x 8 geometry and to confirm the method of calculating friction, i i 1 1
l p ' 1 (. . l acceleration and elevation pressure drop. Furthermore. the fuel assembly support casting orifice is the major 1 fica resistance and,therefore, the flow distribution be- - i tween fuel assemblics is insensitive 't$ differences in the hydraulic characteristics of the fuel assemb1ies. The methods of hydraulic analyses are the same as those previously reviewed and accepted for boiling water reactors and are equally applicable for 8 x 8 fuel assen61ies. . a The correlation used to cciculate the critical heat flux
. . i in the 8 x 8 assemblies'is the same Hench-Levy correlation used in eval!'ation of 7 x 7 a.*,scrblic';. Introduced in 1965, I~
the llench l.evy correlation has been the accepted basis for I determining thermal margin 1or a variety of General Electric boiling water reactors. The 8 x 8 fuel assembly is, except for the inclusion of an unheated rod and the change in hy-draulic diamater, very similar in gecmatrical and thermal-hydraulic chcrecteristics to the 7 x 7 fuel assembly.
!!e have previously reviewed (12) the effect of an unheated rod and the applicability of a CHF correlation such as the Hench-Levy correlation which' is based on average fluid e muunna
D - I
' conditions 'and concluded that the effect of the unheated rod is not significant. We have also reviewed the effect that the changes in subchannel hydraulic diameters might have on thermal performance and conclude that the subchannel flow in the 8 x 8 assembly is more balanced than in the 7 x 7 design and should result in improved thermal performance.
Therefore, we conclude that the Hench-Levy correlation is equally appliccble to both the 8 x 8 and the 7 x 7 assemblies. . Because the Hench-Levy correlation does not specifically a . . : account for non-unifcrm axial heat flux distributions and p rod to rod variations in power, et exist in fuel assemblie:, e f
. a lower limit line to the then existing critical heat flux !
t
- data was chosen as the form of the correlation. In addition, for added conservatism, .the steady state design CHF was to be such that it did not exceed the Hench-Levy CHF divided by 1.9.
In order to overcome these shortcomings of the Hench-Levy correlation and to provide a data base that is more repre-sentative of actual fuel assembly performance, General Electric constructed the ATLAS facility which has the capability to test full size, full power 8 x 8 rod bundles. Except for the
i D 17 - - ( I inethod of beating the rods (electrical resistance heating)- j 3 and differences in grid spacer design, the 8 x 8 rod bundles tested in the ATLAS loop are similar to fuel assemblies. , ; The large body of critical heat flux da'ta obtained from the ATLAS facility for both 7 x 7 cnd 8 x 8 of rods in 16, 49, and 64 rod bundles has provided the foundation'for developing a ned correlation called GEXL (General Electric Critical Quality Xc- Boiling Length) which GE proposed as a replace-ment for the Hench-Levy correlation. A new thermal design
, me,thod (CETA3, Gcneral Electric , Thermal Analysis Basis),
. which uses GEXL cnd appropriate design parameters to determine
{ the maximm po t r capt%y of a fuel cssembly during norm?1 operction and cbnormal operational transients and i . accident conditions, is also proposed. The Regulatcry staff is now revic. ting GEXL, GETA3, the Hench-Levy correlation, end the ATLAS rod bundic data. ! Generci Electric.has informed the Re:ulatory staff that all , operatinr; E'..'n pients have bcen provided with GETAD with the ! instructions that, in the interir.:, opcreting' thermal limits Le datermincd by either the Hench-Levy correlation or GETAB, choosing the method that provides the more conservative result. 1 i
D 18 - ( At this time the staff agrecs that the operating plant thermal margins should be predicted on the basis of the method (i.e., either Hench-Levy or GETAB) which yields the mare conservative result, on this'b$ sis, use of the Hench-Levy correlation for..the 8 x 8 fuel design would be accepts.ble. - 4 e O d * . .
?
l- , E
, e ..
D . d I - 5.0 Abnormal Doerational Transients To assure the safety of the plant, the results of the. aralyses of abnormal operational transients are required to indicate that the fuel and the reactor' coo.lant pressure boundary (RCPB)arenotdamaged. The fuel damage criteria are a minimum critical heat flux ratio (MCHFR) of unity and a cle,dding strain of one percent. The RCPB damage' criteria i is' the system design pressure (as specified in the ASME Boiler and Pressure Vessel Code, Section III). Thes,e damage a limits for 8 x 8 fuel are the same as previously reviewed and accepted for 7 x 7 fuel in boiling water reactors. i I' ' i Abnormal operational transients are the result of single 1 equipn.ent failures or sing 1'e operator errors that can . 1 reasonably be expected to occur during anticipated modes of station operation. The types of failures and errors considered are the same for both types of fuel. The transients resulting l from these failures and errors can cause variations in both . system parameters such as core flow, core power, pressure and coolant level, and in local parameters such as flow and power in a single assembly. System parameters are primarily a function of the core average nuclear, thermal and hydraulic characteris tics. O
D ' Since the characteristics of the 8 x 8 assemblics are similar to those of the 7 x 7 asserblies, the 8 x 8 fuel has no significant effect on these transients. However,
. . for the determination of local parameters, the characteristics of the 8 x 0 faal may be significant. It has been reported (l) that the thornal margin of the hot assembly has been angkyzed using the conservative fuel type and the results demor. strate that the fuel damage limits are not exceeded.
The result: of three limiting events, i.e., a seizure of one recirculot.cn pump, the ccntinuocs withdrawal of a control
- rod, and the miserientation of an assembly indicate that I'
( the consequcnc2s of these events are less sevure for 8 x 8 i assemb' lies thin fcr 7 x 7 assemblies. Analyses of all , i transients hase been mado(3) considering both the 7 x 7 and 8 x 8 asser.Slies and the results indicate that the fuel damage limits are not exceeded. G t e 6 I 1 I
D , . ( . . 6.0 Accidents. Analyses of the design basis accidents are trade to evaluate the capability of the engineered safety features to mitigate : the consequences of postulated accidents and control the possible escape of fission products. The four postulated design basis accidents are the a) loss-of-coolant b) steam li,ne break c) fuel handling and d) control rod drop accidents. 6.1 Rod Drop /.ccident The rod drop accident cnalysis is not significantly.affected a b,y a change from a 7 x 7 to ai g x 8 assembly. The kinetics mode'l uses homogenized cross sections and is not directly I involved with the details of the lettices. The local peaking i factors of interest are also similar for both types of assemblits. Analyses of the rod drop accident demonstrate that'the dropping of a maximum worth sequenced control rod will not result in a peak fuel railet enthalpy which exceeds the d: mage limit of 280 cal /gm. 4 0.2 Refdeling Accident
- The method of determining the num5cr of rods which might fail follo. ting the dropping of an assembly is equally appliceble to both designs. Since the types of assembly are similar, the total amount of fission products released
i D - from the B x 8 assemblies in a refueling accident would not be significantly greater than from the 7 x 7 assen61ies. J 6.3 _Stec n Line Break The radiological consequences of a postulated steam line break cutside of the pritrary containment are dependent on the cmount of princry coolant lost during the accident and the cenccntratinn cf the radioactivity in the coolant. The amount cf coolcnt lost is prirc.hrily a function of system
, par (=eters which would not be significantly changed by introduction of 8 x 8 fuel assenblies. The concentration j of radi'r M tvity .r. the coulent i.; iinited by Technicai Specificatiore and is also unc, hanged. Therefore, the i
rcdiolo;'ical conscquencc.s of a postuleted steam line break
. accident are unchcng:d by the use of 8 x 8 fuel essea.blies.
6.4 Loss of Cenlent The cnni/ sis 0." the pcrforr.ance of the ECCS and the response
~
of the 8 x 8 fuel assenblie: folicwing postulcted less-of-coolent eccidcnts has been made using the cssumptiens and cciculatienti techniques described in "Part 2 - Gencrcl Electric Evelection !lcdei, Appendix A Acceptable Evaluation . t 0
D .
' Models including their Conservative Assumptions and Procedurcs" which is contained in the Commission's Interim Policy Statement, entitled " Criteria for Emergency Core Cooling Systems for Light-Water-Power Reactors" and published in the Federal Register en June 29, 1971. The Commission Rule " Acceptable Criteria for Emergency Core Cooling Systems for Light-llater-Cooled Nuclear Power Reactors" dated December 28, 1973, is intended to replace the Interim Policy ;
Statement. Conformancc with this new rule, which includes 4 revised criteria and revised features of the evaluation model, I will rcquire re-enalysis of the ECCS performance. When the l ny n i e, 'i t , c. .d na t iv,r, one w. billed to i.he Direeiur of Regulatinn, as required by the implem:ntatica schedule contained 1 in the rule, the sicff will make its review and conclusions. Our currcnt review is only concerned with compliance with the Interim Policy Statement. Since the 8 x 8 fuel assen61ies are. a dif ferent design than the 7 x 7 asse.ablies considertd in the General Electric Evalu> tion Model described inljEDD-10329,andreferencedinPart2ofAppendixAtothe'
. i Interim Policy Statement, the staff has reviewed the evaluaticn '
model to determine its applicability to the new fuel design. t The features of the ncw fuel design which are different from the old design and significant in determining applicability
- i I
D . 31 - t , of the e/aluation model are: a) srn11er diameter fuel rods; b) larger nunber of. fuel rods in each assembly, and c) an unfueled central rod. The features of the evaluation model ! which might be affected by these changes in the design of the fuel assembly and which we reviewed include applicability i of the transient critical-heat flux correlation, the thermal ; radiation and the spray' cooling convective heat transfer in an 8 x 8 array, and the effect of the unfueled rod on heat transfer. As discussed in a preceding sec, tion of this report, we have
. revicwed the differenc's e in the thermal and hydraulic charac-l l' tcri: tic: c;t""en an S x 0 fuei assembly and the 7 r, 7 assetbiy,
! I ;- , cnd concluded that the* steady state critical hcat flux l correlation is equally applicable to both designs. In 1 I addition, GE has nearly ecmpleted an extensive series of stecdy-state critical heat flux tests on full-scale, 8'x 8 heater bundles. with varying inlet conditions, and power distributions which are representative of expected conditions in a Bh'R. These tests will provide a large additional set of critical heat flux data applicable to the 8 x 8 fuel design. General Electric and the staff are now in the process of evaluating this data and its applicability to the ccnditions 9
D-2$ *' i following a loss-of-coolant accident. Upon the completion ! of this evaluation and during the review of the re-analysis required by the new rule, the staff will re-examine the acceptability of the current critical heat flux model. He have also revieaed the differences in thermal radiation and spray cooling characteristics between the 8 x 8 and the 7 x 7 fuel assemblies and conclude that the procedures used to calculate the hntup of en 8 x 8 fuel assently following a loss-of-coolant accident are consistent with _the cpproved Gencral Electric Evaluatien Model. Our conclusion is based on independent cciculations using a computer program developed , l' for the staff (1 .0/ cnd the results of full-scale, stainless ( i steel S x C red uray, h: ,ter bundle sprey cocling and flooding tests. (IbIN The edequacy of the t!.crm a l radiation model for an 8 x 8 fuel bundic has l;ecn verified by compcrisen of the predicitons of cled tc;:pdrature using both the GE(IN and staff's(IN computer progrcms to the results of steady-state heater bundle I tests which had not spray cooling. The s taff's computer e
D ( i i program underpredic'ts the temperature of rods in the ; bundle by not more than 25*F, but overpredicted the t-
, temperature of some rods by as much as ,150 F. The GE l program predicted temperatures which were from 50 to 75'F lower thtn the staff's calculations. The temperature overprediction of the corner and unfueled rods may be due to local differences in emissivity. Although comparison of the gray body view facters for individual rods used in the two progrcas reveale'd no rea:.on for the difrerence be-tv/cen th'e GE and staff results, the simpler nodalization of the hester rods in the GE progran could account for the I' '
difference. t The adequacy of both the GE and staff heatup models, including both convective cooling to the spray and rod-to-rod radiation, was demonstrated by comparing predictions t? the results from transient tests of the 8 x 8 stainless stcol heater bundle. The predicitt.ns were bast =d in part on the conservative values of spray cooling convcative heat transfer coefficient specifed in the I/I evaluation model. The other par 5rceters, such as ! heat-generation, emissivity and thermal properties, were best estimate values. lhe staff's calculations are as much as 40'F loaer, ( i
D - 27 - . ( and as much as 80'f higher than the measured ten'peratures. The predictions reported by GE have approxirrately the same j inaccurncy. These differences are within the uncertainties ! 1 of the test results. The General Electric Company has also completed a ' test witnessed by the staff on an 8 x 8 Zircaloy heater bundle, but hcs not yet repried the results. Previous tests have sho.;n that a hcatup modci which is bcsec' on the results of
.tbsts with stainlets steel rodi chn predict the thermal !
response of Zirccloy rods within the uncertainty of the ( > cxperitantal mccse.~erc.er ts. For t.:.st recc.tcrs which have
; jet ;'rps, the heabp transients are short, that is, approx-irr.ately t.to rc.inutes long, result in modercle temperatures ,
that is belce! 2000'F, and the degree of uriccrtainty is acceptably
. sr.m il . }l0.tever, fo?'tN0cidr.ts thich cre lo ,ger and result in higher tergerature:, such as occur in reactors without jet pumps, tdditisnal experimental verification of the applicability of cnalytical methcds derived fran stainitss steel heater bundle tests to Zircaloy clad rods are required. Therefore, , the rctults of this Zircoloy bundle test will be subuitted and ravit.:ed prior to use of fuel in retctors without jet pumps.
I
D t'
!!e revievied the effect that the unfueled rod might have on heat transfar. Inspection of the test results indicate the convective cooling of the rods prior to wetting of the unfueled tod is insensitive to the lo'c ation of a rod relative to the unic: led rod. That is, rods immediately adjacent to the unfueled rod heat up at the same rate as' rods which are septratcd from the unfueled red by one ro. of rods, llo tever, tii e unfueled red is beneficial si ..e af ter the rod weli, it acts es . therrt.31 radiation sink. Wetting of the a uniucled red ir, not included in .pither the GE or staff computer .
progran rcdois. I, l!c conclude th:.t thc. General Eler.tric Ev:1uction Model as describqd i in IEDJ-10'D cad including .the requiic:rints specified in Part 2 of A;;;;rdix A of the Interir: Policy State :nt when radified a CMcrib; din NEDE-10SJ1 to account fer differenc:s beaccon the du.igr, of the 8 x 0 and 7 x ? assemblics, is ep;*licrbi:: to the cvaluatic.n cf the ELC ptrierr4nce of 8 x 8 asremblies in a General Electric boiling water reactor which his j et p t T:_ :. . t J e g**r* A
D ' 7.0 References - f
- 1. "D'resden 3 I;ucicar Pouer Station, Second Reload License Submittal,"
' General Electric Co., J!ucicar Fuel Department, September 1973, an'd Supplement A November 27, 1973; Supplement B, December 6,1973;-
Supplement C, Dece:6er 6.1973; Supplement D, December 17, 1973; Suppicment E, December 17, 1973 (Proprietary); Supplement F, Jan-uary 9,'1974; Supr.ement G, January 9,19.74 (Proprietary); Supple-ment H, Janue.ry 23, 1974. l
- 2. "fline liile Point Unit 1 - Second Refueling," P. D. Raymond to A.
Gianbusso, September 14, 1973. 1 "fline liile Point Unit 1 Safety Anclysis for Type 5 and Type 6 Reload Fuel," Niagara l'.ohnik Pouer Corporation, October 15, 1973.
"liine liile Point Unit 1, Part 1, Hon-Proprietary Response and Part 2,-
Proprietary Response, January 15, 1974. j "lline Mile Point Unit 1, Analyses and Proposed Technical Specification Changes, Jcnuary 22, 1974. a . ( 3. "Iionticello fluclear Generating Plant, Permanent Plant Changes to Accorcodate Equilibrium Core Scram Reactivity Insertion Characteristics," Janutry 23. 1C74.
- l. ! ,
- 4. " Pilgrim Cycic-2 Lic:nsing Submittal," M. J. Feldman to J. F. O'Lecry.
i January 24, 197J.. . i
- 5. HEDD-201D3. "Goneral Design Information for General Electric Boiling )
1 Haler Reactor Relocd Fuel Cwsencing in Sprin9, '74," September 1973.
];
- 6. 11. E. Willitmson and D. C. Ditrocre, " Experience with BWR Fuel Through !
September 1971," l'ED0-10505, thy 1972. l
- 7. GEAP-4059, " Vibration of Fuel Rods in Parallel Flow," E. P. Quinn, i July 1962.
- 8. Letter J. A. Hinds to V. Moore, February 4, 1974.
- 9. !!EDM-10735 "Densification Considerations in BWR Fuel Design and Per '
formance," D. C. Ditmore and R. B. Elkins, December 1972, Supplement 2, " Response to AEC Questions, flEl'i-10735 April 1973 (Proprietary),
' Supplement 2, " Response to AEC Quastions, NEDM-10735 Supplement 1,"
May 1973 (Proprietary), Supplement 3, " Response to AEC Questions, NEDM-10735, Supplemcnt 1, June 1973 (Proprietary), Supplement 4, l
" Response to AEC Quastions, f;ED '0735," July 1973, (Proprietary), l Supplement 5, "Densification Considerations in BWR Fuel," July 1973 i (Proprietary), Supplements 6, 7, and 8, " Fuel Densification Effects l
'g on General Electric Boiling Water Reactur Fuel." l l
l L______ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _________ ____ _ _ _ _ _ _ _ _ _ _ _
D '
- 10. fief;0-20181, Supplement 1, December 3,1973 (Proprietary); GEGAP-Il!,
i "A P.c??1 fcr the Prediction of Pellet Conductance in BilR Fuel Rods."
- 11. "Technicc1 Renort on Densification of General Electric Reactor' Fu'1s, Deceri.ber 14, 1973. -
- 12. " Change fio.17 for Oyster Creek, Docket I?o. 50-219, License DPR-16" Letter from D. Skovholt to Ivan Finfrock, Jersey Central Poi Er Co.,
dated liove.ccr 16, 1973,
- 13. " Sensitivity Study en B'..'R/G Fuci Bundle Response to a Postulated LO^'.
C. f1. I'.cscr 2nd R. W. Griebe, December 1973,
- 14. I,'Eid:-10801, "l'odeling t!?e DPR/G Loss-of-Coolant Accident" Core Spray and Detto a flooding l!:ct Transfer Effectiveness," J. D. Duncan and J. E. Lconard, iiirch 197:., and " Response to AEC Request for Additionr.
Infonc tion cn liEDE-10^01," licy 1973. (Proi,riethry)
- 15. I EDD-1 dei'5, " Core Sprcy and Cottom Fleeding Effectiveness in the Bt."'/
J. D. Duncan and J. E. Leonard, September 1973. - a 16. '.' Core : Therm,1 Analyses of a Stc.-:nless Steel Cind Hecter Rod Dundle," C. l.i. li:cer ar.d R.11. CJiebe, December 1973. l ; i . (- ' l 1 l
,' D '
CI'.10' 0LO3Y - e REGUL ATORYd R' V!bl OF G'"ERI.L ELECTRIC 00",PA';Y GXC FUEL ASSEMBLY. September,1973 Gcneral Electric Corrpeny submits report " General Design Information for General Electric Boiling Water Reactor Reload Fuel Ccorencing in Spring, '74," NED0-20103. Septcraber,1973 General Electric Cerapcny,'ffeclear Fuci Department, sub-taits report "Dresdtn 3 Nuclear Power Station, Second Reload License Submittal ." September 14, 1973 Man to A. Giambusso, AEC from P. D. Raymond, "Nirye Mile Point Unit 1 - Second Refueling." l Octobtr 15, 1D73 Itingarc l'.chn:h Po. cr Corporation submits report "Nine Mile Point Unit i Safe,ty Analysis for Type 5 cnd Typa G Reload Fuel ."
- 04. t n.w u, nu . ru to V. e.n l iv , tJ C , r m D Ru3s ond'T. Nuvak,
' "Rcvicw of GE CX2 Reloud fuel Assc:ablies."
I' OctcLcr 24, 1973 l'.:r20 to V. Stello, AEC, frora U. Minners, AEC, ' Review of GE CT.B Re10ad Fuel Ascen'0 ins." Hove:chr,1973 Mardern Stctos Po.:er Ceg .ny submits report "Monticello Nuclecr Ceneratina Plan; - Seccnd Reload Sub:aittal."
!!ov:*r 16,1973 Letter fron D. S!:ovholt to Ivan Finfrock, Jersey Central Power Cc :.:.3ny, " Change l'n. 17 foe Oyster Creek, Docket No. 50-219, License DPR-iC."
November 17, 1973 Gerieril Electric Company, Nuclear Fuel Department submits Supple:acnt A, "Drer. den 3 Maclear Pmier Station, Second ! Reload Liccnto Submittal." , i Deccmber,1973 Energy Incorporated submits " Sensitivity Study on BWR/6 '
.f'Uel Bundle Response to a Postulated LOCA," Part IV, C. M. Moser and R. M. Griebe.
( December 6, 1973 General Electric Company, Nuclear Fuel Department submits
1 y . . l- . D
- i k
I Supplement B, "Dresden 3 Nuclear Power Station. Second Reload License Submittal ." 1 \ l December 6, 1973 General Electric Company, Nuclear Fuel Departmant submits , Supplement C, "Dresden 3 Huclear Power Station, Second i Reload License Submittal." . j Decemb r 6, 1973 Letter to J. O' Leary, AEC, from J. Abel, Commonwealth i l l Edison, " Supplement B to Second Reload License Submittal." ' December 6,1973 Letter to J. O'Lecry, AEC, from J. Abel, Commonwealth . Edison, "Suoplccent C to Second Reload License Subaittal and Proposed Change to Facility Operating License DPR-25." December 14, 1973 Memo to V. Stello, AEC, from W. Minners, kEC, " General a Electric 8X8 Reload F001 Assemblies." Deccaber 17, 1973 Cencral Electric Company, Nucicar Fuel Depari.rncnl. submits l Supplement D, "Dresden 3 Nuclear Power Station, Second Reload License Submittal." Deceuber 17, 1973 Generci Electric Company, Nuclear Fuel Department submits Supplement E, "Dresden 3 Nuclear Power Station, Second Reload License Submittal." December 17, 1973 Letter to D. Ziemann, AEC, from J. Abel, Commonwealth Edis6n, "Suppicment D to the Second Reload License Submittci' December 17, 1973 Letter to D. Ziemann, AEC, from J. Abel, Commonwealth Edison, " Supplement E to the Second Reload License Submittal December 1D,1973 ACRS mec ing on GETAB and applications to LOCA analyses for
, 8).o assemblies. ,
January 8,1974 ACRS Subcortmittee on Fuels Meeting, Washington, D.C. Janua ry 10, 1974 .ACRS Meeting, Washington D. C.
-(
January 24, 1974 ACRS Subcommittee on Fuels Meeting, Denver, Colorado.
D . i January 3], 1974 AEC - General Electric i',20 ting, 4 February 5,1974 Letter fro:n J. A. Flinds to V. Moore.
/* el 4
4 o O g 9 4 5 ' 1 I i < l a
. (
I 1 l l f
, APPENDIX E -
1 1
/
REVIEl! A'iD EVALUATION OF GETAB
'(Ceneral Electric Then IlAnalysisBasis)
FOR EURs l: < i By Technical Revieu Directorate of Licensing United St6tes Atomic Energy Concissio.i September 19N i t e
i I - l 1 1 l' E-1 i ( i
. i TABLE OF C0f; TENT._S_
i
- 1. liiT RCDUCT 10.
- 2. ANALYTICAL -]
A. Criticsl Heat Flux Correlations a
- a. ' Methods i '
l b. GEXL
- c. Data Basis for GEXL B. Subchannel Analysis flethod '
l C. Data Co'eparison .
, a. Co.rutrison of ATLAS Data with Hench-Levy Correlation
- b. Comparison of Hench-Levy and GEXL for Rod Bundle Power ,
- c. Comparison of ATLAS Data with GEXL {
fI; I p, .Cyt.lu..iun .
- 3. EXPEf;IIiD;TAL ;
i
- A. The ATLAS Heat Transfer Facility !
- a. The Loop
- b. Ter,t Sections 3' C. IC S t I'r C C Ed < ;*LE:
B. Com.ourir.c of ATL/,$ Data viith Coluiabia University Data C. Evaluation
- 4. GETAB APPLICh 10.'! .
- 5. stair FlND1?G", A!!D C0h'CLU510NS REFEIE!lCES
i E 't
- 1. lHlP.0DUCTIO::
During anticipated abnormal operating transients in a boiling water reactor, a criterica of no fuel rod damage is applied. Historic 411y, the therrzlhydrculic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. /.lthough it is recognized that a departure from j nucleate boiling would not necessarily result in damage to BWR fuel rods, l the critical hect. flux at which beiling transition is calculated to o: cur has been ado ned 5s a convenicnt limit. Since 1906 this lin:it for BWR fuel asse.:blies has been based on the [ Hchch-Levy correlation) which was formulated as a lower limit line to I the existing rod bundle critical heat flux data. To allow sufficient margin
! I fo r ta '. ' i n . i c s , t l - s ts cy s ten- ci+, e i.in,. ud i ; .on , o f Gener.;i g Elt : tric LL pr.ny rento:'3 ucre liuited such that during coticipated abnormal t rr r.3 : e n t : iN calcul:ted h1.c t flux was alays ins than the lower limit criticil !
flux lin . Ihat is, during trar.sicnts, the critical heat flux rat.10 U . i! Licy a gr. rte th;n unity. Pese:' an recent exten. ive c:i: !cci he.st flux date obtained with full size, f ull pu..:r rod Lundies in t!.e liTLI.S test loop, the Ger. oral Electric Company hcs deve'.npad a nc.! rsethcd of critical heat fic., correlation. llith this Ecneral Electric Critical Quality (X )- c Boiling Length Correlation, (GEXL)(2') critical panr, the fuel assembly po..er at which boiling transitic is c: spec t.ed to occur, is bcned on the correlation of the critical quality and toiling length. Tho t:asic form of GEXL is identical to the
E F well verified CISE (italy) correlation. In contrast to the llench-Levy correlation, which is a lower limit.line to the data, the GEXL correlation is a best fit to the ATLAS data. GE proposes to determine thermal limits using a ne', thermal design method, the General Electric Thermal Analysis
, Basis (GEiAC) ,3) which incorporates the GEXL correlation. The uncertainties associat;d with the CEXL correlation and the reactor steady state operating para!nuter, are con 6ined statistically. The steady state operating cond;tions ere to be limited such that during anticipated abnormal transients, more then 99.L4 of the fuel rods in the core are expccted not to experience bailing transition. That is, during transients, the minimum critical p6te;r ratio'is to ba greater than a value determined by the magnitude of the u untm taintie:. A typical value"is 1.05.
I'
- Ti.i., roi,crt presents th; results cf cue revie v of the CEXL correlation
, and the G: TAC methnc'. We have review d the GEXL correlation and its basis including the expcri:Lental data and analytical methods used to deternine the correlation. Ile have also revicecd the experir.: ental methods used to obtain the data, including the design cr d operction of the ATLAS test loop. Finally we reviewed the application of the correlation to the design and operation of boilirig w&ter reactors. As discussed at the end of this report, bared on this review we conclude that the GEXL correlation and a statistical application of the correlation, similar to the proposed GETAB method, are acceptable.
1 L
E l' ? , ( 2. AN/,LYTI Cf,L , A. Critical lleat Flux Correlations , L
- a. Methods, Severni rethods, as described by Recys, et al } have been proposed for thc predication Of hect input for, and the position of, critical heat flux .in non-unifurmly heated tu'.scs. These prediction methods have the
- follo. ting fca tures
(a) In the "Acertge Hea t Flux" concept the critical heat flux is esv u :d to be correlate.' thus: (L. ' f (G, D, P, LH. in
, L), -
(1) wherc 't is ti.e total heated length, ' tili in is the inlet subcooling; G is mass ve t itj, 0 is the equivalent diameter, P is the pressure, I and < u_ i s ti.- a s :rr.ge ha.t fl u '. Insp:ction of these parameters
, S F e.! tact th. criticc: pr..c- is assend to be independent of the i
form of tnc h:at flux spatili distribution. Iithough this method j l does nct ; ermit the prediction of the critice' haat flux location, it is simple r.nd h;s bccr shnta, by Leo ), to give praic tions within 10, of c yperircental dat; for tubes with large L/D ratics ar.d madere te ru t-t.7-aver 63 heat flux f ern f actors. (b) In t% "LoulConditichs" concept the expression for critical heat flux is:
., <c * (' ' '
c (2) l This inthad assumes that tne critical hea t flux 10cctich will ! occur a t a local hea t flux, qc, and a local stcam quality, X , irrespective of the axial heat flux distribution. Exampics of l 4 l 1 i
E t I f this type of correlation are those of Thompson and Macbeth(6) , Tong (W-3)N) , and Gellerstedt, et al, (B&W-2)(8) . However, the need for a correction to the predicted uniform critical heat 1 flux, for the case of non-uniform axial h'est flux, is described in Reference 9 for the W-3 correlation (8) , and in Reference 10 i for the B&W-2 correlation (9} . The Hench-Levy correlation (l) , 1 for use in BUR rod bundles, is similar in form to equation 2. fc= f (G, P, XCD} (} t whare g is locci critical heat flux, and X CB is the bundle averegc critical steam quality. The' equivalent dian.eter does not annuar as the correlation is applicable only to GE BWR rod b bundles. ( Since 1906, the Hench-Levy corrclatica, in the form i of a icwor lirait line to the thcJ1 existing rod bur.dle data, has served es the.besis for predicting the tl.erm:1 margin in BURS. (c) In the "Goiling Length" conccpt the critical quality is correlated in the following form: x=f(G,P,D L. ) , C D (4) where cx is the steam quality at dryout conditions and L is ! 3 the length over which boiling tai.es place. Exampics of this type of correlation are those of Certoletti, et al I) ,and
' }
Hewitt . The "beilirig length" type c.f correlation has the dent;n:,trated advantage of being cble to correlatc critical heat i flux data for both uniform axial heat flux as well as non-uniform ( ___.m___._..
E a axial heat flux. Since the axial (and radial) heat flux distribution in a BWP. fuel bundle is not uniform, the corre-lation of DerLcletti et al(II) was chosen as the basis for the new . GE correlation called GEXL (General Electric Cr.itical Quality X - c Boiling Length). /h used for GE rod bundles, GEXL relates the bundle average critical quality, XCP>, to boiling length. A recent corparison of the correlations described above L (excluding GEXL), and adaptations of some of the methods to use with rod bundles, to CWP rod bundle critical heat flux dcta is described
/
by Guarino, et al.'1'3) The compared data comprised 785 points and included
' ., i uniform heat flux, radially non-uniform heat flux, and axially non-uniTo,y; heat flux.
- b. GQJL.
i - ine GEXL cor: oletion is a variation of the critical quality vs boilin) length correlation of Certeletti, et al(II} which was based t on singir_ tube d.:i.1, Lat ucs sha.!n to apply, with good success, to a { 1arge mm:nt or red hur.dle critical hect flux data.UI) Subsequently. tht s i: .il a r ( X , vi. L ,). c fa correlation of He titt, et ci,(14) for single ur.iforrily heated trbas, was crplied to rod bundles by Marinelli cnd Pastori("') co the bass of ascribing the flow rata attending ecch rod to that which c>isti. within a zero shear interface betwe:n rc:'s (the ClLi teiturion). I _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ _ ~
I E ' (
)
'The GD:1. correlation is of the fom:
x cc
= f (G, D, P, Lg, L, R) ,
where the tems are as previously defined, L is the heated length, and R is a weighting f actor which characi.orizes the local rod-to-rod i pea':ir:9 putern with rc':pect to the most liraiting rod. In addition, R is de;;endant on lattice dimer:3icas (7x7 or TayS) and grid spacer I configuration. Since It, in effect, 6ccountr. for the flow and enthalpy distribution ,tithia tb bunJ1e, it can be intcepreted as being the bundle average analog cf sul;hany.1 cnalysis. The forn of the CISE correl >dien(1'1 is: a .
.a L '
y wherec=f(P,G) .
^c ~ b4 L e b = f(P.G. D) whcre i n :s defined.cs the he:te'l length ou r which th stesm quality is a
grebtcr th;a rerc. The form of the GD:L cor rehtion is sitailar to that of cI;r; I,. ...er, in GEr.L, L is defiried cs the distarcr fren the g init.l? tion of bul!, Loiling to the boiling trcnLitic, pirt
- l'.n e> .ple of the s.iili ty c1 the criticcl qu;1it," - t oiling lenpR 'IV. c arrele tian,(II) which is the bc:i: for GE2, to bring critical hm.t data for various axici heat flux distributions into a single cur /c is sho.!n in Figare 1. These date, for 1000 pria stea",
from Keeys, et al,(") inclada the folicwing hect flux distributions:
~ ~ ~ ~
*Keey . . o t al(!i uw Ln as defined in Lg; as defineJ by I'erto'!ei.ti, et al(II)GEXL, . Ilou2Ver, this whereas dif#erence isIR.titt, not et al important es long as the particular definition is consistently applied.
l
E I uniform, exponential decrease, and syrnetrical chopped cosine. Figure 2, from Reference 2, the report under review, shows the identical ret ults for Freon-ll4 wherein critical heat flux data for uniforn, cosirie, half cosine, inlet peak, and outlet peak heat flux distributions as viell represented by a single critical quality-boiling length curve.
- c. Deta Pasir for GEXL The OTXL correlations (7x7, 3xS) are bc::ed on experin.cntal datn which cover the folle'iing rangar:
Fressure: 800 to 1400 p'ia . dass flux: 6 0.1.x 10 to 1.25 x.10 6 lb/hr sq. ft. Inlet Subcooling: 0 tc 100 Bte/lb h LocDI Pev.hing: g 1.6I corner to"iA7 interior s 1 I.xial Profile: Uni fo rr., . Cosine (1.39 r?ax. to avg. at 72 in. from inlet) Inlet Fesk (1,60 n;c.s. to avg. at 51 in. fron inlet) Outlet Peak (1.60 mex. to avg. et 93 in. fror) inlet) Double Hu;cp (1.40 and 1.38 max. to avg. at 51 and 100 in, from inlet) Lattice: 7x7 and 8x8 l Rod Dundle:. 16, 49, 64 rods Heated Lengths: 6 ft, 12 ft, 12-1/3 ft
c E I The uf > L corraletions are ba',ed on data of which the overwhelming portica t:Fre obtained in the ATLAS loop and the remainder in the
~
Colum';ia University test icop. The data used for GEXL are: lattig pp.ofP.eds A>;ial Profile Heated Length No. of Points 7x7 16 Uniform 6 ft 84 16 Cosine 12 ft 223 49 Cosine 12 ft 127 8x8 16 Cosine 12-1/3 ft 211 64 Cosine 12-1/3 ft 1058
*Cel t.rlii a !!ni'. arsi ty da ta.
.4 l Frc a tu a'r ove. it can be senn that the GEXL correlations (7x7 l i and fat) cr2 b:/ cd cn 15U3 data points of winch 1266 i ere obtained seith full s ac ('. in' e4 red:), f cii 1:.ngth (i? ft an ' 12-1/3 f t) rod buHl a . L cept ict C4 point .ich ::cre cbtained :.ith a uniforra axirl .5 e:t fi t.;, the d.t 1 cre sbi.:ined lith a cosine he:tt flux dis-o t r i b u t i :.!; .
Afar t.velr- oT the carr:la tions thcy were cer;.nred tu add ; t inr rl d,: c si;.ich represen t the 1: hole rc:.;;c of pe.rc. ?ters. The data und Tor com;;c ci:0;i at e: i.a t tice f.'o. ci !?ct Axial Frofih: !!a. of Inints* 7x7 16 Uni fore 455 15 Co.ine 121 49 Cosinc 470 16 Inlet Peak 484 ' t 1
i l ' E' I lattice Fo. or Roh Axial _ Profile No. of Points
- 16 Outlet Peak 477 49 Outlet Peak 32 16 Double Hump 434 0x8 16 Cosine 131
- Includes 220 points from the Freon loop Inspc<: tion of theic comparii. ens shows thet the 7x7 GEXL corre-1 lation, whid, was buu r' on unifon:. end cosine axial profiles, accurately pre': cts t:ie whole rang of da:.a. '
a W; tile i.h: formulation of the GEXL correlations (7x7 and Sxf;) rolje:1 very nr:cvily (120 cut of lfM3 points) on datt taken with the
] , 8xc icttice, only a small fraction (131 out of 2005 points) of the datr uscd to che: $ the GEXL c.'a rciatior.s it.Fe for t he 82 lattice. In addition, ab:;at half cf thne confim.'ng dad a f;r the C d lattice, were for Fre:,n u:0 ell wc're fcr 2. cosinc hont flux distribution. GE is performing addit icoal tut; with f.v. lattices and r.on-uni for., axisi profiler. The:c tilitie.1 tests vill include profiles vi:h a peck tomrd the cutis t. Ir.1, t Enhed profilts may also be included.
Although these tests can provide additional confirmation of the 8x3 Gl:XL currelatloa pr::dictive capability, they ar: not required for two reasons, first, the 7x7 G~XL correlation, which was based solely on data frcm uniform and cosine axial her.t flu;; profilcs tests, (
E t acu'ratel .' p. "'! ts boiling tiansiticn for the other tested profiles. There is no icason to believe that the Ex8 GEXL correlation would not perforn similarly. Second, in the application of GEXL, the standard i deviation of the uncertainty in the Ox8 GE"L correlation will be ' inct eawd to account fer the less complete data base. The standard deviati.a. Of 2700 experimental critical power ratios (ECPR) about the 7x7 GEXL correlation is 3.61:. The standard deviation of 1299 ECPR about th:: 8x n GT; XL i s P . C'.'. . In applying the 8x0 GEXL to the deter-ir,inc. tion of L',li' t.hcreal liraits, the stcr.dard deviation will be increased to at least 3./.J. which is the rqi are root of the stm: of the variance
-of the 8)# c; .crir.ntel rcruits er.d the 0ariance of the me'ans of the 7x7 d a l.: for each fl u:' shn,:e.
l , i I l l 1 1 l l i i l
E I B. SuFrh;nrol T..st . ir :'ethod The subch;n wi e,naly;is m:thods used to develop and cc plement the GEXL correlation br,*e been reviewed cnd evaluoted by our PNL consultant. l This evalur tion ir.dit: ries that the prirary General Electric subchannel i analysis r.c,Al ha a r. so.M !c hesis. The basic formulations and computa-t i ol. ; of the r. ic -' 1 c e ty% cal of subchannel analyscs available in the open literature; ho.:ver, for;mictioris of the exch:nge mechanistas between subch , ,:b coni:,fi: 50::.e unique featurer. The incit sien of the p rticular f: ..e : t F ci ti tur. .i c ' u.,'ing and void dri T;,e ch;.:,ce m'cdels is ene ; of the m: :. si' rii'i r r &c: : iE in the C'i iuS:htnrol c.nclysis fcnnulation arJ is the jp i.. .rr re_:ca thnt th GC inc.dpl dots a good jcb of predictino subchw .1 flou crt' enthcl.w d;ta for sir.:ulcticas of CUR rod bundles.
*r_.s . -. . _ .. .. . . . .
}' .y; ll u s , .J s u.
.p .. ,.,..,t* .
l
. e . e .c s e 4 }] s. e s L' , G e&L .r-C l . .% . l c ; ' i . : ,. L ' p : . .' C ;
- ! 2 'l i ' i t.M h Cort'Cl' le . C2 rod h :Y ' datG t
s 'i ~[.C?l p, tii # h , '; D .' . ' telC [ ',", C,Uhli ty C :,"l b0ili r.g l eD g th c'e :o . b r - :,. . : . c..m
- 1 an: lysis. It is itT:ct:nt t note that tub:
,.'g *',., ,,,e d
t' g** ( g* g g J r * .as **
- I ., e i e c L ,',- ). ,g a d sg g. .a' 4 ,l-
{.,,.aa. Ia
!,. . e s
. . . . 4,,. y. .,"3 g e og
. i s <
5 ,,, . ; ,a _s g .as,, ) ,. ,, g - g, = tinvs., e- 1' -. 3 r.. e. ,. .-.r; .Lc .,, -t +ds 4 .: I. ' i. t ., i A6J s-
. - ,: . . 1. . s. s
. . , t. .
- da, ;t. ......,..,.,4.c..,.-
s . vi ; 3 ECT ' ; J fi, l i i C l , I bOl!$C I- G',,i L i On d.7 ti' [O r rC b U r. ~.' ) C S . t P C. De cca pic su;' W 1.ulys; c'.i.;d c>rrelat';s thc 0: peric.wici data v.ithin 4,/ e ( *- - ' g ( I h . ..*.6* g I {
- 3 0 V6 ll',d 1. i b:M for 070 in pt. M.'e!riC d?si 49 SIUdiLE Of R-l fu'.l bundles.
INb y e bb 'NI . Cl ' l3( ' (s .) (b rrf; ). (' [ , y/ (' l[]N [j
, g the subcham,el en_.ly;is nath. J and the Lundle averty GUL tethoci. !
l ( l l
E ! However, it does not logically follow that since the t<co empirical methods provide comparable critical power results that any one of the
~
individual factors iri either of the correlations can be justified on that l' asis alone. Each f unction.must stand on its own nerits in conjunction with 1 the erpiricel or semi-empirical correlating schtree of which it is a part. Ccosequently, the arguments associated with the justification of the bundle ave? ags R factor via the subchannel critical power results cre not only cir celar in nature, since both critical power correlations are based en tb so: e rod bu :dle d;ia, but ar e unnecessary since the GEXL correlation stemds on it s c..n ruits as a rnethnd of predicting critical btnt:ile powar. a C. Deto.Cr oad soa e
- a. _Cgpvison of ATLAS Data lith Hench-l .wv Corre1 3 tion
! ; /, Ll.e llc. A Lcvy co eclation presen~ly iciws tnu casis 1or precia.ng
, the. val limits fcr BUP.s, this correlation was indmndently ccmpared, by cur AllC con:,ultcn t ,
to the rod bundle boiline transition data obtained in I tne ATL/.S t u t locp. The comparison, censisting of SP53 data points, shcued that, c:m pt for the ccse of uniforia axial end rediel heat flux distribution, the Hunch-L'avy corrcletion, which is a louer limit line, is not conservative.* That is, the experiroental critical heat fluxes are generally less than those predicted by the correlation. Figure 3, frca Reference 17, shows critical Fundle power as a func:ien of inlet subcoeling with mass velocity as a paraneter. The consistency shown by these d.'ta is typical of that obtained in the ATLAS test loop. 1
*fhis' argument as to " conservative" does not embrace the use of Hench-Levy l i.e. , the present requirernent that MCHFR using Hench-Levy be greater than 1.9 dur inq steady s ta te operation.
( I I i l
I
~ l 1
l 1 1, . E 1 l A t [ V:'.ri; e' . ..ra v.carcd c.nd pr edictc.J critical hcct flux (ilencklevy) Ter 1:niforn axial and radial hcot flux distribution, is shown
]
in figu: e 4.(171 ! It can be seen that al':.ost all of the measured heat i l fluxes ei n cr: ;r thun thc predictions tl.as sh0 Wing that, for these j conditian., the hench-Levy correlation is conservative. This result is I not tc urpris:rg or the e;.perim. intal basis for establishing the Henen-Lev; ccrrelni co e.as primarily comprised of uniform axial and radial heat j flu, rod turdle d:ta. 1:cc.cve: . as sho.;n in the folimiing two figures , the gr: ;ter t.!;e dupi-h "e irca unifor, c::ial end rcdici heat flux distribution, l th6 c.v ':: L:c-en ' the dispr: 0 i.vtween the Hench-Levy correlatica and i "mysur'. : .cri t ic::1 t.:a t fluxer. (9 th Henc'h-Levy being the higher. Fi y,'e 5,(M 5.':ic.h ces:. e crilical heat flu:: for unifor., axial ( h:?t fi end 1.c..-P .i 's..:i r< J , _I pding t o predic t ;n:is, sh:w:. tha t about ons-l.. ' t.' th J- t a , ' n t:- a re i c . c. ti.cn the Hinch-:.uy prediction. Wher
' ., : n i: L Tra, : m.,
c e r r. . ., ..
.. i s . mn-u'.10m sxit.., rc l,iux distributions
' is ,
i ni r. ; [.. ( c c s i r.e ct tiet pn , i:' tle h a;,), th:.' corparison in Iicare (9 ' r e. : ' I ! ' . Ib ii i s H :. tl.M , for ' .cf r'ajority of uete p:.ints, t! ? i! .;.-l" p r. .e-ier.s t.ut st m.tici,ly ex:ed t?.c r: -cured critical hsat fl u: lia.s , i t cr - br- conclu j.r! th= t, fer red L' uncle hea t flu.s distributiord i and Ic.: .W lue ' L., 1.hich correspeaJ to t':sc found in a GE BWR (non-unifc m sirl b. . ilm \,ith corn r pealing,12 f t. heated icngth) the i I!anch-Lc '/ cc ,,. .. .'ca does not prev:de a locier li: nit line. Tw. pr ' . i c.i . tersunr. thy the Hench-Levy correletica de e not provide a loaer lir.it linc to the ATLtd data which were obtcined with reactor fuel as:,emoly-like rod bundles are: (
E 1) The Approximately 700 critical h:at flu:: data points for 4 and 9 rod bundles, which for.a the basis for the llench-Levy correlation, were obtained with vaiforc; exial heat flux with sca:e data obtained with ! corner pnkirm or interior rod peaking. As sho..n in Figure 7, from , i P.eference 4, t he criticrl heet flux., for a single tube, is substantially ) greater wit.h unifor.T axial heat flux than with a cosine heat flux distribution.
- 2) f er a 19 rod l' uncle, using 1000 psio steem, Patzner et al f )
dr o.1s t r c '.' d that, c,'er t hc hea.cd length rance of 1-1/2 f t. to 9 ft. , thc bundle Svera.:,e critical hest flux increases as heated i.er.gth decre n u . J.t a givcn steam civility, a pressere of 1000 psia, and a 6 rms valotity of 1 x 10 lb/hr so ft., a decreasr in the hected l ir-ngth fr n 9 it. to r ft. int e __v t he : r 1 : a ! t r :t 11: hy &:.:t u. s , Si..ce th- h' 9 :"d len:'hs of the 4 and S rod Dr.dle critical hest flux data which formd the b6 sis for t.he Hench- Lr.j corre16Licn s cried f rom 3 to 5 fact, it folle'..'. that thou daia prov;id gre0Lcr critical hett fiux than .:culd brie been obta:ned had ti.e Sc6ted length b,cn i:. ora representative o,' ac:.001 fuel rXs ; i .e. ,12 ft. From the abov co midert. Liens, it can be conclud;d that the critical heat flux data und to develop the !!cnch-Levy correlation were high with the result tbt the correlation is r.ot a lower licit line vihen af. plied to Cita obtained in lor,r: (12 fi. ) rod bundles with non-unifor:n axial and redial heat ficx profiles. (
4 E 1 b.
.O.a M ri sen o f Pench- Lev ._ .a.n.d...Gr.XL_.f...or. Rod f.undl e Po_
It was previnusly shown that, for non-unifora axial and radial heat flex distritwtions, the critical hect flux predicted by the Hench-levy correlation geacrally wn substantially greater than that measured. Since
' U.F.L is e close representation of the experimantel boiling transition data, it niight be cr.ticipated that, on a bundle pcuer besis, Hench-levy will provide hip':u vclu? than GEXL. That is the case sho'.cn below.
Figures 8 and 9 shca. critical poeter, for a 7 x 7 rod bundle, as a feuct;on of iniet subcoalii.9, for reass velocities of .5 x 106 and r 1 x 10',lt/ar-sq.-ft., rexpectivnly. The two cpper curves in Fig.;rc 8 rga rese... t!.e " w':-L' vy correlctic.n and C.XL, respectively, along with experiuent..I di.6. It is seen that the Hench-Levy correlation predicts v.: : t L p, - c-^ te p :cr tSt., LJ.m ur the da a. The tia 10001 curyn represent the cpwating cunes for e;ch of thc cwrclotion:; Hench-Levy t is base:1 cn critical hect flux divihd by 1.9 ULercar, GET! ii. in a typical cm , based on critiesi Lendle pot.er divided by 1.2.* L'ute tnct, with respnct to t.Se o.m 'ating curves, GEXL. permits ' igher pow 2r t': s. Hench-Lay a t 10." inlct '.u: ev. lings vihile the converse hc W true at high inlet sub.w!lng:. It c nss v lo:ity of 1 x 19 '6lb/hr ri. f t. , as shown in Fig.m 9, both the Hench Levy coreclation cod its operatir.g curve are always hig!n.r thi.n tL r2spective GEFL corvus,
- c. [g g isan of fTLi.S Pata with GEXL Our /.tn. co: :.ultant,(lo) indepenJatly comnared all cf the A-'.AS dMa, cons is t it.9 o f LN.. di.ta puints, to the GEXL correlation (February 1974 ver. ion). tlc /.!C cor.:parir.on has besically subr.tantiated the claim by GE that f
%f disow.ed in Section 4, ow of the arithme[ h mean, r;th;r than th3 yeom e t r i t a an, in the sL0tistical analysis, will incrcase the required CPR f nra 1.70 t o 1.24. This ,.uuld correspondingly tower the GEyL opara*.ing curvn no hm.h fimere 8 tod 9.
l 1 l ' E l l [ L l l thc Gr7L correlation fits t he ATLAS data with a standard deviation of 3.5':, the value qual:.C prior to the. second data subniittal. ANC hc.s determined l l the stand.nrd d; viation to be 3.7" based on the data from all 69 asseciblies and the Fet:ru::ry 197", version of the GEXL correlation. l r l , No sienific;nt err.7. ti ends in the correlation are observed with resptct to th; it,put vai iables (pressure, mass flux, axial power shape. ravir.1 power shEpe, inict sdcooling, quality). Waile small systematic difftret.:es L;tc.:^n asserbli.E are sh:mn, these differences are not associated vii th Ley particu! - p hraon:inon. ; The vari. tin. of the ratio of the difference bet..sen the raeasured end a ci:lculne f pr...w t.o calculcted p?icr yith axial flux share is sho in in Figur- 10. Uhcn cor,arcd to the Unifor and casina p.:2.cr shapas which fo, - 7bu i. 4 e-- t; e 7 7 and r. 5 [ yL c n: c r.1 [ I is
;.rs, the in'et pect o .: , - ; ;, , t.o !e slipaly car.,ervati,c u*.ile th0 outlet pe?k and I
det hie h,. ' r,b. ; ;s a ra s l i:,dly non-co:.serva ti vi.. The sut:u relative I l v5 n c . c::i al pc :r sher.c s i t, po .1. . e:15 p 3.19 t.!.. she..n, in terrs of 1 4 1 p: c . l r.t a t <. .surad ; , :;, on p. 5-5 o' F.cfur:nc: 3. H r. . ; v;.- r it shculd 1 be t.t t e t' t.1:: ther; c0.:is c i!0" . cre arplicable cniy t: t:u 7 / ' lettice { c' e's ta % inle p a , w ic' L, c:.d douch h- p exial shapes hase not, s yet. t een obtcint:1 v.ith the 3 x 8 lattice. . Gann al ricc;ric h :s limited the nlicat:en a the CD;L co rel3 tion to cond tion t.!.or the inlet rub;cslit:g is 10^: Stu/lb cr less. Hor: eve r ,
- .s rn t rei t' -
it h rcaitect in ir.lti su'coling is cbserved ov:, the entire range of subcuoling s this restrictic- daas not appecr to be f:,cessary. l i i ( 1 I l
4
-l 1
E 1 t The sensitivity of the 7 x 7 and 8 x 8 GEXL correlations pressure, to inass flux, an't R factor was evaluated. This parameter study has shown -
^
1 that at a particular coi:bination of conditions it is possible o predict t
n0
1 x 10 6 lb/hr sq. ft.), and at high values of R factor (R r 1.2). i The consequences of such a condition are l that it is ir:.cs?.ible to otir.in a convergence of th:- GEXL criti 1 cal qualii.y N iling Irvjth com e cud t' e energy balance qu6lity-boili length car m. This prcbls.? occurs only at short Loiling lengths where ! bojlirg trcnsitim' does nLt occur for E.2 conditions . If required, it w y te po;sible :o achiest a sclution by igraring the first 15-20 inch f t si'1 j 1; m Oni tem es ATLA 5 dm paini h: u condition, where i c wa tive q; alit y r:as prcdh ;"d by GEXL. ; For two of these points, a recsm bic solutica
- ss Uctained b ignoring the first 20 inches of boilin luigth. For the ot!mr tv:
0 paints. convergence was obtained, but t:w rer :lts acre umatisfcctory, Another ancialy in the behavior of GE%L i was +scrve,i at e high R fcctpr (R = 1.25) 1:horc e t COO a, and 1400 p ttc curve for C = 1 x 106 - lb/hr eq. ft. crosses tne curves for G = .75 x 10 6 and G = 1.25 x 10 ' lb/hr sq. f t. lio"tev';r, the dif ferer.cc betweer the tv.o highest mss nt:x curves is very sm3ll over the entire boiling l ength range. Unmcdified, the GEXL corrciation fails to predict accurately cation the lo of Loiling transitic.n. To correct this, GE has fernuleted a location pre- , dictor correction to modify the GEXL correlation for use in th e prediction i ( of the boiling transition loca tion. An error trend with respect to bciling l I 1: ; i s . O i t. eb ' : d in li ' . 12 Uhe s: the locatirn retidual is plotted vs. boiling lem,t h. This fi< pre shorts that at charter teciling lengths, the correlation, usii,9 the location predictor correction, gencrally predicts the locctica dc, a:.trenni of the mersure.d position while, at the longer boiiii,3 leng%, the correlf Lico generally prcdicts the locatice, upstrea:i of the nc,r.va a p:siti.-n. - The chortcr boiling len;t!; are characteristic , 3 of'the inlet i ected 3xial profile. Ficure 13 shr..s the locction residici vs, axiel pnier shape. Genurelly,. the scatter for the prediction of iccation it ars t6r th, t u ' t!.e pr. diction of p c r. L' : i l e t hi < r. t i r:. sho. s that the CCXL correlatic" does e p d job of pu dict.u... tv critial power for the /,TLGS datn, it rust be l,orne in mind that ih corrrli.tien- is c;..~pletely e r?irical ar.d na atte::gt h:$ been
- 3. e . . s, d e,'... ,,
(i, I , J.. .'. & 44 .. , . . . .G ,w p . a s. e , w.. e = 4 o ) l I tya4 .& L h., yIess v6 wi,J e',.i.,. ticr; for an.' o f the ti.frt in the cervelation. Consequently, the Correlatio: 1 i $hCI'ld be Us! 6 C e l'c IO r ' ' i tr0.nsit.r or '.li; i t.r.0 whi cti crc within the renf;. Of t',e th..1 1 - ny'd, n !'l i . (Onditions frc/ U'iich it '.<a5 d:f r IVed. O. E *.*d ) if . i P } li Y i J.' Cf tl3 f a C 1. 1.'en . #' . I'GhCh - !. e V,'/ cGrrclatic0 d Ms r,0I, as oriNol:y tb. M , prc n c . l .u lit it line to the recent (ATLA 5) i crit:cel h; i flux et; dica wee ebicirj in rod bandh' w'iich closely L vlt - r e ,. in' : s l a .; , i i :. i r m ,1 te . ', , as w anc3 l l l is tiare that t h:: r,c.. Correla tion, GEE!.. Uill not be fo':nd to ' 3 ai r.e dequc te , l i wi th rewr ! to Li:- icr 'iu iw of f uel ancOly thct: 2.1 largin, et son future date?" Thi; ques!ien cai: N cusumd ty the fullt,;;ng; ( 1 1 l l E , I While fairly c:itensive, the dcta used to develop :he Hench-Levy ! correlation were obtained from rod bunJ1es which did not duplicate the nurter of reds, the axiul and radiel het:t flux ptcfile, or the heated length of reactor foci anscublies. Heucvtr, test operating conditions (flow rate, ! . , presse:e, ialet terOccatere) did daplicate reactor conditions, In contmt, GEXL is bi. sed on nbre than four thousand boiling i i transitim d: ta points, r-ny of which were obtained frc,n full size, full l 1 Icogi.h, rod budles with a wide rcoge. of axiol und radial heat flux profiles. f urth:" ac;r'; th>L;mcc gri'M us<.d in critical heat ficx. tcsts were f very sir,ili., Ic t!asc used in a fu,21 assc's.bly and in addition, had the { sy 0:.ial ,s:,:cing. A wide rcnge ci op:ero. ting conditions (flow rate, pres ,sra, ir. lei tr ,"ceduce) dupl!Iating those of reccter conditions '.ert . . , , . .. .-. > t,c.. ,s l' . ,, m. u, s . . u , , . n ., m ,..a,.... a.ua . .. . ,; m s m
- n. .u c. - : t. .. : s.o / . . ..- ,
. . .i _ s e;: cept for thc rt:.cd of hming,' vi, t ually d;?liccues of fuel asse.dlics I t,! .: 0 the ex:ci hect flu / distribucien nf the tc:t and fuel assenbly t ai mr, 'a , ti A7vh test deta con b3 cinsiden:d to be in th; nature cf ) calibratico. L std cc. the cbom evidence, +bre is high assurance that the ATLAS test ast rnlie; and testi duplicat' the there.ul perforr=.ncs of fuel essemblies, h am this, it follow: thst the GEXL con elations (7 x 7 and (: L' ) , i i .. at e h sC c.n ATi.A3 C r . cc he evected to f cit! rully c,irror tiie thermii parformenre of D'..T f0:1 useelies for condition 5 which fall within the pnscribed limits of the correlations. Cased c . the very detailed. independent evaleation of GEXL, the Iench-Levy correlation, and the ATLAS data, by our AUC consultant, the revelation of son:e anomalies in the ( GEXI correlation under certain extreme conditicas, does not seriously flaw 1 _ _ _ , . - . - - - - - - - - , - - - - - - - - - - - - - " ' ' - - - ' ' - ' ' - - ' ^ ^ ^ ^ -. - - - - - -. E l J ) 1 I GEXL or it: , utility as a prediction r.ttho:! In total, the ATLAS data - GEXL' corrointion coa.bination provides a distintL improvement over the presently used Hench Levy r:etht.d.
- v. t y;z u...r. p.,. w...
A. T._h._a_ .A.t.l a r, I k .:_t_Tr..a n_t_f , e _F.a_.c._i.l. ._ i _t y . The GLXL toiling-transition correltiticn is based on data macsured in ! the IsT;. c) heat t re...s fac f r ei li ty. ATLAS was constructed Ly ti;c Ger.eral I Electric N.rpeny it , t:ie p.:rpose of coing steady-st-te and tizn:.ient tha.e l-n t c:wlic r:st; of feil sc..le clutt ically hc.ated rod tunCin v hich s tru,oi t e i t,tet nr tre i. a In p rcpm. . c.1 for this review, a tem of four fdC P.egulatory staf f tic ~.rs in:. : c.nu.itants visi ted San Jc:e, the site of ATLAS, to witnes . {. i x.,4 ...:3 i . . . i. a . . w s.no, o ; :.. un ;c;e c,.c r.3 t i v.: oi Luc icup m.i.h at nn es cf the llLM userati pu staf'/. The fellr ing discortico is bc.:c:' on infor C'i , obiainad Lt thr.t tiu toget'ic ' riith a writ ten dei.criptien of /M.l.S p.'en ded by G: and ir. di: x ted at areer rdlating to 6ccuracy 6nd reliabi','o er the ^TLT2 tcs. rn uits. F. ore deteils en /,TL/.S cet h. found in a letter, dried duly 21, lH3, frca J. A. Hinds to Dr. J.14. Hendrie. p"
- a. .T.h_e L--eo n ATLAS is an all stainics5-steel loop designed to opcrate with water at wide ran"n of cc nd;tions up to the following maxim.::
22'iG psig sys te:9 pressure - -4 l I G5L"F sys tem to::;>ccr.ture 1000 gpm test section flou 17.2 1N test section power \ ( l, I l E . t It can therr.fbre be used for the full range of steaoj ttate testing appropriate to boiling (t:nd pre *.arir.ed) vater reactors. Further : ore, po :cr, flow, and prest.ure can'.cols are ovailchle to simulate a wide variety of transient end eccit.'ent mlitions. 'lhe pa.:sr sup;.1; cc:,3 a.; of four silicon controlled rectifi;r units cach co:cpdrtd cf 93 SCR cells balanced in impedance to equally share the load. i'c i t e ct te the te t section is centrolled me uclly by operator ce!ju.st*.:.iit of a 0-10" dem.nd sign.si to a fcedbach centrol systw. This c'.n t rul r;:ct ' i.l ac; 16 fir-:'. p%:0 f t the gate; of all SCR cells so ~ es to ra the errce b;ttren 100 dcrand cad ortp.'t v01tage to.within f.1//, , Fat t. i..':. i cat i : ; t: th2re is prnvisicr for,.auicratict.11y follo'<.ing a progre~xd p'r t ? r hir cry w: e c. tira cca tant of les than 10 ms. [ 1 l.t: r .i. .. r . r the rec.ii.vi vy tg.s n S io tv : g rd to a full p.:'r v 1 + 3 o 10Jv. Th. centrihtion of this ripple to the t es t rec t ::n pc. ?r v:.ric.s f ro- C.C'l ai, f ull power t o 1 t- 30: full pm.er and is cccouat:6 . . by a car.p:nsctLJ,1:all-eficct ..attna L2r. The output of this v.' h di .:tyee cc a digitt,1 recdir.g h, U k.z:tt < rnd is . O avtilisle tc ti:? t, < ccybisit ia cod cc. irci sNtt:r C:-libr.ted CC shnt: a re a l s o t::2 J t o cc l ui; . .: the tes t sec tiac ;; .. r. Th y ~ ear.ure the cerce 't frt.' e ch SCP unit and the resultant calculcted pc,s cgrees I with th ur tti.: tet t ,surn.an:. within i 1: . Ra:un&nt rre surwentr, are al,o isde of other ::are eters vehich e affect LLiling trtasition: - ter t sectio; inlct te .perature is measured to 11 F 0by cr, RTD arid checked by three Chroml-constanton thermoccuples ( - test section pressure is rear.ured to t 5 psi ty a Heise gauge and the pressure drop by the dif forential pressum transducer l l i 1 E I - test sec'it.n flow is ua.'sured to .4 1% by both a turbine flowneter r.n:16n or4' ite/stevor:innueter
- b. l..e n.* S e c .t. .i o n. s B :.c s t Le - '. i oi s e,.on i s t o f a nu:.':',n' o f hea te r rods at rani;cd in an '
. o r ci,y i . v. t i r :;) 10 LMt of thc nucicer fuel b.:ing .sirmlated and held by grid p, s of t't c.pprcp ia te des ign enc: lo:ation. The neater array l l l i s hop'.w: with4n a flct chanA l .sich r,ccurc ily iuelates the fuel c h: n..,11 e . ! ' . 4 t I ' n i. h f. c . ' ' - d tc . ht " hr'. bcca 0: ne ct ed ch .icr ily in the heater l l l uci!. T ! .: t .a c 1 ;h i!.utior f tn heat ilus, is, ib':refor , descndent i Ont.h3 l . . ' I. ! . l i C . i ( t :~.O '.i h ! h i ",i; ' i' dC i c t ' !!tPd Iy drCfir.!, th0 tub 8 OVOr d 'N r ; 20 ": nd. U I;'t i l i r,p t rc h ition 1* netected by electrically in- . , , , . s . .. , ,e '. l '. vl, .,11 . .. . .. < . sn , s ,..,L . $ e s. - ...w. itL e. sl* 4..ww... .s.'L.,... I . silv0r sulie' :.0 t's inor. $. xf :e of in.' hE: ter .'t:ll . For c; iall.y uni-f u n., [ U . '. c J i ' t.r i .u . i &, i.a i l i * ': : i w. ~.
- i' . i s kM-:n ; c ec c'c r a t t h e 1
1 do,;n y.;c;.i e: ' r. d.e tes t s-: ' i r n , e n e .neral , so t he-r,:: is no prcble.1 - l i:, la.n . ~ Lhc th::cc m: E; c c t;.? ccc n . t asial loc tion. For oc';- un i h.n.. . .:s es hs. :v ' , h'J i l ino. t r; .J i t " ,,;cu n ov r a rann. 0I c.xial , mca tN. ., uwe l s y , . 2,cc e.1 9., c . I o.r. t v *s t.nt chann:1 ,iengt, , ,.hc i re - I 4 t f ore, in these ce'!. s , E ins talls a lar, er number of thernaccuplc:, l 1 selecti.c a 'aric ty of axici la'.'ticas ' m.i on their expa icn:e ..ita non-unilom te:.L scctior; 0.,ce tesi.s begin 01 a particciar test section, i t soon bec',.,'es a:v ' reni. Whc/c hni l iim transition tenas tc occ'. t r.d t he n.ucou p l e r, i n ! n, .i s reg ion c.re uun i t crea, pre-erent. ,.u. , i y . TFe error l ie.so( la ted wi th this detoction precedure is ininit:al for tuo reasons: ( l 1 E l - cor.ide .b1 u.perie..ce has chc.'.. that boil-ing transition it I ir,i la LU jui.t e,vtrem of a soncer grid on one of the higher powered rod.: . c.nd '.' <_rcouphs t'm att:tch 1 accordingly. - if, in spit- 7f th. P.' * ;-rvien're, !+il ino tr: nsition occurs bet'.;een : ivo ph ie c: ;h';Mccou; l's, coly c mini! cal pow r increase (1 to 2 ' acca':'ing to eg. rN ) vill ceuse the !,ciling transition zone t o adunce tr. th' th'. rr 2.uple Ph.ne. Te t to,tioci. ' H r i e , i n /.TL!,3 h . -l n cl ud :.d ' x 4 cd 7 k 7 he tr l $ .-. .,t., , i e '. ' s 1, i , ,. t,i * , , -/ .s ....P . . (. ,A. .. i. r ,l .. . . , s . i a t t. . i ,y ui ; ,, O T. . nU , r.- b'.!:1S U , < ': C .i t ' 'l1 t'r ? 8 X 0 ' Ol O s C 07:5.f O r ry , i t s Cf.;h cnSC L 5i'.; the . d, .A ,. ., ,. . ..s. ,e q ,> , ,. . . , e n . . w. . twi It i ., e , r!'. IH 5[ ."'.d u t t ! 2 !' :t' 0;;i Pl Ifi ' . '* e d W l ( '3 C.I thC h . i lit bi,M dl
- l . , , , .~..
.r.. , .,,..e. .. .,i..,. . < .r. ~. .. .. f t . , n. r. . ( . . .. 1. r p.. . v . c. . r..:} r.c. , .i.- 6 t n' r.'..,. , t , *.* 5 t, :) , j'.. ,
- e. . ir. . a r , + 4..
!.y. J r. ,,.J i. . s p.. .,en. . ......i,,o.n, . , . .s ,tst. r ._ ., ,. . ,.i. .u e ., , . . . , , , u, m .c , to n. .. t ..1 . - o (, s ,. . i .'. . o, . ..s, ,,.,.cp.. .e.. .i, c,4,4 .t ,= i i . .. , a p, . . , . i. g . . , , in. q.. , ' . -. . ' '. r. . 't h. v. +. .s .e., t. F, . . n ..: 1. - r: . '.i nis..i. r 1 n r '.' S i c + 1 . . *l,.. i. . . . ~ . 4 . r ! .) e. c.. . > i .; s-t< . A. y. . . v *., ( ., O.. p i :. ,. !3 u# -. . ~ . ,- l d.is . . . ,r a. .p. g { $ ),. :i t' V .i O O ,. , f u. t , .'*.L,- , '3 I. t :' T' . , ,!,i-i, l i ,e IL,
- 1. - . -
s.-. . = = . - t, J' . J ; L'- ,.(.. *. i - (.a ,i.. t. i,j.., v3 . ..: I . .ni y%.dO h L.r .n. e,' 'l r. . i ., . . I. i . 4.. / i/ . e, -) .- . . , ,, . . . 6.- ...,;., 6 . 'i .+... , e. ,;, . . .. , . . +, ,; .-., .. I 4.1) < .,h ,.e. l.;ndh 1ithd' c. !. '. i f a r s p r i r. ' , , , The u : f. . h n. WJ era t:::s3 shm.n 4! Fiyr: 14. Th e.v, uc.re c!.u c n to , i : e:c L tim '.:ider t rang:- ;1 shape n .:; -ir ne,i durino the i i core 1;n . Cc . i ' ig '.he wW cc: , oi [ : Ain't fc c . ncli:Jed in the ATLA 5 pr. m i. n ch te:,tr appear to I, .ralaie es cic n '; a possible the geomatry so, po..er distributions c,':pa ted to occue in 7,'R f uel . . l G J i ) l ~. I n P4- j i l- ( c. .T_e s t.. f_r. : c c du_.ro.c_
- The fello.en.; ;*cocedun ir. used to iost.ure boilintj transition at steady state cc nditlen :. The inlet t w erature, fic. and pressure are selected c.id h;l' ccm i.ut by U lo g ni.rators. Errcrs betweta the selected and r.:c a re ) .fluet s ' <;n :J nlar;;.. ehich the 0:e ? tors rc.ay emoc1 by correcting t ht .e t n a:r.* ?rs. T'a test ser ion pc. r is sloaly inctw sed by epcrator i
- , w al c.'f' e a.. M unile th2 cy:ratcW cc:sinually n:eniter' the operating condi Her and t'e t,L ip -h: rt r9co'cJs of th. ther: at:Uple si:ysl: in ter.dc d to ,,v m. . a ...
,:.... . i l . m; ..rans e u:-1 .I,le c:n. '. a , u t :1n . ,:. .aisi ti ca 1 ; 1acn .,.c, i m. 1: U. i t e r i.e:: 00 a , m e m t.u ar t c. at. .e.. b;caim . . At this pc1:;t, all t!.g.. ;c ,'> si .is c re chech d an + l'et, - One to c'.5"re thct ra t.hermaccL l t. t1:,0 u , o. ling . cen.:cr...c to t,e s .r rp~ t 1e r ,t r. cerv,er is 14.ulci rc,a , ;.,i l s e ni . l . ,, , a: " y ..- a-:..e . 4 . . ..* t v.o . .- I - th ; vr t i r.c.it c.c t.c . The e .G E. 'ing d-U req'. .r"l to ...coss the result and 4: ! caco J to tH : "
- .w crc r;;itutd 0.'t: th ! s'.d't coli ng i s che'{ : t: c a ,., . ., .e
..q. 3. , 3 .. . . , . . s. .s ...a r c. . ...+..2 . , -a <o.
- t. . . . . ,v.. ., , . . r . , .,..u.
. ., ) 4c.>i . ,, e, : ,... , , . i.t.. n s-, .: o f S Ww i , r s i s c c ;c re d . The f l o'. rL t c i .s t : ".n c hi ;: C ?nd th. prcced.irc I.e s (. . . u.. f c r tri n 1c nt tcr,.s, the tro.! rct 15 verle:. e,y t mer circuits nic ,o ottc/ t an t.ir-c: er:Ad fic"; control valve er.d the power it varied by a procrea .1 funcLhr. 0;; crater. The rw Gte, includi:1a hoster ti.-moccuple ~ I Sifjnll' con bC Scmpl 0 d 3 9 O f t::' c 5 59 ti.': per S6Ccnb End FCC0rdcc On I!::i'jnGt ic l.B pe ful' S Lh' 03,nOn t pTGr.cL 5it.'J. . ( E ! i i i, 1' i 11 .C.o.g.b..r..i.e.o.n.._o.f. ./i.Yl. A.S .D.a.t..a. .w.i_t.h .Cn.l u.n6. .i.a. .U..n. i..v..e rs i t y D_a t.a. /v, evidt.nte of the accuracy of the ATLAS locip results, GE repeated i a 'tc ,1 tcries run earlier in the Colw;bia University Heat Transfer i fec:lity. A ccNgarison of the two sets of results is shown in Figures 15 i thcough 19. Estiw.tes of the percentcgr difference betwee;i curves drawn through each tet of date, shown in.the lower right hand corner as A. i lie bei. ween -E (the ATLAS clata ere lo:::.r) and C (the ATLAS dcta are hip..>r). In: vie.. of t'ie scactor ;hich tyr.if:es beiiing i.ransition ; tc.:. ct... th: "o' c.. ' : c U.:i vc esi ty cod /G LI.5 100; : egres rzn ..r!:cbly well . 4 C. EvcP' tion a , I n g u r.:1 , t h isTLAS Het t Transfer Facility comares fr/crably wi8 i cay itci1i ty b ciw ':n ?^ c: tt"icL: c. ttsdy :t 12 n 'd 'n *in' - I Lv il ir.g trr.nt i .un tests. Furthen-0cc, it 'incocperates cr. cial features: l > cuiu ? ic clarm c.ys tem to esure tcqtrimd test conditions ere clescly r.d - sr.ccially desity.el controis and du acquisition systr r to facilit- % trsngie ri tests - highnt te:St sec.' ien p:. cr cf any loop which ;:al.c /JL/.S superice to other fccilitics and ensure e valuable source of date useful in the safe design of Cl.Ts. l I l 1 E ' /. i G r. T r'.'.. . 'r M i t <m, - . n.", . Cereit;l Elect <ic Un:.esci to establish design and opart.tienal thermal lirits b m.J cn thc re u correlation. Thora limits were previously based c- t h: : e r. -l et co rdcticn. The GLXL correlatipr. is based on a icrger . ;nt of '.xc r tim cen. : Live date ths.n the F'cuch-Levy correlation. The GE.':L r.orrel .t i .c is a tc.;-fit of 1.h:> da ta .thile the Hen:h-Levy correlation i it alu, 1is,it of t h .: d L *. e . 1 C: p ,w e's La !? cit the. ther 1 linit in ter:rs of th critical ro ser 1 l l .tio (CP ) 'i<!, i: .101 c.il,. r 0.r c .yc n c e. c .' t h.: fern ci the nec ' /L . l l ( 0 ; '., ., ' .u t . .. airo wt ra- :.cn t;.r.it > c' the av:.ildle therr.al p* 3. Pee ico:iy, th th.w . l li;iil \;a9' slo t?d in tcre cf the criti',si ht ;lt - ' u (CN :;, ,5ici is r.c dioectly relcted to the the"o ! l n ra n. .. 3a Uu .; c.s 5 o c rr ' ! . .innt ryc caccel:. I i . e:: ca t ! . r: L. f i (: c '. il, ' ' i" . .!l i Jr a ,..'s.i I:b Ci c. ) v . pr0; c.: 9.. ' . _ cc: ;in; t! ef fe( t e i toe. vi, c 3. u i r. ? i u i r. : b'- CD" .ccrr.it ir c.4 re ; t is t'p t. n c e. ,it :ics in the Occior 0;. ira ting ve i9 le., in de'.J.c.iai ' in; i.!- :1 liait. . Previc.4y, only n.:.~;.nl veluc.; ef the c.;.c r.! ,i rt . , n _ t ' '. ' >' uLed i; . :: '! 2 rmi n 1.uct flur. relativs to t:. CHI lir;it li .; li l t . . 1 < u ' , i t i c t. ' cr lysat i.a ,7 b;.2n a" plied to tbs previout DiF there 21 li..: 1c in orc' c to e,;luate the e?fect cf uncertaintic: ia the cperating '..a bl( ., the dira t incnrporttica of unccrtaiatin in 1: ; pec:cssd CPP thcnn,:1 iir,it acur.os th. t uric tairtiet m e considered durin] design ' a NI (> h h f o e r, ,$r IfUr(hI GF pro;9r.et tha t trei.s ient:, c4 ;ed by Uni;1e oper ator error or Pqu il'.Nill In't } [Uhf. l I .HI 'hr!ll h0 li:MIUd SUCh [hdt COD 5lderiW) UNCOrldinliCs i ( 1 i l E .27-in definina the (cra op ratintj state, Jr. ore than 91.9 ' of the fael rods would be eg.,<.tei to evoid hailing transition. The application of this desitn bmir, to the C.ter,1ir,d. ion of steady stcte operating lia.its is in two steps. iir;t a strtit.cical mahl is ustd to calculate the minir.;um critic &l pwi.r ra ti0 ('.~"H) for u' it.h less then 0.1". Of the rod;. are expected to expericace Fnilii.c trar.si tion. Set 0Ld, a tran",ient model is used to cal:ui r i the thonge in Cri' resulting iron tr. nsients. The stead)-state operatirc li-i-is df ur ind as the seN of the lcrgast ch. age : in CPR due te any of the t rie ," t x , i t c. . . . d . -w cLd U., !; cpi; at v';ich les . th.m C.1',' of the i ads are .rv exp:c, , to e: p. . r.a boili' t r: r.s i t i t:6. i n. t r:+,:. icn t n;odal (lEDD-i OE ~4 i' ~ is ti2 ..ir ' p evinen's t.:. cd i n tc i r.W ti n r +F i.'ench-Levy CHFR lini t , a . cnd is not c '; Bjt;' cf :li s rn :r.'. Ti:e staff it revie' ling this subject ev.. i 1, , . l c. : . ', ,, ...', , , , , . ,...,....n.-
- i. . : . : ,
. . . e., , 7. .,. . I th: CF' of the ;;"ndle: m the cor cr.su . m; c rJ vd i.. , r di U > i'. .: tion an _ va 1 r e, of :- q 'r ';i ng va , : hi e :. . sirt th'. crlcult'.d vt.lv:.3 of C?P., the pi,::iail ;iy ai l~ -:l ing i.r. :a *. iun 0;cer 'ing is Etmc f r i.11 ro.;r in N: Pcanive tricis "r."_ : cedeu v - tiro; i i t' c.Orrtim v^ ri c' S ' e : .:. d .cntil un t s n c.'; c v# : i eri , U-pra v 1:1 Q 0;- p ri ! . o; t r . r. i :. . .>n o c a.1 e n g i n '.., - :.re ir s0. . j The prc 4bi ~ii ty c ? boiling tran< i t ica c:curri.. i: c61 c u l .r ' t.a r e d 4 i on the e,tcad:r.1 c. .ritticn of thu /,TLlo .'cta rclative to the O.'. torrelatic', ( j ast,t.:.in; a nw x) d i"; t e i but lca . Erfi, w caly tests wit'r a sy," etricS1 cair.e axial proPlc are ir.cled?d in the C ." deta, th : ani'u G o~ the unc. cluin'y in the 8 x 8.C01 correlatico ( l i l l E -PD-
- i. incnias :d' to be co ;menble to Ute-larger variability of the 7 x 7 dat a which in !n !ed fwr oti:ce axial profiles. CE originally used the anti-leg cf th; :or.n of the logarilh:n of the omucr of rods expected to ev:.r . ce b41ing ironsitic.n, in the dete'c..'in.: tion of the CPR limit.
ii. ? m :f ', Sis p.a . M c r.mc crdu xs the ur. certainty intcrval. The pv c M.n i ::s b wa . . Ji; i ed, by CE ,(o'I " to use the crithmetic n,ean. I! : ', rwfi M ,' i o en h. ceca".c in ths- i'.Ci>R for . hich less than 0. l: of the N& We t , ec t k c;:'x ience I:Ciling tr:tr;iticn. The n .' u * ' t ', : in uy , 'ing v6.ri dle s a"c h s r.' on es;im ':s (Lsi, \ * * > ^. }h f h! tb U' i . s, iii k. ' ( n [. IlI ts . " 'f , /g fU' du [f,d (C Udfi(O'O . b C f '- - \ Vi s ' , ' f. . , . n i ; '. I t f.O Vi . t bl P 3 vi.' #. . CDnt!'ihdt? SigDifAC: 'If 10 th: ' ' 'i l ax ; i t. i o ; y h.". .: b.:.n considercd. The est - '.ed v iur et the:c f,
- L .* 4 i d s. i,/.
,..s .) s. '.4 3 4 4, L...; e ,. L* .%,II* l' ( t, s i. h 4 #4 ! L e, s L. t i6v 4.t il c. kJ. '.* . iyi CI6Y .m .; . .ii. ., , c. . ,.. . . , - 1 . . . t....< -r. , ..- ,r . e,sj sL. i.,. ,, ,.,.c. , ". .... , g. . . p. e. . 8 t ., , ,; t, c , 4 .. . - n $'i t g 11,4 * ;t .- k #
- i. .,4 p.,* -s**. (
t.$ ') .t..f.. .b #. e s (l* .,s + ' '
- s. ( 1 ( *' (
. 's k '4 - 'i t0 LO a Cuphrtur.'J frC',th0 i .a- 4 n . . . e'. . .i f. g,,, .- /. . a ,a...hg, ( - . c. )'I . ... u .& ..-.J.s c"g , , , . . pg , .A .3,.g p 8 i ., g 0 . . l ,.. *, g 81 at ,v. .. g. g g$ b '.' ! I i ! . ) i i' il l i Or. \ ' !l t.' brt r; $ 3Oni tl ,' i b.. c,q: c t erj [n CcCur On cn'. T OJ in (.$ r. . ' . i bn . t (n o 1, i g r;, 1 ] , ni c (i,g , t h3 t ys'; {g;] a ; 3 0 ',,b ': / h ; j l a cai xiete.I;a.4P or ur.'-y). L' uce the p"spv;.cd basic for cere, wid:> \ 4 tra:Sicr.t: 0. l'.' of e rids in d.2 cc,: ; ws ,.l .! .9 .Get td tu e:<isi Gn'.e boiling trentitica et th thurci limi t (e.g. . a MCPR c' l.05 on t!.c worst ab h y lt a t 'Pi ([ [(((t9f), ,yO y p f , th} p((r[e{}d )ij ,i { i n c
- u p .,
-------------------m i 1 l l E ~20- i ) I l i unc:tlaint ie:, in t.h, reer. tor eparating variabler, and th; previous limit ). d oe s r. ': '. . .!I un:a:rti.intics in c;mr:: ting vt.riebles are not considered, there is %' c: < fin t.:c that hith a cpl! of 1.05 there is a 95C probability t a l tv: 1 : r.g t r:nu ;i i vn ce: s not c o;..r in L:. worr. t bun:!1 e. . Thereicrc, i a l'.'!;' ','o? 1. t 5 is rcughly c rp.ivalc !t to a MCHIP. of unity Or.d both , i 'f r . i g r : - . previde sin.ila' Ow." .nce th:.L boiling t ransition would .nct c:ut foily n; corc u >: i n n';i t. d r , No.. .r, far loctl trLnrie:U.;> the pror: ';ed design L5 sis is a c:aartura ft - t' ,h ;> .> i t t ! Prcr:.,;<ly, the cair.uir e<j U.Q r in 6riy c.,s e bly L l i n '. , ; '. t. ;'n i ty t' c3 y: . u ce c p c;e( to erp. 1cnce b:.ilin',, t p :". 'n . ni > , v ;.rc3 .w d '. . c i , c l } c i t h.- rods i n a f e.1 c m ' . ly e ic !., o- , ;c ei 1:. . ' g. i m e 'eiling ir::n:itior, uit: ea;; viciet .ig the L 1<. . , .. , ., 1, .',, , , , . . - .. s .- . s
- i. .i
- li. , . s, % i ,.. (, ,
..i e-; G...,.. ..y t. :.,,, , s e, a o _i U . 'l ; of ib ' 1. r.si e cc: J. for c:, i t. I e : ~' cf 0 '6 U. rr: c:ic a ,f d ior f .:o c t fu'1 c ' l;' , tPr is, boil m; tru,qiticc: e,;id ha ; : e;? -' ed ,, ... l ,,o...,._.. s. .~, , t ,o.. a. : r vw i s u . i. . , , . . . . . , s . , .. : . . ,s, t. 5> . ..: .-m,. < a... ino,s... ,s. t ,
- t. ,a_ p ,., v, ,.. a. . .c.
,'7, i o . b . .c i rov , As lets t:iu,' ' Pariin ivi t . r,. a L.culined ,. m,ia.' , p. r. o.os . ,.. .i,e ,a . .b u. . . . ' h'a c e:,'. ! " d .: that th > pre;'c 'ec c"siga tmi (i .e. a r. . th .n O' . cf tl.e fuel inds in tl. curc would te; expected to avoic' $ boilir.g tr:;nsitior. C O U", a, e," .s i n],: a 0p; rig r. . Or c(!Uip'.;i in0llu.C;iunS) is ecceptch,e. C P F 01- r. whr n .. .plic:.i to cor (. < ide traus n ,'ts su;h w a turt,ine-triv or ru' :'- coas tdocin i nn iu,t. 1. also cotlude th:.t the rathc.d c'u to cciccic to the MCPR thern::1 lia;il is an c:cepiable 1.;cthod by 54i;h pt. cr dic tributica l 1 1 1 4 1 1 I E -30.. J .q , i 4 i!Dd (th ? , l' , ' p f, , , ' .; h d i' r r-l i; t . 71 nnd tit > r .i.-<~ - t ..- Oper., Ling - Nnc. :ters or s . . I D C I L %I i r' + *h C ' k' t :tr,natio;; of g. ether the design - i bu 's i 5 1 L O . . ;l y.Jg (y- ( 1 ; ,- A. .e. l , V'- .1' '-"'i- O .s: n] g 3 ,j g,. ,, , , , u..t as centiul. rod it h>'h av a 1<> & -- : <h. inijj?pr; prig; y, . '( h v .- .s.i., s, ....t. r. .p , ; .- ; " t ;se: 4 u- n '.. j, h.yi;43 ] ] j g g, , I FU T , .i a ' r, 01.,s '-
- dt 's. . v 5 P L, L3 1 j pw *v,',vt rr n;
. -. ..+ g, ... , -' , , ,, ,,,,,),,. ,,a- '- a><. t. . 0,- )0:g] g, ne j ggg , 1 -1 h t d ..e e 1 l I . i 1 ) t 1 l l. i i 1 4 1 1 1 l I t-?1-i r.i . ,.I h... .- "- i,. A;c 7.J.i...e h i u t, '., . . . , U S. . ,. . , c10. . .) .s._ Th: O iit 1, ; twie : the 6:icral Electric Them.1 Analysis Basis
- r' i a r 14 :: , ;c -inr dec ian end cper; ti:n. Included in the
, ..ci.'e'1.c p r r '. cct m: ' '. it.a . i.:n t ii i s t he N.-i s for GIT /. ; t he L4 'c l y '. i ri l I A t ' ' '.!! LEO.i t0 C'.WClO[ t!.I7 CCr tlDt**:O; thG EXpriMEni7 I ' r '; ', '.' l ' ! Pt;r 4, 'i, ; dh2 t=,'rt*;l f.I'. . tyhihS.I :; Bnd th0 CXp!;rinCntial , .... ? (...., t, c. i.ll.,ws, ...' s..... .ld ,J'.,- .' tO L.:w. 4i. r. - .i. r t.' s * " . r.- H o .*. d 'a i ,, - r o .' s, . . s r. (. ; 1.M,. p. . . /.3 +u.:sl r e,1i.. e'- : <., c. .i ! ., t,o .) 4. i,s . . -i.i.u L i u. .,, i L ., i. 4 . ';,.ca .. . t ..: .s e. r . .,; , s, e n J.. - . i ea. . ..:. 'L.... , 4p i, t,. s e3..
- 0. ,1. ,.. s4 n .. i, . ...
.s. 4. s.... Cs. . .c . , . .i... pu.. . ,,3. .,-.. i q. 1 : ,., , - .(. , . .e .n..,... >, ' . V 6; g, g6- ' i. t .s mi ,. e Tr , ., L, .. s1a 3,.'s , sf. v , d! .. ' , . , L' . La " Pt s r.. ;.,..,.,..J 4. s ..,c.,-. 1* L. . . l ..s, .......:... i .i - .. e. . , , . , , . .....<e..,:. l' ... . . . . . . . <. v.... ,;. . .x i 3 '- 'e' " O f t . i .' C l . p ; i ;i J.3 l'U U.ii' 1 ; inf , a .,'.,ii, i.is , 1 61C.I'. . . g 4 .. J t -9 .
- 1 y90l 's U , .5 8I
;I r s'.1 l' .( P ~' 3[ U i .L 8 Y 'I l8 .S 'l d I l II I
- 1. . , . . . .
y. .i..... ..it: .. ,,.l . . , , y.. . .. .LL! sI t. . . v., ..- c. h. . , i. 1., . 6 Ie 1 . . . ,.L ;. v 4
- g. . , ,., , .t s.: ,w .,
44.,i . .. u . , , t. . '6 . ,- n o .; e. 3, .;i .. i ;, ,. . _ , 7 l ,. , ,.. 4. c .-. ' / i..,,,..
- i. . . i f. ,' /> C '.' i . ' ' '
! ;., - i p; ,,te /.7 %. . i Jt..1 .. f.. ,,. J. t p. . ,, f. . . . (i.- . . ..g...a u e . . . . e t, f .m,.%. ., g , , t sc. s . , v * + - w i C' 'p. 'J i .8... ,v I ., t. ' .. I. 7 , e . 'n,.. i. s...- > . - . !. + 4n .r a. I e .'-6 ic , / D.' .,,, d.I .I L i e . t . l . t.L.4(.., & i. .a h ,. a s . t 1 ,. 8 I..L PJ . s a .y - i L g ,. ,ttq- .,! 4 eia. b r . g,.. .,,e
- p. ' a. , J. tn. Aa .s.e .r6 ..iJ ...
7... . . ,i , e. c...,. g . ,....~l,i.,. . T.l .. ,9. rt3 .#L. ,yO , .. t. ,3 C C' m. e ' , J e, _ tl.ct the a m'i.. - . Le i n ruits rg r. M th . t' ".1 p r k .:cc of GE C x 3 , . f ' .h k. i a hI 9 . 4 ' 1 hk hN yn n f / '. Il s . j. (hb',h I ' e (.' . f *) f' 'sbte tP thC GEXL hhf" 1 i 1 4 ~ pQ h(iQr , tl0 (O*,C , (;d y (h3{ (hg d}((; (g fg hj (QggP (y ', { g ()y (pQg f Gli g g p 3 pp{,;) ) g I J *1 4 4 1 Il E t lisf ributed cimt the corrrelctions t.ith a standard deviation of 3.C and 3.0 for the 7 x 7 t'ad 8 x T.: CEXL impr:ctively. While u.nll systematic d if.:e c es h.:t u n ar.!c hiir,t.iih dif;crent gcter distributions are. lici..
- t h.: i; :, . . .: 1: tic '; sli:;htly ccest r#tive 1:ith renM:L to the lost p A - ';1 c 6i . u u ' t a., : s (i .e. , inle+ pd. and r,y ncri:cl cosine). Al-t b' ni'f t P,.; r,ry , *c l t ;it , l,.s c , ,.."; r n r,. -
1i.05 0.f 071" ;;;;' CCliTI# l i o n f. , GEhl
- t. r.i n . t; '. ' ! Li ,1 b i r, a dcij r,,,d t ;g;pi,c { nty the tl.';rr.*il p01'f tJrFi3DCC Of GE i
a l Uy '. in o' / ), / i't;;;l i f f. - .'il l t . f 0 F '. h 'J C;;p .,CcLc d 1,1dge Gi FNCtor r,0rc hl J ... , . .t., .., . 4, s.,. t, ' i .,' i:. 6.. a. a i,.,. .6 . , ..,...,i n [. , )4. 01 ieo .,j, . . i,,..n. . . m...s.. , 8 p.... y ~ ,. ,l 1 i 4 t,... e t. t iv., 'I <dl,, , .; ..I. ,r .'t.. w . a. .6 ,,....,9,;)*) ,. S g . .3 p c. , *! (.i ,;, d ; , ,,,. t .. ; 4 . ' '
- s.*',+ ( d . . < ' 4 '6I' <s.
- 1. r. s. %t. is'*ir- l . .. I .t .- l 0,*.s o i- *. i t 3 *- 1
' .* .- l i s. )* . . #i nt , 4,19* * ( 1. . ., . .. . . , y. , , . , ........, y ,',u. %. . c...,,-6 . , . . r .mi ~. . ., . m .. .. .. , ca. , . ,. .. g. . v . . ,s,. m. . ,,' ' 4 , , ,9 9 da je- j , R C, -.- l , "; 'g* r [. ; i, ut3 ( 7 4,17 ' d ' ; ' ". i f) .]lC g N fQr'Ettr sj;[CT b e.. C,. . . . .'. I 1 ,,,I' a . *f.t-f,,...,,ei s , lbe ., a,.9;. I 'te 1. 6.., ,:. & ' a ... ,.,8 ,J h I' 4 j. * , C P, 6., j .'J'. i , . !,P,y, d .eaya. ,. . I."" l C ' ' a !. l t. >, b 8, N i r, ', 's l i .' (8 '? C)[ ? J * ( i 7 , ' tf If , 'IC,FC {b;.R .' . I
- OE IhG f t. O r(, . 7 .!c h> r;;;c'..,s. '
', 0 c.y; i t' [ e 4Ii;,g tri : s i . ' t,q ) F U I', ! h' C r . DJri O:, ! W: a , :blF.' to '.O r a - t 'i d e t ri: n i :,- 3 . . iia c 's. /te, e,e rc.;9ir,' tL ' the E PR ] lii. t twi n.d f or cor. t.ia t ran:. . n; clso i.c w'j e s tM % f uy L i. ;i t d, ,'l i c.ib l C 10 i OC.3 l t ra r:S i r;,it h f, U C h U .:, a CGull el TGd tii thdrai?,l . h's also cm id t h :. ! .h s c:- Listical l'.:,Jel c'.Dd to deri c. the DR limit is c.ceptnH e. : i e .E-33 ~. P_E_T I_Ri.'!.L..T_.L. , e
- 1. J -
lba;:. r, J. F. liac'i, E. Anst.ca, cad S. Levy, "Dc.ign 'Dasis for Cri t ical !'..'c+ flux Leaditimi in Uniling 1-lcier ilcactort ," APED-5186, C l w., 11, l w . ?, - "C: w r,,1 .in .' ic U T!.dra'n i Analysis Bc.ik (gel'.0): Datt, Corre!ctiri n.. 1:.3 rm "p;.l k.u t .um " 2 O-1095L' T."lED9, Cless 1, I;0vmh2r 1973.
- 3. "'" , .'t.1 T.1 + h :' .Thw , i i .:.1 ys i c, !w i<,: D 't: , Ccerelatirn, end Dr.ign A: , :n 'ti;.u" :7 ! r 10 . . , (. u 7 2 m 1, Clas.s H1, Motc.-ber 197').
R . f . F . I'< e.. - , J . C . f. : ' , a nd D. t' . r.n:,'r t: , "P'cr.t D.t rnna t Hea t ir, ' Or in t!i '. Per s un g ",:,a , :l w ;e y,i> t m M in a Tube tiith Cm.ine 11 c '. Fiu D i!: t ri bu ti on ;" l,% -PC ',11 ( i 971 ) .
- 9. D. 1,. I ; ; . . , , ;- * -
g , t i ., n : .a r. " larced Conve'r c; eu mout ! ,'. f ',. :'- in , . , i' f i ," , P, 1, ;, l(*p; ~l',.h,je g ,. j l, j g g j ; gpg 3;}<j f g.gr,j ,Eg ., p e .. , t . .. # .1, I . e. .u * ..L,, * *s, -l. , , -/, t, j ';' , . ,j ) , 'e '#'. . ' i. .. , $. ! L , " l6 ' ' , k.!,* f.C ',' l !, ,: { [pli p f,h , hg l* $ }g g t j n . li ' - :<, ' / !L ' .\ U.-l 1;* : /. r r. .
- 1ct.[cn 0j ',.m.; q m n t. t h J , . '. s (# .) e a ., , ; y * '.t l g .j' ,
I t ? * '.~ 1' 'l'- a 5 'i- ./., ut 2 . t...n. s . p .. . t . ;, - i. . . s ip' ay. ta: ' < f.<, .;isec '. . *l , to r- s 5' - ><t-- i 6 N' i ~ 1. 5 , i s r. , , st t} , i,- ~. l , L jlQ1 ., {>) gnoeng 1 ;, L / ) , s l t *. I. [.O;, l .' , .l . (* . , [. r r,' , , i , ', il } l , ' ,,,7lg , ( r).j i ' "C ' . , Of Cr itict , IP ilu i n a M;- :. (c;, led by i.c. -. -. ' ' . , i. . !- ,. . i s .ic. * #i, , : . , ,,;,,..., r. .
- t. a., c. *s tr ., ,
h.~.. .. (,i.,. ~' , ;' i . i ")l.*i,,,.,,,.,-. , l. ; , 'ir . * ., M f - l '.- 3 .l.,, .s... . , . t,. 4. i ..'. , . . s /. 3 ' i .;- 4 i i, , !! ' i .1 L'_ ' C'- g ' aj , I ' -(, ;1, "' , S.; r j , *, , IP). b !. U i . I' '. i, i ., i ,' , ' I
- 10. i: 4. '
.il . .1 - .s'. ,i. s.,t.. . .., .,.: e, ,, i,. . . r r .. . .... 4, c , , F, Fir. ; ,, e .n d.. .' !s .a p - , , ,, g g , " , :n pg, , . ! ; g,. g ll.n.,,t T r--i. .. ! u , ., ,- , - 2, n. - <, . n' ' e , v, g-:, .,' h e . ' . . 11, ' ,. W i.1.:i:,, G. P ( .3 , ,, ; : ..'.; . G. f'e cr1or.cc , M. Eilvetri . c..e ! A . T,. , , . "l; . ! Tri.e . .; r C i n i s u t;, c ' ..c.c.i , m ,.c li! xt u res ,' E nc r9 i e M . . ..i . t , Vo;. l l' , is, 3, (p_u ), ll G. F He,6.t. , l' . t W n .' 'ep. r ; . ,:v,r.r.9' I, m i n T :- T+ : s c !!ect Tre - c - f. r Ur, 6
- y fia l;.J '; ,1A,, /, . [ p,t..p.1,,(19g),
D. D. Gu .rine. Y. f'6rinclii,. and L. Pe t ori , " States (.f / rt in Carnout P re d i c t i t,,0 for I:'.? Dud iniJ i es ," i rim... A'. Nucl. Soc., Vol. 17,(1973). I l l m___ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ , - T - 3 'i ll. G. i. !!n ; t . , !: A Ue w .y, and .1 C. Collier, "Correlat M/, of Critical !F u i 11 *
- ,e y, , ' : ,:1 r io ; o f Water in Unilk mly heak i Chaniels,"
/*,,i .l..g r. .r. ' ' :, tI l. 6 f. i i' . llarin( li a ' i. P,n tm i , " Simple but liccurale Method for Predicting I % rm ou t. In ! T ./J DenW u ," irane. Am, Nucl. Soc,, Vol.- 16 (1973), ir, A. ".,. y ,,,,s , . ' .4 c 3 I h.ir.. e/ halmtic , of th? htchannel Th .rnt:1' Mt, drt ulic.e . April ,. .m. ~. .- ! e r. N.. . r .. c.1 f i c , . c i ;c.c h, Iilor;: tor w., . . u . ysis . . , . , 4 l
- n. C. .c.c i - t o '_ . H. Sulli nn, 'Lvi .I of CC ATL/S Datc,"
- 17. l e 4. u : L: .
i ; s ,. . l 1 ,' 1. ; 1; D . . ' , u n t. - ,s i Ce.:oriinc, E , O . I :.n :. , a n d (: A. Wit!w y,er, L' f. f i Il i! i [' t t i tli l! ',I [ll' s !h'C y ' f . . . ;' i ,[ f '
- l i C j (Q li p(;i l *l d j
, y .- L . 'u . '! /, 'c .", . , , p,,cr (g..,Ljgj_ 4 (lg.c), I ;. - . , ' i .N jy ' .i. '..g l ; *, f i i ., " ' ' .:, ',t i g , O ' ( f.' G ' ,_ (',, .- (c ) f . ( ( .. u .' i 's , . i.,.. t 1 j. '.'.,: 1 e., e. a1
- 5'
- s. .;. .
, Pt;m s. , e .3t 1 i 1 s
- y { 1 * 't' ,.,.
,a 3, r e ,, r , y g.,. e. ,, . / * , - i i**i.1 .e t L- , . , t , ,' ; 4 .4 , 4 11 l.. . . I
- r. , s r .i . i
. t/ 4 .. ;,,I, e;,4 .s,-..m. Ci-'l34mi. i i...,...,.3...c ir . . t. L l . , 4. s ,, ;b. f. ti ,,, ., # ,i l .- i - . t,, , ,.,
- a. ' or3
-i-i,., t ,4 , 1. i i t I i , v. [, f.- ... g, .; f . e 5, L Q * , , ., i./ ,* .,,,,,,;.. um 1. 1 .. : e. i n , ,. 1<! t 8 ( i '
- t ' *
- r.
- e i(. ., j.f.,
, e' *i 1-Aj - ., i8 p , c ,,,d[. I . , 1 . + , t,'., /;. ,.. ,*,,p !. ,,.., *l r r * , i, f a n. < .. i i,t, t, . ,. ..t, r .. ,i.) r ;r e. ., c. s, .- - e-
- i. 't n. - l oe .
, . . .- s... t , a w.4 ,e.i.,~ . .. i ;. ; . - m %,. .H. , , . ( S,. :.r , . , r. -t 2:,- cs . 's,. . . ., a.: s. s s .i,. . s.. , i.,= s, , v.. . !- 3;. i .~, , ^ * . , r,s
- 1. i* . ... ri ,I. '1 - 3
.6-..,o i' e , - i Is , l, , e L. . r 't pp,, ',,, 4 i J. ->.l t'i t,,, .; }. -"1 ,4 r. a r . >t,r-, f1i i -) i i.. . j r?- L t. s %... v, 8 . ,ee,,,J, u , .,,., i- . . . .,'ip ;- i ,,e ' e e. i L' - - ,, . (.si i ,
- ,,.. l
. t ,.4 ( } .i. . o,
- r. *e , 1 :. . i r . '!
i > *- . .' . ' . , * , +aI J""*. j r,. .>T , m i . +- ] . 5j :, * ,. ; ,- . . . .p..,L . . y. j.,,,,.,.s , I?. 5 .. g 6,s, J . 5. + >, j i e i .-.r +....,is . .y. , 7. . . g...i.. . . , t. ; ., i .s s; . 1 [ . . . .pa p ,.,, , en..,,..., 4 im. 23 C f .- . .. (. .s:. . s C. i .,. a. f,,,.,.,- ,, .n t,, ,A t 3. v . .,. H e ' .1* *. s q>. .n.'. c* "; .* , . i .. .f a S. se r_. l ' - .J . L] 1 '. . : s, , I 1 i 0 I l ( l I 1 l l i E- 35 l~ l i 4 - g __ _ ._ _ o q 7 _ _ __. . _ _ - _. 7. . . _ i I, b:..i. U.: ,,, y ! i: ia < i
- l. 8 6, L., ( f:
c d -
- c. H t
1-7 < t c;<!. < a : 1< c. 4 ,
- t. . - ' -
e e ,- w - Q) i i ., c- " r, : a I iG < v: i i- ,.~ lL:j'7 --' ' . , , f. 6 '; s .' i ,s_ , ; 7 i 3 O l f, f , e9 .i in i ! ' i '~ ;l ~, i s: O l ' ' I, . [; ' l" cc l i l, [ c~-: ' '. - - t
- . -j h
L . .J cc .N i jw O ,*--l t i o i q t 1 j ,; :H l a* '!;.1 c' i ga a L., l a i .,,, - I . .. e} } },q F ,5*,...1 , . L. t.- ee "1 9 it.,... \* A t L- ! :t, p' l '. .
- t. g c j
i j , ". . , , . , ;i ; i.~ .. ,"' I .- ' H s- l t l... .~, x. J i c:. e j 1y;
- l P
r, , I ; ','l'*',I ' P, ,il.,<;j .o q; a 8 l c. i i r.., '\ I ,4 i,., 3 J . g i r v ! l, l , l. - ; s .'L,;~.- :. l . _n.L1_J 1 1 - s ..a .). ; * .,. 7 -- j g I .. c,. \ i - '- i 1 l i I- ' ? I ) g Y .' t< f6 i f t i .. c i ! n 1 m > t ' *} t p M ! l s. c- e -e - .i _o l ia .: ... \ j \ o. p~ . , i G 5 (, l j i . _J C D ; tv .r'. C ..m *: *i^ s., . ! t4 1 . , . . .! l L.* .O. , u. t': i o r; A i
- t. U,. i i C5 g .I .a
~n n~ i 1 g ! s' ~. c, <, a l' g A. ' I c x..- ,c._ J o _' ; o u. l e- \% U l n 1 .E u. Q y, _ _ - - g t c. ; t; k *) s& D 3 o c; d W *' g ., ei y c ci * ~. O C-, ? r1 04
- O
', 1n 0l Un3 J Y AiFIVD O J/2 n . , l l l D-30 g .i i l t I 1 1 1 J \ I s U< U .T' / ~ / f h I, / e~ L- / 1 /,' / / r , .* 4 / I .
- g, .
t I > s D: '- 3 ! , . + . ,i .- ! ' I i FT v- i u . t/ v i ! ., ! .. r + ( y i .r _. . ., . . _ . ~ . . _ i. . i , n ; f. . .u. . . L . . c. c-. i . f , . . , . . .I ; e e .. . . - - . ' .. t : -.,,'I. ,u. ,. . .,. . i .,., / !r, .. .s, .d,. . , i ./. ._... ... _.- . . . . . . ._ . . . . . . . . _ ' . .+ ,i i , .,f,,
- i. . .: a- .. I l t, * .
i 1 'l l L' !. ' i i. . lf I ; . . L , . a. . i i l i / l ,. , l' r.. ( +. i- . .. j .:...... ..- . .._... - . i i /'s) l O'! .fl* ! i " t/ . U -.V. / e' I i?;L cT ; c/ i.. / / -i n ,., e . c ;j
- t. . . , n.r....,,. .
,l
- l. ,
- ' Mi i:i. U !FCMi : ll g L, C. L.. I ,-v T. ui r, ....i,..<, .. . u. 01:. I I I ! I d ! I Cd 1.0 'U L .: . D .t 0 0.0 L0 7.C S.O !.' . 0 ( !:D' LIP.O '. i iJ.l i t il"I .) Fi.pi,' 2 Cf;e'i C/W OU/\t.lTY VS. CO!L!" L ENGTI , FFIOTJ 114 /J'!'ULUC D AT A 4 I i ~E-37a \i \ L' t ?.- I,t l 't ['L ., .-.. l N 1 \ , . . _ . _ _ . , \. .. _ .. ._ -.'.'< _. _. _ . .\ ._ . . . . _ . . . . _ {.. .. . t ' yi ': b o .A. , ?- ' _J \ .i t, \' .r'- r .; r: s s. L' f bi V v. .. .....',.'.j_.. . i ,i ,l C t l .I. \- ( f. b* v', , .y1 /.. ( 1 \ ' 4 m
- p. ..
%l }l . 1 ~\ = . . , l t V 'k.~) } .- (, r !. l l i l4, n : .4 i 1 ea .. r- ) l' .\ i - j . _ < plt t ,. la p . ..._ . . ......i.. .,_..._o,.. ! .-\,... . ! .. 7. . k . l', q g. } L" ! ) '\.I ' l I c ., l l} -- ; ' ., a. , i i .\>- ) jl c --c 5 l l l I 5 u - , -1 f i % i I , ,.s r t i I ,,. " = ' z.* { t , '-j C l '! V1r l \u . W O' !....._i C.: , t _ . . _- l _ ._.. . .,' , i j \ \ j
- m C, l
' } . I. ' \ c-) :. . L' '3 , l Is . L. \ ..l I [' ... ~_ ,3 s-I -S P CG. 't '.} \ .I ;, > ce o- nr -{i n1 ()
- ' :: 3' fy M.
y : , G: l' - ' L;
- D
, , 4 :. ...__.-. .._'__._ . L ...a_ o i .F. - -. L; 9. i t # i C -g i. i c.> n ov *. L' :, r,. 6 i ,, j ,- .,e ,; 3 - - t l << \ 3 4 3W %. - 1 Y " .. i' 1; .l I i O ,! L u a o e, o m sf q {k-. - p e- y n \. L., y* 6: (m - l'
- i. . 1 I; i .
e-W.E G C 0 C; (1 1 ,' O e, ~. , ' ,b
- k. '\
f ~ . . kb ; c, ;,~. C N . O y Jy s i [ o* g, i \ s'. . ( LL 's\ t W1 H 2t'.'y 7 v.' t.u c i l 1 i l l 1 l 1.~ l l 1 E-38 . ; ii c ! t , i i .: . . c.f ,s ! G. g , O i I \. ,, . . i \ 2 i \ i D ' \ \,l c. [
- o O
\' f C 2* I I . r ;-. . _ . . . _ + . . . . _ . . . ... < ) i i ig \ i 1 I, l 3 d 4 I i s I.. C (.s; l '. A \ t i i < \ ) < I~ I \. ! ( ! Ti i, i 1 I' o g , c, s ! I 4 s, . ,,, e '.' > s . f% r J.. ...._..!.. . .. . . . . _ . . . _ . . _ . .. . . . D i ..
- e. ii '
M s l , i . I ! . . , s . I i i \. l , . .. r. ,1 l \ g ., ' t. < i. t .u 1 c , .h .. i ,i i c . ,...a .. r ~ . , _ . . . - - -L........... - i. i i ; i i \. 1 f. .a c .. i ! o l i o n . -' . -g. . e , s. u e l . , 3 . i i . l I Q g . i . s i ' , t ! l \ i I :~ . i [ i.< 9 .'. - I f ' t l . l I i : \ . I i s a . l ' i i \ q t-i , ! \'i . C. I i f g } ; 44. s 9 , l . . s .' ., l l t \' i ". , I l l s s - t , . . . . . . _ . . . _ _ . . . . . ..._.._.. .. _..,l......_._.~_._... s.. '. - I t ", . . i ,.6. i l l l l ,\ C,; E. i: l . \ . Ia f.... i s .
- L .e ; .t.o.
8 l i 6 y 6 e ) 1 f w J S. M. .. i. 6 ' .a. - - - - ....-...=*..a = = . - = = . . . - . a.a====8--- g ,. J q N 'd *. !' - <J c. - i . s, .o . . , ' g u q j o. p g j e .. . l e s . \ b. ' I l ( l \ -. . _ _ . - L _ .. . -.....w.~~.' - . - - - - ~ (gb ' %. ' ). fi l ! J t / f i1 li'D;I4D Od13.G3Ud l ,r--m,.,.------_ l J E-39 1 f W J -.~-....-~....7........- l> o U-(. l '. O l . \ c' \ O C. O c n cc w g P. OO x e . I ' ....C.',..-.... , . - - . - - - - - . . . . - - . - - . _ - . 1 i I f l '\ ; , 6 1 7. ,, g . h l , '*1 ' '[ i Q < i i , \ d (' \ .. o l- = [O s?. l ,i
- . V o :c e C (1
I \ . t I (A, .i i: \ te ( -. ', ! .C i %s ' _. .- . . . . - . ' . - - ~ . - . . _ - --4 (. > l I' \ - d h l , t 6 O -\ 3 -l l - .'e. i i st \ i n_ , l l 7' %y i '\\ ; i (, .,'
- v i
.g i l l, - . , \ ' O.. :. c. l fG f. ( , , \ ,, E V. c. l . . . ._ . . ' ._.'- . '. _; .;.__.. _. .. _.t . _ . _ _ .. ..__._. .. l .. i . - - . if s* 3 g-I s/ <3 a gs ' s. 8 ..v i I ( h l C c >' 4
- a. r.
I l l ' c - . a. -I c4' C(* 1 i ; ;, ,i ,s . u, s, ' C, f .'7' , , f' l i a , ! ; i. ." l I . 8 t , 6 { p s. .'s I 1, < '3, c I I a e = . .'".= ; O I \ 8 f [- ' hee _7__.---. u ( , , l o i- ,! b i - sw l (. ' -) 3 l 4 =I 0 f I (. f. , l . i . s d. Ci , i e i L l * *_ . l l, 3f, C ! . t. ) ,' , i , i n .. .i_.. ._...L _.' ...r.__ . .c. . ._. P, l l t ; , l a . tp.* . i ) . g 1 g I ! Q ! t w I \ i' C a \ l . \ . ].
- e' l l
_.____.__i..._._.... - ._ . O, I De ' 8 ! p i \ w g < i n m. er - ( \ a nn 1 4 ~ ~ ~. \ - . g , g _____________m._-_.._ _ - . _ _ - _ _ _ _ _ _ _ _ _ _ _ _ . . _ . _ _ _ _ - _ _ _ _ _ _ _ - _ k'.** ' h w 5 e e ' 7..,.,_.,._m.. . L. 's I C \ l1 - 4 C. c i' '\ ' \ f \ ." O s'. O ,n. , a. _. . .~.,a. . . . . . _ . . --s,. =S l I 1 \ l Q l M. - I . =\ t I \, \ O
- l. l c.
\ f.*. i ' \s o. L .s. c' ,._ --......_..g. \, .. . _ . . -.. ...--- 3 y' s (* ! \ . C. C - ci *o i \ i- i . g C.. .,. r)
- U
; *(- c, l .\ 7, e-1i ' \ l . i t. , \, i !<- ; o .\.....-.. ( ,. ' . . ( e a .e y - l i
- f. t l
- l (*s ' *
. C' .- l ,6 , s e . g ('j at 4 t a
- f* .
)' . , { ..: 4 ., f., a l (. . ' .. .a y- 5 L , l ,, l f* .' I i . ( s. . a'
- 1. !
- i . ( . .! /
- g l l .%.. +
4 s t - ' s. . l l. *. I u "m i n ' g C. *E L. . . . , . . .'.. . . - . . , . i *. . L. -_ _ . . - - - I g** e ( * ' r. % ! n , , - ., ? <: e. I ! , c ,f S . a .> . i - , g* . # 4 % 1 e 9 ' ( r. . l l # f . v %. ' + { Lsa ) P l g l* g . 6 . l . p .. .. .._. ..--...._. ... . i. C, (. , I ,# t g ',. 'f Q + , .#6 g', , v 6.* til l . 'r C' l* I L'- s s ., l O.. j I ' . t* ., w - ~ . ' .q ~ s C ! L' , ; [ n.
- 1. .
1 .J.'...... .. O * ) C- \ .a l (.i s i i r, , ,1 - J e, t ^ %e h (. U f (i e. - ea ,,: ,.- n. . .,. , ~- w v. u, C. ..
- c. .
c. 9 \ v** . _ _ . \ s 11'9 ?* l l (glj "l!"I'.h X O i d IVi!H WD t.dl.D C 2iDI L i> '.'J I i. "O a i i 1 1 -1 j*~th [ d l [ T, o A +- _., i...s,.r . : 1.p a ;. l p.. . i p , e .g s. .s .. a ,r l .... .i s t . . v I r.s.?, ...i.> *s . ....\,,.. . . . . ..i... 9, . " ll<J:. m- ^ 1 !it'.'!TS , o v.D Ns i: Os. .. e ., s.. .,.. .- _ . i-.-. .c / J. ItP. % m> j p g, a (. l. . . 4.b .. .. [,,:., .. ( -- p, . ! , - \, CRO ': ; '.t!" C Iti .F C '.TE 2;'OT fit - t- I ' TO U: U T. f ' C /> T,' .- - . . - 0 r.' ' ;i :c ' 00'iTh 0 L'_ E ? C .- [ , C , , r, ! r ~... N e'. L g i e , , , s s i l .. \. . \ s .
- t. s . x, p.
p, i 3 I t r s)- {l 4 . \, +
- H. . t i s T
o D. \ . ,I i i . . < ( ;;; n f. . e > [- vs , ' '}}