ML18040B025
| ML18040B025 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/13/1983 |
| From: | Curtis N PENNSYLVANIA POWER & LIGHT CO. |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| PLA-1870, NUDOCS 8310180426 | |
| Download: ML18040B025 (87) | |
Text
SSES-FSAR a ~
Accompli;sh a controlled, orderly and safe initial, core loading b Accomplish a controlled orderly,. and safe initial criticality ancL heatup Conduct low power testing sufficient to ensure that design parameters are satisfied and safety analysis assumptions are correct or conservative Perform a controlled,, orderly,, and safe power ascension 14 2~/ ~0 G'ANIZATION-AND-STAFFING'-
The Superintendent of Plant Susquehanna,, has overall responsibility for the Initial Test Program.. The Plant Staff and Integrated Start'up Group {ISG) conduct the different phases of the test program. Responsibility for the ISG may be delegated to the 'Assistant: Superintendent, of Pla.nt-Outrages In addition to
'he'se basic'or'qanizatronat:units.:the Superintend'e'nC oE Vlant.-
Susquehanna is assistecL by two review organizations,. the Plant Operations. Review: Committee (PORC) and; the Test Review, BoarcL
{TRB) Th.e organizationauthority.. responsibility" an.d: degree of participation. of each of these organizational units during the Initial Test Proqram. are described in the following sections
- 14. 2. 2..1 Plant Staff The Plant Staff consists of the permanent onsite PPGL personnel responsible for the safe operation and proper maintenance of the plant. Chapter 13 describes the Plant Staff organization. This section also establishes responsibilities, reporting relationships,. and minimum qualification requirements for principal Plant Staff supervisory personnel.
The Plant, Staff also includes, the Startup Test Group which. is a..
temporary -group: established to prepar~ for- and implement the Startup- Test'rogram The Startup; Test Group Supervisor reports to the.= Technical. Supervisor and supervises. the activities of the Startup Test'roup . Activities include," preparation and implementation of startup tests;; review ancL analysis of startup test results; preparati'on of startup test reports; and participation in test planning meetings.
The Plant Staff is utilized, to the fullest extent practicable, during the Initial Test Program. Specific responsibilities of the. Plant Staff during the Initial Test Program. are=
14'2-'3'
SSES-ESAB 14~2 11'FST PROGRAM"'CHEDU"CE he Preoperational Test Program- is scheduled for 15 months duration on the Unit 1'nd Common components and for. 12 months duration on the remaining Unit 2 components (see Figure 14 2-4a and 14 2-.4b) .. The subsequent Startup Test'rograms are sch duled for six months on each unit The'reoperationa1 Test Proqram sequential test schedules presehted on Figures, 14 2-4a and. 14. 2-4b offer one possible plan for an; orde ly: and ef ficien proqression 'of the p-ogram., Rhile-these secruences may be preferred,. numerous; altezna;tives exist The- schedule- vill, be updated. periodicalI'.y at he jobsite to.
ze" lect corstruction s+atus, manpower a;vailability,, and'. the.
req:uired test prereq'uisites he safety-related structu"es., systems;. and components vill be preoperationally tes,.ed., The Preoperational Test Procedure" are scheduled to be developed from Sep+ember: 1'977'o J'anua" y 19-79 f'r Unitnd'. from: tuI'y 1982, to July 1983..fqr,.Unit 2 "Rhere electri'caI'm'each'anica1.'"physicaI'r'-'admirr'ist~ativh communication.
exists, be+.ween Unit 2'nd the operating:, Unit 1 the Unit 2.
PreoperationaI'r Acceptance Test wiI:Z, be. divid,ed: into, 2. or. more proced'uresr to faci:lactate.- proper ad'm'inistrative: corrtroL and sched'ulinq. Any'est procedure.-" which.'irrvoXves; ar. interplant
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communica;tian: wilI: contain. the:- suffxx B on eke. procedure number The schedule of Unit 1'nd: Unit 2. S'tartup Tests is presented -'n Figure 14 2-'5. This schedule establishes the required ".est'g as funct'on of tes.. cond'ion.. The test condi+ions are described on Figure 14.2-6. All testing is assigned +o a'pecific test condition for convenier.ce even though some testing, as identified in ficrure 14..2-5 is performed'utside the- bounds of the assigned.
test condition., blot all subtests of a Startup Test are po.rformecl at each assicrned test condition. Startup +esting vill be divided into =hree- 5ajor, Test Phases,. and,. within. the Po~er Ascension Test'. Phase in o distinct test plateaus The testing included in each: ~a"joe T'est Phase and test plateau is. desc-ibed; in: Table
O'-Z-4.." Ev:en though th'is'.'basi*c order: o te'sting is. required tliere;- is--still corrsrderabl~ fl xibility in. sequencing; the staztup, testinq: specified'o be- cond'ucted at" each:. plateau. Detailed..
star+up test'nq, schedulescommensurate with the requirements of this schedule vill be developed't the job site 14,2,12 T'i%DIVIDUAL TEST DESCRIPTIONS The indi:vid'ual. preoperationa.I'ests to be conducted on. safety-related structu-es;systems and. components are lisped. in Table 14 2.-1. for: Unit 1 ancL Table 1'4.2-6 foC Unit 2.'he abstracts of j '.
.2-.21' P'4
SSES-PSA'R'hese preoperational tests are con+.ained in Subsec ion 1'4 2 12
'numerical order 'The Sta "tup Test Program procedures are 1'n:
listed'n Table 1'0 2-3., The abstracts of Startnp T'est procedures are contained in Subsections 14 2..1'2 2 and 14 2 12. 6 for Unit Unit 2, respectively in numerical order. The abstracts 1'nd identify each test by. title and. number,. describe the te t objec+ives.. specify .he test p-e equis'te=, provide a summary descri.'ption of the test method,. and es+ablish the test acceptance criteria..
Un'it 2 preoperational program will be scheduled and. performed in
- a. manner,. What wil3 no+. affect'he safe operatio'n of Unit of the Preoperational A'cceptance T'est will, be. subdivided 1'everal
'nto h and. 8'ests The A'ortion. of'he test will. not affect the safe operation of Uni,". 1'he- B. portion of the Preopeza~ional Tes". is dependent upon an irterface with Un't 1'nd may require an outage on Uni 1'o perform the tes .. Zn addition to Test Rev'iew Boa:rd: aoproval of the Pzeoperational. Test, B desiqnated tests. will require a written Safety. Evaluation submitted and tes-t;pprova3'w the Pla;.nt operations.8~view'ommi,tt:ee All -permanent:
'interface .c'onnect.'one betwee'n.. U'nit, 1-'ant: Unf;t vill be:..: ":
'2 accomplished:; in accordance with SSES Plant Modification Procedure Prior. to performinq." the; B designated. PreoperationaL T'est the Fork A'ctxvity Review Committee" wi 1L be'- briefed on: the.
impact arrd:- equi:rement's. of; the. test;.
1L 2 1'2 1 U'n'it 1'zeogerational Test Procedure Abstracts 125 Vol.t DC System Pzeogezat iona 1 Tes" Test Objective To demonstrate the ability of he 125 Volt dc system to perform he followinq:
The bat+ez'es can endure a complete discharge,. based or.
their ampere- hour rating without exceeding: the'at .ery bank minimum, voltage,. limit (Pe rfozmance Test) n The batteries, can- provide" reliable stood;.'nerq'y to selected load's, i:nd'.Heated'n Ta.ble. 8: 3-'6,. in ..he event of, a: des'n.
base- accid'ent= (S'ervi;ce Test)
The battery: chazqezs can deliver their rated output.
The batte y chazqers can fully chazqe :heir associated batteries from desian minimum charqed s..ate (i.e , after the service tes+) simultaneously pzovidirq. power to the distr.i.bution'anels f'r norma:1 station. loa.ds
SSES-PSAR QP.100'.~1, =- Cold Functional Test Test. 0+b. ective- To Aemonst rate that the plant systems are capable of operating on an integrated basis in normal ann emergency modes to demonstrate that adequate power supplies for the class IE equipment will exist ancL to assure that optimum tap settings have been selected for transformers supplying power from offsite sources to class IE busses.
~rereguisites Required system preoperational tests have been complet'ed and. plant systems are ready for operation on an integrated basis Test method Emergency Core Cool&.q. Systems (RHR 6 Core Spray) are lined up in their normal standby. mod'e The plant electrical system is lined. up per normal. electrical system lineup (For Unit 1 this lineup may be different than the lineup for two unit operation) Loss of coolant accident signals are initiated with and without a loss of offsite power Voltages, and, load.s are adjusted as practical to simulate the anticipated ranges of variations Proper response of lhe'lectrical distributiorr.
wi'll'e
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'sjsteqr -diesel'gener'atora "and 7CCS "pumps verified'Acce tance -criteria;.Systems performance parameters are in t accordance
-mD:- "'
with the appIicabIe, lt desi;gn. documents All, those tests comprising the Unit I'tartup Test Program (Table 14 2-3) are discussed in this section.. For each test a description is provided for test purpose, test prerequisites, test description and statement of test acceptance criteria, where applicable Additions,, deletions, and changes to these discussions are expected to occur as the test program progresses..
Such modification to these discussions will be reflected in amendments to the FSAR In describing the purpose of a test an. attempt is made to identify those operating and safety-oriented characteristics of the;. plant which are .heing,-explored ll
",I applicable, a definition of'-the relevant acceptance
'here-criteria. for the test. is. given. and. is designated. either Level of L'evel'. A. Level 1'riterion normaIIy rela.tes to the value 1'r a process variable assignees in. the design of the plant, component systems or associated equipment. If a Level 1 criterion is not satisfied, the plant will be placed in a suitable hold-condition until resolution is obtained. Tests compatible withapplicable this hold-condition may be continued Following resolution, tests must be: repeated. to verify that the requirements of the Level 1 criterion are: now .satisfied J W p
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SSZS-PSAH
.Level 2 criterion. is associated. with expectations relating to A
the performance of systems, If'. Level 2 criterion is not satisfied,. operating and testi.ng plans would not necessarily be altered,. Investigations of the measurements and of the analytical techniques used for, the predictions would be started.
Zoz transients involving oscillatory response, the criteria are specified. in terms of decay ratio (defined as the ratio of successive maximum amplitudes of the same polarity) . The decay ratio must be less than unity to meet a Level 1 criterion and less than 0-25 to meet Level 2.
/ST-1$ -"Chemical and Radiochemical Test Objectives- The principal objectives of this, test are a) to secuze information on the chemistry and radiochemistry of the reactor coolant and b3 to determine that the sampling equipment, procedures. and analytic techniques are adeguate to supply the data zequired to demonstrate that the chemistry of all parts of the entire reactor system meet specifications and process req uire me nts ~ a h Specific- objectives of the test program inc1ude documentation of radwaste. liguid discharge documentation. of baseline piping radiation. levels determination of steam quality evaluation of the Condensate Polishing system and, evaluation of the Reactor Water Cleanup system Data. for these purposes is secured: from a variety'f sources.- plant operating records,, regular routine coolant analysis, radiochemical measurements of specific nuclides., and special chemical tests Pre~re uisites The required preoperational tests have been completed Instrumentation has been checked or calibrated as appropriate Test method prior to fuel loading, chemical samples are taken to ensure that reactor coolant and puel Pool cooling and cleanup System sample stations are functioning properly and to'etermine initial. concentrations . Additionallysubsequent to fuel load:ing during reactor, heatup, and. at each; ma.jor power level
'change,, am compIete set of" samples are taken. to. verify that all plant sample stations are- functioning properly and'o determine the.. chemical and radiochemical. quality of reactor water and reactor feedwater and'erformance of fi:lters: and demineralizers Ance tance Cribte ia Level 1' Chemical factors defined in the Technical Specifications and Fuel Warranty must be maintained within the limits specified. The activity of liquid effluents must conform to license limitations.. Hater quality must be 3cnown at all times and. should. remain within the guidelines of the Hater Quality S'pecifications
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SSES-PS'AB Te~t~ethod Before the first fuel assembly is taken from the fuel pool and insezted into the reactor core components (fuel support castingsblade guides, control rod drives, etc.) will he installed, tested, and/or verified. This procedure begins with the steps required. to assemble and, load neutron sources, includes the activities necessary to monitor neutron population using specially constzucted fuel loading chambers (PLCs), and culminates vith the insertion of fuel assemblies into the reactor core. Fuel loadinq continues until the coze is fully loaded, verified and zeady to perform subsequent Startup Tests.
Control rod functional tests, suhcriticality checks,, and shutdown margin demonstrations vill be performed. periodically during the loa.di:ng Acc~etance Criteria LeveL T The partially loaded core must be subcritical by at least 0.38%%u delta k/k vith the analytically determined, highest worth rod fully withdrawn
/ST-4+ Full Core Shutdown Margin
~: Te~:Objective-.' The.,purpose- of '.thm-.test is to: demonstrate that .'-.
the reactor wilI be subcritical throughout the first. fuel cycle vith any single- control rod fully vithdrawn II Prege~isites- The follows.ng prerequisites wU.1 be complete prior to performing the fulI core shutdovn. margin test-..
a) The predicted critical rod position is available b) The Standby Liquid Control System is available c) Nuclear instrumentation is available with neutron count rate of at least three counts per second and signal to noise ratio greater than two to one d) High-flux scram tzips are set conservatively low e) Instrumentation has- been checked or calibrated as appropriate.
~e t~ethod'- This test: vill be performed in. the fully loaded core in the xenon-free condition The shutdovn margin. test will be performed by vithdraving the control. zods. from the all-rods-in configuration untiI criticality. is reached. If the highest worth rod will not be withdrawn in sequence, other rods may be withdrawn providing that the reactivity worth is equivalent.. The difference betveen the measured. Keff and the calculated Keff or the in-sequence critical vill be applied to the calculated value to obtain the= true shutdown. margin..
TO 2.-6-3'
,a ', m
SSES-PSAR
~hcce tance. Crltegia Level 1 The shutdown margin of the fully loaded cold (68~P)', xenon-free coze occuring at the most reactive time during the cycle must be at least 0.38'A delta k/k
! with the analytically strongest rod (oz its reactivity equivalent) withdrawn.. If the shutdown margin is measured at some time during the cycle other than the most zeactive time,.
compliance with the above criterion is shown by demonstrating that the shutdo wn margin is 0 38% delta k/k plus an exposure dependent correction factor which corrects the shutdown margin at that time to the minimum shutdown margin.
Level 2 Criticality should occur within +1.0% delta k/k of the predicted critical QS~T-5 " Controi Rod DIive-Sistern
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Test ~Ob ective The objectives of the Control Rod, Drive System test are; a) to demonstrate that the Control Rod Drive {CRD)
System operates properly over the full range of pzimazy coolant temperatuzes and pressures from ambient to operating, and b) to determine the initial operating. chyzacteristci'cs of the entize CRD completed'est--method--
The CHD'ests perfozmed. dazing the staztup test program are designed as an extension of the tests performed during: the.- pzeoperational CRD system tests verified that a.ll control rod drives operate properly Thus, after it when is installed, they are tested periodically during heatup to assure that there is no significant binding caused by thermal expansion of the core components A list of all control rod drive tests to be performed. durinq startup testing is given in Table 14 2-5 Acceptance Criteria Level 1 Each CRD must have a normal withdraw time gzeater than oz equal to 40 seconds.,
The mean scram time of all. operable CRDs must not exceed the values: specified in. the- plant technical-.specifications . {Scram, time is measured: from. the time the pilot scraar va1ve solenords are deenezqized )
The mean scram time of the three* fastest CRDs- in a two by two array. must not exceed. the values specified. in the plant technical specifications. {Scram time is measured from the time the pilot scram solenoids are deenergized)
Level. 2 Each CRD must have a norma insert speed of 3.0 a 0.6 inches* per second indicated by- a; full 12-foot stroke in 40 to. 60 seconds. With respect to the control rod drive friction tests, if the differential pzessuze variation exceeds 1'5 psid. for a a
l 14 2-64
SSES-ZS AR.
Unfortunately the decay heat load is insignificant during the startup test. period Use of this mode. with lov core exposure zesults in exceeding the 1000P/hr cooldovn rate of the vessel ),
The shutdown coolinq mode vill be demonstrated after a trip or a cooldovn from Test Condition 6 The RHR system steam condensing mode is used to condense steam while the reactor is isolated from the main condenser and reactor vessel water level is being maintained by RCIC. This test vill demonstrate system operability and stability Acct tance.Criteria Tevel 1 The transient response of any system-related variable to any'est input must not diverge Xeyel-2.- The RHR system shall. be capable of operating in the steam condensing suppression pool cooling and shutdovn cooling modes at the heat exchanger capacities indicated on the process diagrams. Both simultaneous. operation of RHR'oops and single loop operation shall be tested in the steam condensing and shutdovn cooling modes Each RHR loop shall be tested independently in the suppression pooL cooling mode System-zelated'~variables. may c5nta'in'sc."illatory modes" o'f response, Zn:
these cases the decay ratio for each. controlled mode of response must be less; than or equal, ta 0 25 t The tame to" place the RHR'eat erchangers, in the steam condensing mode with- the- RCIC usinq'he heat exchanger condensate flow for suction. shall average= one half hour or less
/ST~9 Wat~eLevel Measurement Tegt~~bectives The objectives of this test are to determine actual reference leq temperature and recalibrate instruments necessary and'o verify'onsistent response of the upset range, if nazzov range and vide range level instrumentation.
Prereguisites The required preoperational tests have. been completed. All system instrumentation is installed. and, calibrated Test-rlethod -- At rated. temperature and.. pressure undez steady state cond'itions;.. the reference Ieg temperature vill be measured.
and'. compared. to the value assumed during initial calibration If the difference of the two temperatures exceed the Acceptance Criteria, then. the instruments vil1 be recalibrated using the measured value Data vill be recorded at rated. temperature and pressuze and at steady state conditions to verify consistency and proper calibration of reactor vessel level instrumentation.
Acce tance Czite ia T.evel 1 Not applicable
SSES-PSAR Level. 2. The difference between. the actual zeference leg temperature (s) and. the- value (s) assumed. during .calibration 'shall be less than that amount which will result in a scale end point error of 1% of the instrument span for each range The Narrow Range Level indicators should aqree within %1 5 inches of their avezaqe reading..
The Ride and Upset Range Level indicators should agree within k6 inches of their average reading
/ST--10~If Perf o~mance M
Test~Oh ectives.- The objective- of this test is to adjust. the Intermediate Range Monitor System to obtain the desired overlap with the SRM'nd APRM systems Prese uisites The required preoperational tests have been completed.
Test. Method- -: Initially .the IRM" system is .set.,during the
'Pdagr'as'--SRH" XRH'.6nd'RH.-APHM" first time sufficient neutron flux conditions arise
'Preooperati.ohal'.Test the overlap"is"'erified After the A'PRE caLibration the'RM: gains wHZ. be adjusted as t .necessary: to optima;ze the'RM: overlap with. the Ac~ce ~tqcy~Ctezia--
so that. overlap with the LeveZ Calibration SRMs 1' Each IRM.
and A'PRMs-channel'ust is assured'ST-111'PRM SRM's and APRMs be adjusted Test Objectives The objective of this test is to calibrate the Local Power Range Monitorinq System grer~e uisites The required preoperational tests have been completed Instrumentation for calibration has been checked..
Test Method- The LPRM channels will be calibrated to make the LPRM'eadings proportional to the neutron flux in the water gap at the chamber elevate.on.', Prior'o: this. calibration. LPRM response to control rod movement: is, verified-'alibration.
factors wi11- be= obtained 'through: the use= of. either an off-line or.
- a. process. computer calculation that relates the LPRM reading to average fuel assembly- power at the" chamber height.
Acceptance Criteria Level 1 --Not applicable Level 2 Each LPRM will be within 10% of its calculated. value.
~ST-12) . ApRK Cali~batian
SSES-PSAR'umps to avoid. coolant temperature stratification in the reactor pressure vessel bottom head region .
Pre~re uisites -- The zequired preoperational tests have been completed.. System instrumentation has been calibrated.
Test Method During initial heatup while at hot standby conditions, the bottom drain line temperature, recirculation loop suction temperature and applicable reactor parameters are monitored as the recirculation flow is slowly lowered to minimum stable flow Utilizing this data it can be determined whether coolant temperature stratification occurs when the recirculation pumps are on and.
prevent it if so, what minimum recirculation flow will Monitoring the preceedinq information during. planned pump trips will determine if temperature stratification occurs in the idle recirculation loops or in the lower plenum when one or more loops are inactive Acceptance criteria level r The reactor recirculation. pueps
'h'al'1 'riot'.be started .ho'r flair. irrcreaseri unless" the 'cooI ant '"
temperatures between the steam dome and. bottom head drain are within 105oP The recirculation pump. in an. i.die Xoop, must not be started', unless the loop suction temperature: is. within, 50~7 of the active loop T.evel- 2 Not App1icable T~et Objectives The purposes of this test are to demonstrate that reactor recirculation, main steam inside containment,. and those piping systems identified, in Table 3.9-33 respond to thermal expansion consistent with stress analysis results., (Note
.that this test now includes piping previously contained in ST-38 )
11 calibrated Test Method- Kanger. positions. and. locations of piping in the Nuclear Steam SuppXy System and piping systems identified in Table 3..9-33 inside and. outside the reactor drywell are recorded prior to initial heatup and after a planned cold shutdown.;
During intitia'1 heatup, a visual inspection is made at an intermediate reactor water temperature to assure components are free to move as designed Adjustments are made as necessary.
Devices: for measuring continuous Pipe deflections're mounted. on main. steam, recirculation and. other selected lines Motion" a uring heatup; is compared with calculated. valuer.
SSES-PS AR Acc~etance Criteria Level.1-- There shall be no obstructions which will"interfere vith the thermal expansion of the main steam and recirculation piping. systems.. Piping systems identified in Table 3 9-33 will not be restrained against thermal expansion except by design intent Hangers on piping systems identified on Table 3..9-33 shall not be bottomed. out or have the spring fully stretched Snubbers on piping systems identified in Table 3.9-33 shall not become extended or 'compressed.to the limits of their total travel.
The measured displacements at the established transducer locations on the- main steam and recirculation systems shall. not exceed,'he allovable values calculated for 'the specific points level 2.-- The measured displacements at the established transducer locations on the main steam and recirculation systems shall not exceed the expected values calcu1ated for the specific points The measured displacements at the established transducer locations on the piping systems identified in Table 3.9-33 shall be. within the acceptable. range calculated. for the specific
'p'ants Hangers on piping, systems: identified. in. Table 3 9-33 shal1 be in their operating range ST-18 "-TIP Uncertaintg-Tegt Obgectives The objective of this test is to determine the uncertainty of the TIP system readings Prereguisites System installation is completed and required preoperational tests are completed and verified. Instrumentation has been calibrated. and, installed..
Test Method The TIP uncertainty consists of a random noise component and a geometric component, the geometric component being due to variation in the water gap geometry and TIP tube orientati'on from TIP location to location Measurement of these components is obtained by; taking; repetitive TIP readings at a.
szngle TIP'ocation, and. by analyzing pairs of TIP readings. taken at TIP Locations which are .symmetrical about the core diagonal of fuel loading and'ontrol rod symmetry The- random noise uncertainty is. determined. from. successive TIP runs made at the common location (32-33) vith each of the TIP machines making six runs at index position 10.. The TIP data will he obtained by simultaneous operation of the Process computer OD-2 program which provides 20 nodal TIP values for. each TIP traverse The standard deviation of the random noise is derived hy taking the sguare root of the average of the variances at, nodal levels.- 5'hrough. 22 where. the. noda3'ariance is obtained 1'4' 75
S SE S-FSAR MLHGR MCPR MAPLHGR Prior to the verificatiou of the Process Computer in ST-13, an independent method, will be used to calculate these parameters After the successful completion of ST-13, the process computer will be used Acceptance Criteria Level 1 The Maximum Linear Heat Generation Rate (MLHGR) of any rod during steady-state conditions shall not exceed. the limit specified; by the Plant Technical Specifications The steady-state Minimum Critical Power Ratio (MCPR) shall not exceed the limits specified by. the Plant Technical Specifications Tge Maximum Average Planar Linear Heat Generation Ryte .(MAPLHGR)
's'hall.'ot'.exceed; the:. Rim'its 'specifx,'e6 by 'the Pla'nt 'Techni cal Specifications Steady-state: reactor power shaX1. be Ximited to the rated. MRT'nd values on or below the Licensed'naZyticaIIy determined; powez-fI;ow tine Level 2 Not applicable
/ST-20) - Steam production Verification 1'es~tob 'ective The objective of this test is to demonstrate that the NSSS's providing steam sufficient to satisfy all appropriate warranties.
prereguasites Required preoperational tests have been completed. All required instrumentation is installed and calibrated Test-Method-- A. NSSS'team output performance test of YOO hours.
of continuous: operation at the. warranted'team output will be performed'.
Acct tance- Criteria---Level-1 The average reactor core thermal power {CTP) sha13 not exceed 3293 MHt.
The Maximum Average Planar Ratio {MAPRAT) shall be less than or equal to 1.0 The Maximum: Fraction. of Timiting Critical Power Ratio (MFLCPB) shaI1 be. less, than or equal to 1'
SSES-CESAR The Maximum Fraction of: Limiting Power Density (EPLPD) shall be less than or equal to 1 .0.
Level 2 The NSSS shall be capable of supplying 13,483,000 pounds per hour of steam of not. less than 99 .7% quality at a pressure of 985 psia at the outlet of the second main steam line isolation valve as based upon a fina1 feedwater temperature of 3830F measured as near the reactor pressure vessel as practicable, and a control rod dzive feed flow of 32,000 pounds per hour at 80~F QST-.2~1- -Core foyer-Void Mode Re~souse
~Test-Ob ectives-- The objective of this test is to verify the stability. of the core paver-void dynamic response Pre~eguisites The required preoperational tests have been completed. Instrumentation has been caIibrated Test Method The core power void loop .mode that results from a
~ combination of the neutron. kinetics and coze. thermal hydraulic dj'nemmies ih lea'st stable'ear 'the natur'a1 circulati:ot'n end'. oZ the rated. 100 percent power rod line., A fast change in the reactivity balance. is- obtained by. moving a. very high worth rod.
on1y 1 or 2'otches'nd'. by simulating a. failuze of the. pressure regulator hc~ce tance Criteria Level 1 The transient response of any system related variable to any test. input must not diverge Level 2 Not applicable.
1ST-221= Pressure Requlator Test O~bectives=- The objectives- of this test are to demonstrate the takeover capability of the backup pressure regulator upon failure of the controlling pressure regulator and to demonstrate smooth pressure control transition between the control valves and bypass valves when reactor steam generation exceeds steam flow.
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used: by. the. turbine I
v Prereceuisites;. The -required'reoperational tests have been completed ., Instrumentation has been checked or calibra.ted's appropriate.
Test Method The pressure set point will be decreased rapidly and later increased rapidly by about 10 psi and the response of the system vill. be measured in each case., It is desirable to accomplish the set point chanqe in less than 1 second.= At specified test conditions. the load. limit setpoint vill be set so that the transient is handled. by contro1 valves,. bypass valves and both The backup regulator vill be tested by simulating a 14 2-78 v
SSE S-ES A R t level setpoint will be ma.de. to demonstrate proper response and operability of the feedwater system at low reactor power At Test Conditions 2, 3 and 6, with one feedvater. pump in manual and the others in auto, a +5% change in the manually controlled feed pump will be made.. The response of the feedvater system to these steps vill be analyzed and compared, to the applicable acceptance criteria. The recirculation system vill, be in manual for these tests At Test Conditions 1, 2,, 3, 0, 5 8 6, with the recirculation system in manual, +5 inch changes in the water level setpoint vill be made to demonstrate proper response and stability- of the feedvater. system At approximately 80% to 90%* power vith core flow. near 100% of rated failure- of extraction steam valves to one of the feedwater heater trains is accomplished by closing the heater train steam inlet isolation valves which last three stages of that vill train.
isolate extraction steam to the Recordings of the transient vill be analyzed and compared criteria to the predicted. response and acceptance
' .. 't.demonstrateCon'dtion Test, I
8"one'feedvater the capability to avoid pump. vii'3 'be'ripped'-t6:'-
a scram and prevent a low reactor water leveL trip" due to the loss of one feedvater pump A, maximum feedvater runout. capabi1ity test. viiI be dane to demonstrate that the actuaL capability is compatible with licensinq assumptions Acceptance-Criteria Level 1 The transient response of any level control system-related variable to any test input must not diverqe.
Eor the feedvater heater loss test,. the maximum feedwater temperature decrease due to a single failure case must be less than or equal to 100oE. The resultant MCPR must be greater than the fuel thermal safety limit.
The increase in heat flux cannot exceed the predicted Level 2 value. by more, than 2% . The. predicted value. vf.l'1 be based'n the actua3 test, valuer. of feedwater temperature change and. pover leve.I The feedvater flow runout capability must not- exceed. the assumed value in the PSAR Teyel 2-- Level control system-related variables may contain oscillatory modes of response. In these cases, the decay ratio for each controlled mode of response must he less than or equal to,0 25
~'
0 2-8Q
Closurei time for any MSIV'hall not be less than 3 0 seconds Peedwater controL settings must prevent. flooding the main steam lines during the full isolation test..
The time delay between the close initiation signal and the extrapolated initial valve movement from 100% open for any MSIV shall be less than or equal to 0 5 seconds..
Reve)-2 The positive change in vessel dome pressuze occurring within the fizst 30 seconds after the- closure of all MSIVs must not exceed the predicted values Predicted values will be referenced to actual test conditions of initiaL power level scram timing and dome pressure and, will use beginning of life nuclear. data The positive change in heat flux occurring within the first 30 seconds after the closure of all. MSEVs must not exceed the predicted values Predicted values will be referenced. to actual test conditions of initial power level, and dome pressure and
.will use beginning. of .life nuclear, data If water leveL reaches EeveL, 2 setpoint during the MSIV full closure test RCXC .shalL. automaticalLy, initiate and: reach rated fIow During, the MSIV full, closure test the relief valves must reclose properly (without any'etectable leakage) following the pressure transient During full closure of individual MSXVs, peak vessel dome pressure must remain at least 10 psi below the scram setpoint During full closure of individual MSIVs,, pea'k neutron flux must remain at least 7.5% below its scram setpoint.
During full closure of individual MSZVs, steam flow in individual lines must remain at least 10% below the high flow isolation trip setpoint Daring, fully closure of indivf.dual HSIVs. the simulated. heat flux must remain at least 5%,less than.* z.ts flow. biased scram" setpoint
/ST-$ 6) --.gelieZ Valves Test Qbgectiyes The objectives of this test are to verify that the relief valves function properly, reseat properly after operation and contain no major blockages in the relief valve discharge piping..
2 completed. Instrumentation. has been checked or calibrated as'
SSES-CESAR'ppropriate Factory test results- on SRV'low and. operating; times have been. re vie ved Test Bathed Testi:ng d'one at low reactor pressure, in conjunction with plant surveillance testing, consists of cycling each relief valve to vezify proper operation The transient monitoring system, will be used, to record the results of this test The data collected vill compare the operation of individual, relief valves against the operation'of all relief valves.. During relief valve operation core povez and therefoze steam generation rate is maintained constant The pressure control system wi11 close the bypass valves. an amount pzopoztionaI. to the relief valve steam flow to maintain constant reactor. pressure 'Chir. bypass: valve motion. vill. be moni;tored and a comparison of the- response for each: relief valve operation vill be made Zf differences-. exist,, i't could suggest a. partial obstruction of the relief vaI:.ve oz its tailpipe Tailpipe temperature vill be recorded to verify'he relief valve has proper1y reseated., Reactor variables vill also be recorded: to verify system stability during opening and closing each relief yalve
~ ~ V.
C ~
N
~ N ~
Testing. done at. rated reactor. pressure consists of manually operating:.each: reli;ef. va3've. at rated;. reactor pressuze The d'eczease i:n. Haiin. Generator output viXL.be. monitored': during,'he operation of'ach. reXxe& valve'- to- provMe- an; indication: of relief valve flow.- By.. comparison of the generator. output. response for.
each. zeli:ef'alve operation any flov: obstruction in the va1ve or its; tailpipe can be identified'ach. valve vill. be opened. for approximately 10 second's to allow for variables to stabilize Reactor variables vill also be recorded to verify system stability during opening and closing each relief valve.
Acc~e~ce- ~Citegig Level 1 There should be a positive indication of steam discharge during the- manual actuation of each valve Level Pressure contro3 system-related.- variables may: contain oscxllatozy",modes: oC response'='.; Zn, these. cases the'.decay ratio .
for each. controlI'ed'mode-.of'.response, must.-be less than:,or egual 0'$ -', '- '.'. ';,',"' "'**',
to- ,"
N The temperature- measured:.by thermocoupZes. on the discharge s'i'de of the vaNlves shalL return", to. vi thin 10o'F of the temperature recorded before the valve vas opened.
During the lov .pressure functional tests, the change in bypass va3ve position for each SRV opening, shalI be greater than or equal -to a., value corresponding; to the. avera.ge change minus 10% of; one bypass. valve I N
/ N a
a I
.~
'I N N
J l
SSES-CESAR s d)i The positive change in simulated heat flux shall not exceed the Tevel 2 criteria by more than 2% of rated value..
The two pump drive flow coastdown transient, during the first three seconds of an RPT trip,, must fal1 within the specified limits Level '2 a) There 'shall be no MSIV closure in. the first 3 minutes of'he transient and operator action shall not be reguired in that period to avoid. the MSXV trip b) The positive change in vessel dome pressure and in simulated heat flux which occur within the first 30 seconds after. the initiation of either generator or.
turbine trip must not exceed. the predicted values
{Predicted values will be referenced. to actual test cond'.tions 'of initial. power level dome pressure,. -scram timing,. 'and'he"tim'b.fzom the>'start. dftop/controls valve motion to start of control rod motion, and will use beginning of life nuclear data )
Pox the Generator trip within the bypass. valves capacity {initia1 thermal power values. less than. or egual to 25 percent of; rated) the reactor shall not scram The Total Delay from the initiation of a Turbine Stop Valve Closure or Turbine Control Valve Past Closure to complete suppression of the Electric Azc between the fully open contacts of the Recirculation Pump Trip
{RPT) Breaker shall be less than 175 milliseconds.
Recirculation pump trip, HPCI and RCIC starts shall not be initiated from a low reactor water. level.
Feedwater leveI". control. shal1 avoid'he:- loss of feedwater flow due to a. high; 1'eveL (TS) trip.,
1 ST-28 --Shutdown from-Outside the Main Control Room Test Objective. The objective of this test, is to demonstrate that the reactor can be shutdown, maintained in a hot shutdown condition, and cooled down from outside the main contzol zoom.
Also, the adequacy of the Emergency Operating Procedures will be verified 14 2-86 m
SSES-PSAR prere<euisites- The reoulred preoperational tests have been completed . Instrumentation has been. checked or calibrated as appropriate Test Method Phile operating at approximately 20% power synchronized to the grid vith normal electrical system alignment the reactor vill be scrammed and the MSZV's will. be closed from control The control room vill then be inside the main room evacuated, and reactor level and pressure vill be controlled from outside the main control room. The Shutdovn Cooling mode of RHR wiLL be placed into service vith cooling vater supplied from the ultimate heat sink During this demonstration, some supervisory and operating personnel vill remain in the control room to if protect non-safety-related. equipment from unnecessary damage conditions arise and to assume control of the plant if conditions.
varrant A test vill be run to d'emonstrate that the reactor can be scrammed and isolated from outside the control room Acceptance Criteria Level 1' Not applicable Ievel 2- During- a simulated control room evacuation, the reactor must .be brought to the poin't where'cooldCiw'n is initiate'd-ancT ~
control, and the reactor vessel pressure and vater level 'nder are controlled using equipment, and. controls outside the control room . The test is deemed successfu.l vhen; reactor pressure is less than 98 psig (permissive. setpoint). a:nd.'he RHH shutdown mode has- been put in. operation 'ooling The reactor must. be capable of being scrammed and isolated from outside the control room.
/ST-29) -
Recirculation Plow Control System The objectives of this test are-.
a) To demonstrate the flow control capability of the plant over the entire pump speed range including individual local manual and combined Master Manual Operation b). 'o determine arethatset.allforelectrica1 4
compensators d'esired system performance a;nd'ontroUers and stability completed.
All instrumentation has been calibzated.
Test method-- At Test. Conditions 2, 3, 5 and 6, the stability. of the recirculation flow control system..is demonstrated by performing step changes in recizculatxon pump speed. This s
J 10 '2.-8T e
SSES-FSAR testing. is done in individual local manual at Test Conditions 2 and 5 and, in combined. Master Manual operation at Test Conditions 3 and 6 to demonstrate operability and. stability Acc~etance Criteria Tevel 1 The transient response of any system-related variable to any test input must not. diverge Level 2- h scram shall not occur due to recirculation floe control maneuvers The APRE neutron flux trip avoidance margin shall be greater than or equal to 7;.5% and the simulated heat flux trip avoidance margin shalL be greater than or egua1 to 5% when the power maneuver effects- aze extrapolated to those that would.- occur along the f00% rated. rod. Xine-The decay ratio of any oscillatory controlled variable must be less than or equal to 0 25 Steady. state limit cycles {if any) shall not produce turbine steam flov variations greater than + 5% of rated. steam flov
/ST-S0~ReciRc~ulat o ~sstem
~Te ~tob ectives The. objecti.ves of.-this test are=
t a Obtain. recirculation system" performance data during.
pump trip,, flow coastd'own and'ump restart
- b. Verify that the feedwater control system can satisfactorily control water level without a resulting turbine trip and associated scram.
c.. Record and verify acceptable pezfczmance of the recirculation two pump circuit trip sytem.
d., yerify the adequacy of the recirculation runback to mitigate a scram
- e. -
Verify that no recirculation system cavitation vill occur, in. the operable region. of the power-flov map .,
preiecruisites. The required preoperational tests have been completed Instrumentation has been checked or calibrated as appropriate.
Test Method Single recirculation pump tzips vill be made at Test Condition {TC) 3 and TC-6. These trips will be initiated by tripping the M-G Set Drive Motor Breaker. from the control room.
Reactor parameters vill be recorded. during.. the tzansient and analyzed to veri;fy'on-6'divergence of oscillatory responses, adequate margins to RPS'cram. set points and ca:pability oK the v p 't "14 2-S8:
S SE S'-ZS'A R feedvater system to prevent a. high level trip The capability to restart the recirc pump at a. high pover level will also be demonstrated At TC-3, both recirculation pumps RPT bzeaker's will be simultaneously tripped using a temporarily installed test svitch The data gathered will be used to demonstrate acceptable pump coastdown performance prior to high power. turbine trips and generator load rejects Appropriate conditions will be simulated at TC-3 to demonstrate the proper operation of the recizculation pump runback circuits.
This is done prior to an actual planned feed pump trip at rated pover Both the jet pumps and the recirculation. pumps will cavitate at.
conditions of hrqh flow and lov pover where NPSH demands are high and little feedvater subcooling occurs Hovevez, the zecirculation flow vill automatically runback upon sensing a decrease in feedvater flov The maximum recirculation flov is limited. by appropriate stops which will run back the recirculation flow from the possible cavitation zegion At TC-3, it .vill be-where
" Operation verified. that these limits are sufficient to .prevent
'zecir'culltion pum'p'r jet puipp'avitation occurs';."
A'cce~ta ce. C i eria L'eveI I The response of. any leve1 related.
variables during. a; single pump- trip. must not. d'iverge.
The two pump d'rive flow coastdovn transient, during the first of an RPT trip, must fall vithin the specified bounds. 3'econds Level 2 The reactor shall not scram during the one pump trip..
The APRM margin to avoid a scram shall be at least 7.5% during the one pump trip recovery The reactor vater level margin to avoid a high level trip shall be at least 3.0 inches durinq the one pump trip Peak simulated heat flux must zemain. at least 5% below its flov biased scram. value.
Runback. Logic: shaI3'ave. settings adequate to prevent recirculation. pump operation rn areas: of potential cavitation The recircuIatiorr pumps shall runback upon a trip of the runback circuit
)ST-31)= Loss of Turbine-Generator and Offsite Power Test O~b- ecti~vs- The objectives of this test are to demonstrate that the required', safety systems vill initiate and function, properly. without manual'ssistance the electrical distribution and! diesel. generator systems vill function, properly and. the HPCX TQ'-.89
SSES-FSAR and/or RCIC systems will maintain. water level if necessary during a simultaneous loss of .the. main turbine-generator and offsite power The reguired, preoperational tests have been Prize uisitesInstrumentation completed has been checked or calibrated as appropriate
'fest ethod Vith the unit synchronized to the grid at approximately 30% power, the main turbine-generator will be manually tripped coincident with a manual trip of the unit's offsite power source breaker both trips initiated from the control room To ensure a full simulation of the loss of all offsite power to Unit 1'uring Unit 1 testing all. Unit 1'nd Common loads will be transferred to Unit 1 Auxiliary and Startup Busses and appropriate breakers racked out to prevent automatic-transfer of the loads to Unit 2 sources..
Reactor- water level the operation of safety systems will be monitored to verify that the acceptance criteria are satisfied.
~
checked.'nd The proper response of the electrical distribution system will be The. loss. of offsite- power condition vill be maintained for at least 30 mf;nutes to demonstrate. that". necessary equipment controls,. ance indication are: available following station blackout to remove decay heat from the, coze using; only'mergency power supplies and distribution system Acceptance Criteria Level 1 All safety systems, such as the Reactor Protection System,. the diesel-generator, RCIC and HPCI must function prope ly without manual assistance, and HPCI and/or RCIC system action, if necessary, shall keep the reactor water level above the initiation. level of Core Spray LPCI and ADS Levee/- The temperature measured. by the thermocouples on the discharge side of any SRV that actuated. shall return to within IO~F of the temperature recorded before the valve opened Permanent instrumentation for. reactor power, reactor pressure, wa.ter level, .control'od; position suppression pool, temperature high pressure coolant i.ngection,, (HPCI)" and reactor core isolation cooling (RCIC). shaZI: be=demonstrated, operable following." re-energization'f the. SkV busses by the diesel. generators
/ST~3 ~Containment Atmosphere and. Wain. Steam Tunnel CoolincC.
Test Objective The objective of this test is to verify the ability of the drywell coolers/recirculation fans and the reactor building portion. of the main steam. tunnel coolers to maintain design., conditions in the drywell and reactor building portion. of the mainsteam tunnel respectively cLuring operating. conditions and post. scram conditions This: test also demonstrates that
SSES'-PS AR containment main steamline penetrations do not overheat adjacent concrete.
Prer~e uisites The zequired preoperational tests have been completed. Instrumentation has been checked or calibrated as appropriate Test Method During heatup at test conditions 2 and 6, and
.following a planned scram from 100'%ower, data will be taken to ascertain that the containment atmospheric conditions are within design limits.
Acceptance Criteria level 1 not applicable Level 2 The general. drywell area is maintained at an. average temperature less than. or equa1 to 135~F with maximum local temperature not to exceed 1500F The area beneath the reactor pressure vessel is maintained at an average temperature less than or equal to 135oP, maximum local temperature not to exceed 165~F, with minimum local temperature.
above 100~F The area- around the recirculation pump motors is maintained; at an-average temperature less than; oz." egual. to 'l28oF, with; maximum local temperature not to exceed. 135oF r The inside base of the shield. wal'1 izr the- RPV'kirt area is maintained at temperatures g'reater than, 100oF The reactor building portion of the mainsteam pipeway is maintained at oz below 1200F.,
The concrete temperature surroundinq the main steamlime penetrations is maintained at less than 2000F.
QS1~33 piSincC Steady State Vibration Test-O~b- ectives- The objectives of this test is to demonstrate that-..steady state: vibration. levels on,reactor recirculation, main steam, inside- containment and. those piping systems= identified i;n Table. 3 9-33're within acceptable- &mits (Note that; this; test now includes piping previously..contained. in. ST-40 AXso note that. dynamic'ransient; vibration testing previously contained. i.n this test have been. merged. into ST'-39 )
pgeg~equisites Instrumentation has been installed and calibrated.
+est.. Method- Devices for measuring continuous vibration are mounted on mai:n steam lines, recirculation lines and lines of
SSES-PSAR'ystems identified.'n Table 3.9-33 as- applicable, and. vibration during steady state operation is compared with calculated values peak) or each. remotely monitored point on the main steam inside containment and reactor. recircula.tion lines shall not exceed the allowable va.lue for that point Level 2 The measured amplitude (peak to peak) of each remotely monitored point on the main steam inside containment and reactor recirculation lines shall not exceed the expected value for that poi.nt The vibratory response of non;.remote3.y monitored systems or portions oZ systems indentifiecL in Table 3,9-33 shall be'-jud'ged to be within: acceptable I.imits by a gualified test engineer.
The maximum -measured amplitude of the piping response for each remotely monitored. point, on systems identified in Table 3.9-33 shall not exceed the acceptable value foz that point.
/ST-3~4= Consol Bod .S~euence Exchan'cue (This test, number was. previously assigned to the. RPV'nternals Vibration test which: is. now. performed during the PreoperationaL Test Program The: test description for the'PV" Internals Vibration, test, is. now: in, TP2 t6 which. foLIows the abstract for.
pal.T' Tese Oh~ective The objective of this test is to perform a representative sequence exchange of control rod patterns at the power level at which such excha.nges will be done during plant operation and demonstrate that core limits and PCIONR threshold limits will not be exceeded Prese uisites Instrumentation has been checked oz. calibrated as appro'priat e Test +ethos The control rod. seguence exchange begins. on the design. flow control line. with coze fIow near minimum- ControL rod's wilI: be" inserted's necessary to: increase, the, margin'; to local coze thermal limits ' Core power"zs maintained'bove the low power setpoint of the- Rod. Forth Minimizer and Rod Seguence Control System and below the power which. wilX keep fuel assembly nodal power at the PCIOHR threshhold. The exchange is. performed in accordance with the plant operating procedure RE-TP-009. Data taken during the exchange will be reviewed to verify that the Acceptance Criteria were satisfied..
l Acceptance-Criteria Level Completion of the exchange of rod pattern for. the complimentary pattern with continual one
'T0. 2-92
SSES-PSAR Acce~ta ce-C ite ia-- Components: perform: in accordance'with aPPXicable design documents
~ ~
14 .2'2..6 - Unit 2'tarts Test Program Procedure Abstracts kll those tests comprising the 10 2-3) are discussed. in. this Unit ? Startup Test Program,. (Table section. For. each test a description. is provided for test purpose test. prerequisites,.
test description and statement of test acceptance criteria, vhere applicable . Additions, deletions,. and, changes to these discussions aze expected. to occur a: s; the test" program progresses Such: mod'ification to. these. discussions wi11 be reflected in:
amendments to the PSAR Xn describing, the purpose of a'. test an, attempt is. made 'to identify'hose operating and. safety-oriented characteristics of the plant which are, being; explored .
Where applicable a. definition of the relevant acceptance criteria. f'z the test 1"is given. and is designated either Levei o'r Eever 2: X. Level criterion'ox."mali'elates 'to the. value:.of.
- a. process: variable assigned, in:. the design. of the plant,. component.
systems; or assocxateK; eguzpment, TZ a Kevei.: 1: criterion, is: not.
sa<sHeK the; pImt; sxlE; he pZ'acecE xn e suitable halL-:con&;Man-untiL resolutxoz Ss obtaxned" Tests compatible with thxs. halL-.
cond'f tion. mav" be- continued Fo1lovf nq: resolution applicable tests: must be. repeated'o-. verify that the requizements of the:
Level 1'riterion are now" satisQ'ed.'
Level 2 criterion is associated vith expectations relating to the performance= of sYstems Zf a Level 2 criterion is not satisfied, operating anL testing plans would. not necessarily be altered Investigations of the measurements and of'he analytical. techniques used f'r the predictions mould be started.
For- transients involving. oscillatorY response,. the criteria. are specified in. terms'X decay ratio (d'efined's the ratio of successive maxi,mum: ampXitekes'f the. same polarity,),,'Zhe.- decay.
.-- -*'.-ratio" must.,he. less than."unity" ta,meet a...:Eevei'.1."'criterion.. and; less. tham O.'5= to- meet LeveL 2'
~ST'- - Chemi:,ca "-
d Biochemical;,
demonstrate
- *h .:.,-
that the chemistrY
'pa, of'll parts
'P of'he entire reactor sYstem meet specifications and process requirements.
Specific- objectives oP the test program include documentation of radvaste liquid., discharge evaluation. of the Condensate; Polishing system and eyaluatian of the- Reactor Rater. Cleanup system Data for these purposes is: secuzed'zom; C
a variety of sources= plant,
SSZ S-F S'AR operatizq records,, regular routine coolant: analysis,.
radiochemical measurements of specific nuclides and special chemical tests Prer~e uisites The required preoperational tests have been completed.. Instrumentation has been checked or calibrated as appropriate Test+ethod Prior to fuel loading chemical samples are taken to ensure that reactor coolant and Fuel Pool Cooling and Cleanup System'sample stations are functioning properly and to determine initial concentrations.. Additionally subsequent to fuel loading,. during reactor heatup,. and at each. major. power level change, samples, are: taken, to- determine the chemical'. and.
radiochemical quality of reactor water and: zeactor feedwater Acceptance Criteria Level 1 Chemical factors defined in the Technical Specifications and Fuel Warranty must be maintained within the limits specified.. The activity-'of liquid effluents must conform to license limitations Water quality must be known at all'imes and. should remain.within. the guidelines of .the Water Quality Specif icat'ions.
L'evel 2' Not applicabl'e
~ST'-2}- Hadiatian Measnreaents-Test Objectives.- The obj'ectives: of this. test are (a) to determine the background radiation levels in: the plant environs prior to operation for base data on. activity buildup and (b) to monitor radiation at selected power levels to assure the protection of personnel. during plant operation.
g~rer guisites The required preoperational tests have been completed Method A survey
'est of natural background radiation at selected locations throughout the plant wil1 be'ade prior to fuel loading 'ubsequent to fuel. loading, during'eactor heatup and.'t power levels of'pproximately'5$ -60$ and 'f00%: of rated power gamma,ra@.ahtion: level. measurements: and where fast neutron. mea:suzements- wi11 be. made at selected.
appropriate,'herma1'nd.
locations throughout the plant.:.
Acc~etance Criteria=- Level. 1 The radiation doses of plant.
origin and the occupancy times of personnel in radiation zones shall be controlled consistent with the guidelines of the standards for protection against radiation outlined in 10CFR20.
Level 2 The radiation doses of plant origin shall meet the followinq limits depending upon which Radiations Zone- the radiation base survey point is. located=
1'4 2-208'
SSES-FSAR Note=
IV
Radiation III III
-Zone--
15 1.00 L'imi:t mRem/hr...
2 5 mRem/hr mRem/hr mRem/hr.
All areas designated Radiation Zone V have potential radiation doses of 100 mRem/hr. Readings taken in Zone V during the Startup Test Program may be less than 100 mRem/hr; however, since Zone V is defined in terms of potential levels, there are no Acceptance Criteria for Zone V base survey points ST-'3 --- Fuel Eoadi~n t
gest~bective- The- objective of this test is to achieve the full and proper coreefficient complement of nuclear fuel assemblies fuel loadinq evolution.
through a safe and Pre~re uisites The required Preoperational Tests have been completed.. Xn addition,. prior to starting this test. procedure,
'th'e"foll'owinq; prerequisites wilL be met',
Fuel. and ControL'od:. inspections-
- viU'e complete b, Control,'.Rods",; wxlL'e- instaZXeL, and: tested
Reactor vessels. water level vill, be established. and minimum leveLI prescribed'..
The standby liquid control system will be operable and in readiness.,
e Fuel handling equipment vill have been checked and dry runs completed.
f The status of protection systems, interlocks,. mode switches, alarms, and radiation. protection equipment will. be prescribed and'erified.
":.q,.-. Rater'u'ality:: must meet. required:. specifi*cations I
The: follovinq 'prereguisites vill be; met. prior to commencing.
.fi, ~
actual. fueZ loadinq, ta assure that'his operation is performed in a safe manner; The status of all systems required for fuel loading will be specified and will be in the status required.
b.. AK least two movable. neutron detectors will be calibrated and; operable.. At least two neutron detectors. will be-connected. to the- hi'gh: flux scram, trips., They vilL be I
2='20 9. r I'4
~ r r Ir I.
I
SSE S-PSAR located so as to provide acceptable signals during fuel loadinq c Source range monitoring Nuclear instruments will be checked with a neutron source prior to fuel loading or resumption of fuel loading if sufficient delays are incurred The status. of secondary containment will be specified and.
established e., Reactor vessel status w'ill be specified rel'ative to internal component. placement and this placement establ'ished. to make the vessel ready to receive fuel f: The high. flux trip'oints vill be set. for a relatively lov.
power level q Neutron sources vill, be installed near the center of the core and at other specified locations'~est-ethod- Before the first feel assembly is taken. from the-
'anceinsertecL" j;nto"'thh-'dactodore'.. components'. fuel "."
-. fue3" po'olcastings, blad'e guidescontrol rod drives, etc.
{
will be
)
verified'This
'upport installed: tested. and/oz.. procedure begins. with:
the. steps": required'o; load.: neutron. sources includes: the actxvi'.ties- necessary'o'~ monitor. neutron.'opulation: using.
specially constructed'. fuel. loadi;"ng chambers', {PLCs)': and culminates: wi'.th the inseztion.. of: fuel, assemblies 'into the reactor core Puel loading continues. until the. core: is- fully loaded verified and ready. to perform subsequent Startup Tests.
Control rod functional tests, subcriticality checks, and a shutdown margin demonstration vill be performed during the loading..
Acceptance Crit'a I,evel 1 -- The partially loaded core must subcritical by at least 0.38% delta k/k with the analytically be determined, highest vorth rod fully- withdrawn.
gegt-.Gbgective--. The purpose. of this. test's; to d'emonstrate that.
the reactor,'will .be. subcriticaZ throughout the f'irst f'uel c'ycle with any single--control rod'ully'withdrawn.,
Prereguisites- The following prerequisites will be complete prior to performing the full core shutdown margin test:
a) The predicted critical rod position is available b) The Standby. L'iquid;. Control System is available
SSES-PSAR')
Nuclear instrumentation is available with neutron count rate of at least three counts per second and'ignal. to noise ratio greater than tvo to one d) Hiqh-flux scram trips are set conservatively low e) Instrumentation has been checked or calibrated. as appropriate Test Method This test will be per formed in the fully loaded core in the xenon-free condition. The shutdown margin test will be performed by withdrawing the control rods from the all-rods-in configuration unti1 criticality is reached Zf the highest worth rod. vill,,not be withdrawn. in. sequence . other rods ma:y be withdravn. providing. that the reactivity worth is equivalent The difference between the measured Keff and the calculated Keff for the in-sequence critical will be applied to the calculated value to obtain the true shutdown marqin.
Acceptance criteria Lev~e. 0 The shutfosn margin of the fully loafef coif (68oZ) xenon-,freel core occuring; at the 'most m ~ : ireactiveiti'med'using. the." cycle: 'mu'st "be-" at lea&') ".38% 'de3.ta'/k-vith the analytically strongest rod (or its reactivity equivalent)'ithdravn , If. the- shutdovn: margin is measured. at some txme durinq. the cycle: other than. the most reactive txme compliance with the above criterion. is shown by- demonstrating; that the shutdown. margin is. 0 38% delta. k/k plus. an exposure dependent correction factor which. corrects the .shutdovn, margin at that time to the minimum shutdown margin..
I.evel 2 Criticality should occur within +1.0'A delta k/k of the predicted critical.,
/ST~5. Con~tel non'fiv~aS stem Te t 0+bectiye The objectives of the Control Rod Drive System test are; a) to demonstrate that the Control Rod Drive (CRD)
S ys t em o perates properly over the full range of primary coolant temperatures and pressures from ambient to operating, a.n )
determine. the- initial. operating characteristics of the. entire CRD.
System.,
~Pe eguisites. The reguiref preoperational tests have been completed ~ k Test Nethod The CRD tests performed during the staztup test program are designed as an extension of the tests performe durinq the preoperational CRD system tests. Thus, after it is verified that all, control rod drives operate properly vhen installed., they are tested'eriodically during heatup to assure that there is. no significant binding caused. by. thermal expa'nsion
SSES-ZS AR' of'll v
ofy the core components . Al list control. rod dr'i;ve tests" to be performed'uring startup testi;ng is. given in Table,14.2-,5 Acc~etance Criteria Level 1 Each CRD must have a normal withdraw: time greater.. than or equal to 40'econds The mean scram time of: all operable CRDs must not'xceed the values specified in the plant technical specifications (Scram time is measured from the time the pilot scram valve solenoids are deenergized.)
The mean scram time of the three fastest CRDs in a two by'wo array. must not. exceed. the values specified'n the plant. technical.
specifications. (Scram time is. measured: from the. time the. pilot scram. sol.'enoid:s,. are= deenerg ized)
Level 2 Each. CRD must have anormal insert speed of 3 0 a 0 6 inches per second,, indicated by a- full 12-foot stroke in 40 to 6Q seconds., With respect to the control rod. drive friction tests if the differential pressure variationexceeds. 15 psid for a continuous- drive in a settling test, mush be performed, in. which case'he: d'irfferential,'seedling, pressurte-':should'ot,~'be: less'. than '-.--"
30 psid, stroke=
nor should it vary by more- than. 10 ps'.d over a. full i
ii
/sr~6:--SRR'erformance snd'-Control',Rod-Se uence.
The testing'reviously contained', in thi.'s; test. has. been merged'nto ST'-10
/ST-~7 Reactor Water Cleanup System Tes~t. Ob'ectives The objective of this test is to demonstrate specific aspects of the- mechanical operability of the Reactor Water Cleanup System. (This test, performed at rated reactor pressure and temperature,, is 'actually the completion of the preoperational testing that could not be done without nuclear.
heating) .
'r~requi;sitys'; The-"-required preoperati'onal tests. have= been, compJeted" .. Instrumentation. has been checked or,-calibrated. as appropriate,'-" " -.
Test Rethod Wfth the reactor. at rated temperature. and'ressure process variables vil1 be recorded: during: steady state operation in three modes as defined by the System Process Diagram:
Blowdown, Hot Standby, and Normal. Additional system configurations will also be aligned to verify proper performance of the bottom head flow and. temperature indicators.,
Acc~etance Criteria. Level O' Not applicable 1,4',2-212; ' a e,
i v
a t I
SSES-P SAR Tevel-2. The temperature at the tube side outlet of the non-zegenerative heat exchangers {HRHX) shall not. exceed 130 P in the blowdown. mode and 120~F in the normaL mode The pump available NPSH will be- 13 feet or greater during the hot standby mode defined in the process diagrams The cooling water flow to the NRHX's. shall be limited to 6g above the flow. corresponding to the heat exchanger capacity, {as determined from the process diagram) and the existing temperature differential across the heat exchangers The cooling pater outlet temperature shall not exceed 180~P'uring two pump operations. at rated. core flow, the bottom head temperature as measured by the bottom drain line thermocouple should be within 30<P of the recirculation loop temperatures.,
Bottom head flow indicator PI-28610 shall. indicate within 25 gpm of RQCU flow indicator PZ-2R609 when totaL system flow is thru the bottom head, drain.,
~ ~ : 1::::a-M "" .'em n
ectrees The: objectives of this test are to demonstrate
"'est~ob
~ ~
the, ability. of the Residual. Heat Remova1 {RHR) System to= 1) remove heat from'he reactor pressure: vessel, and. the suppression.
poo3'nd'); operate in the suppression. pooI, cooling mode steam-condensing. mode and'hutdown. cooling mode Prere~uisites- The reguired preoperational tests have been completed. Instrumentation has been checked or calibrated as appropriate..
Test-Method The suppression pool co'oling mode and steam condensing mode will be used to measure the RHR heat exchanger cap'acity Data will be obtained to determine the heat tzansfez rate with rated flow on 'both sides of the heat exchanger For the suppression pool cooling mode test attempts will be made to establish a large. temperature d,ifferential between the service and- suppression; pool water by'xtended. RCTC or. relief vaLve operations - Heat exchanger capacity in. the steam condensing mode-wiII be. measured: with the- reactor in.'ower., operation, supplying a steam source to the RHR heat exchangers Due to the: insufficient decay heat load". during the sta'rtup test period, full heat exchanger heat capacity in the shutdown cooling mode cannot be measured without the risk of exceeding the 100~P/hr cooldown rate limit of the reactor pressure vessel. scheduled Shutdown cooling mode trips and operability will be demonstrated after cooldowns duzing the Startup Test Program 14 Z-'2.1'3,
SSES-PSA'R St~am. condensing mode control system stability will be demonstrated with the reactor in power operationsupplying a steam source to the RHR heat exchanqers Acceptance-Criteria Level 1 The transient response of any system-related variable to any test input must not diverge Level 2 The RHR system shall be capable of operating in the steam condensing,. suppression pool cooling and shutdown cooling modes at the heat exchanger capacities indicated on the process diagrams Both simultaneous operation of RHR loops and single loop operation shall be tested in the steam condensing and shutdown cooling modes Each BHR- loop shall be tested independently in. the suppression poo1 cooling mode System-related variables. may contain oscillatory. modes of response In these cases the decay ratio for each controlled mode of response must he less than or equal to 0 25
~ST-9 water ~rev 1 aeasuremeut Tegt~~bectives- The: object'ive of this, test, is to determine
.-actual..referen'ce I'eg..-temperature and-. recalxbdate'.i'nAruments:
.if'ecessary eagreuisites The required preoperationeE tests have been:
~p completed calibrated AXE. system instrumentation ir instalI:ed and.
Test Method' At rated temperature and pressure under steady state conditions, the reference leq temperature will be measured and compared to the value assumed during initial calibration the difference of the two temperatures exceed the Acceptance If Criteria, then the instruments will be recalibrated. using the measured value Acceptance Criteria Level 1 Not applicable Reve/ 2 The difference between the actuaI. reference leg temperature (s) and. the value {s) assumed during calibration shall be less than that amount.- which;,wi;.XI'esult f.n. a scale end. paint error, of N, of the-.. instrument, spa+ f'r each, range
/ST-10)-- SRM- and'- IRM= Performance" and Control Rod Seguence-Test Objectives The ob jectives of this test are: (a) to demonstrate that the operational sources, SRM and IRM instrumentation and rod withdrawal sequences provide adequate information to achieve criticality and increase power in a safe and efficient manner for each of the. specified rod: withdrawal sequences and (h) to adjust the Intermediate Range Monitor System as necessary to obtain the desired overlap with the SRM'nd APRM
SSES-CESAR systemsa. (Note that this. test nov includes testing previously contained- in. ST-6)
Prereguisites The reguired preoperational. tests have been completed.
T~es method Source range monitor count-rate data will be taken and compared with stated criteria A withdrawal sequence,has been calculated vhich completely specifies control rod vithdravals from the all-rods-in condition to the rated power configuration Each sequence vill be used to attain. cold. criticality Movement of rods in a prescribed sequence is monitored by the Rod.
Worth Minimizer and rod sequence control system, vhich will prevent out of sequence withdraval Initially the IRH system is set during the is Preoperational Test verified the first time overlap Program SRM-XRM sufficient. neutron and ZRH-APRM flux conditions arise 'Ber. the APRM;
'a'Xibrati.'o'n -.the:.'IRH'ains 'wi1I.: be. adjusted'"as" neaessary-to optimize'he IRM overlap, with the SRHs and. APRMs-Acc ance-Criteria-- TeveIY,.There must: be a: neutron signal.
count.-to-noise count ratxo of't. least: 2'o'I on the required operable SRMs There must be a. minimum count .rate, of 3 counts/second'n, the required. operable SRMs.,
Each IRH.channel must be adjusted so that overlap vith the SRMs and APRMs is assured.
The IBMs must be on scale before the SRMs exceed the rod block setpoint
)ST-11) LPRM Calibration T~est Ob 'ectives- The objective of this test is to calibrate the Local Power Range Monitoring System h
L-completed w Instrumentation for calibration has been.
channels will be calibrated'o make the checked'ush-method--
The LPRR LRRR readings proportional to the neutron flux in the water gap at the chamber elevation. Calibration factors vill be obtained through the use of either an off-line or a process computer calculation that relates the T.PRM reading to average fuel assembly power at the chamber height Ancetance-Criteria Level 1 Not applicable.
2-'2'15
I.evel 2. Each= L'PBH'ill, be vithin 10% of its calculated value
/ST-1~2- -- A'PR N= Calibration Test Objective- The objective of this test is to calibrate the Average Pover Range monitoring (APRH) system..
P~e~re uisites The required. preoperational tests have .been completed Instrumentation for calibration has been checked..
Test. Method- A heat balance vill be made after initially achieving- power. level associated with each- test plateau Each APHIDS'hannel. reading will..be. acLjusted; to be consistent vith the core therma1 power as. determined. from the heat balance During heatup a: prelimi'nary: calibration will be mad'e by adjusting the.
APBH'mplifier gains so- that. the. APRH readings agree with the results of a, constant heatup rate heat balance., The APBHs should be- recalibrated in the power range by a heat balance as soon as ad'equate. feedvater, indication is available
" Acce t ce-.C teria Level
-- .1' . The APBK channels must be
'calibrated," to,:rea'dq'ua3-'" to. ov.*korea.'ter than the act.ual" co'r'e:
thermal power Eeve1. 2' -Not..
applicable...'est Ojb ective- The objecti;ve of this test xs to verify the NSSS performance of the process. computer under- plant operating conditions Prerequisites The regu'ired preoperational tests have been completed .
Test method The Dynamic System Test Case vill be run to verify that the results of NSSS performance calculations are correct..
Acceptance-Criteria'- Level, 1 Not applicable Level '2; .;--
(1) The tfCPR'alculated'y. an independ'ent method'nd.'he process computer either .
a Are in the same fuel assembly and do not differ in value by more than 2% or, b For the case in which the HCPR calculated by the process computer is in a different- assembly than that calculated. by the independent method for ~
both. assemblies, the NCPR and CPR calculated. by 14~2=21 6
SS ES-P SAR the two methods shall. agree within 2% For the same assembly (2) The maximum LHGR.calculated by the independent method and the process computer either=
a, Are in the same fuel assembly- and do not differ in value by more than 2% or b Por the case in which the maximum LHGR calculated by the process computer is in a different assembly than that calculated by the independent method,.
for both assemblies,. the maximum LHGR and LHGR calculated'y the two., methods. shall, agree within 2% for the same. assembly The MAPLHGR calculated'y the independent method and the process computer either=
a0 Are in the same fueI assembly and do not differ in
'aIue by more" than 2% or.
P L
b Por the case: in which the. MAPLHGR calculated by the process computer is; in: a different assembly than.. that: calculated." ,by, the independ'ent method for.
bath'assemblies;. the MAPEHGR ant APTHGR the two. methods shakl. agree. within: 2% for the caIculated'y same assembIy (4) The LPRM calibration factors- calculated, by the independent method and the process computer agree to within 2%.
Test Objective The objectives of this test are to verify the proper operation: of the Reactor Core Isolation Cooling (RCIC) system at the minimum and rated operating pressures and flow ranges, and to demonstrate reliability in- automatic mode starting from. cold standby."..when. the reactor.. is at, power cond'itions..
~Pe~guisrtes The 'regula red. preoperational tests. have been completed- Initial turbine- operation {uncoupled) must have been performed to verify satisfactory. operation. and. over-speed .trip Instrumentation has. been installed an>Cold>>'s- defined as a minimum three days without- any kind of RCIC operation..),
"After'he m'anual.- st.'azt" portion of cez'tain: oK the above. tests-.is':,
I and while the system is. still operating small step disturbances in: speed a.nd'. flow. command: are input; (in. manual and*
'ompleted
~
automatic;, mod'e-'respectively)';n:- ozd'er to- demonstrate satisfactory stabr1i'ty This. ir, to- be.. done- at.. both-. Tow-': (above minimum. turbine speed)'nd; near; rated.'low initfa1. conditions to- span the BCIC operating range During.",testing: at 150 psig,, this. is done: only.
near rated: flow initial. conditions , '
demonstration of extended operation of up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (or until pump and turbine oil temperature is stabilized) cf continuous runninq at rated flow conditions is to be scheduled at a convenient time during'he: Startup. Test Proqram Acceptance. Criteria Level must be equal to or greater l The average pump discharge flow than the 1005 rated value after 30 seconds. have elapsed from automatic initiation at any reactor pressure between-150 psig (+15.-0) (10 5 kq/cm~) and rated..
k The'.BCIC,'urbine'shel'1"; not'.tzzp""or; isolate.: during auto or manual start tests.,'=
Hote=
be If any" TeveI'" criteria are not met, the reactor will only allowed to operate-. up to a restricted. power l'eve1 defined by Figure 14.-2-7'ntil the problem is resolved. Also consult the plant Technical Specifications for actions to be taken.
Level 2 In order to provide an overspeed and isolation trip avoidance margin., the transient start first. and subsequent: speed peaks'hall not exceed'% above the rated., RCXC turbine speed 1
SSE S-P SAR The speed and flaw control loops shall be- adjusted. so that the decay ratio of any RCIC system related variable is not greater than 0 25 The turbine gland seal condenser system shall be capable of preventinq steam leakage to the atmosphere The delta P switch for the RCIC'team supply line high flow isolation trip shall be calibrated to a differential pressure corresponding to less than or equal to 300% of the maximum required steady state flow, with the reactor assumed to be near.
the pressure for main relief valve actuation
/ST'-'l5~ HPCZ'SStem'est Objective- The. objective of this test is to verify the proper operation of the High Pressure Coolant Injection (HPCI) system at the minimum and rated operating pressures and flow.
ranges, and to demonstrate reliability in automatic mode starting from cold standby when the reactor is at rated pressure conditj;ons Pz~e uisites The required, preoperationa1 tests have been
~e completed Initial turbine. operation. (uncoupled) must: have been.
performed to'eri'atisfactory, operation and. aver-speed'rip Instrumentation has bee@,'installed; and.'al'ibrated Test Method' The HPCI'ystem: is, designed to be tested in., two.
ways= (1) by flow injection into a. test. line leading to the Condensate Storage Tank (CST) and (2) by flow injection directly int o the re acto r vessel.
The earlier set of CST injection tests consist of manual and automatic mode starts at approximately 150 psig and near rated reactor pressure conditions. The pump discharge pressure during these tests is throttled to be approximately 100 psi above the reactor pressure to simulate the largest expected pipeline pressure drop. This CST testing is done to demonstrate general system operability and stability Reactor vesse1.injection. tests are also done. which: consist. oE manuaI and'utomatic mode start. near rated.'eactor pressure to demonstrate operability and: stability After all final controller and system adjustments have been determined, a defined set of demonstration tests must be =
performed with that one set of adjustments. Two consecutive reactor vessel injections startinq from cold condidtions in the automatic mode must satisfactorily be performed to demonstrate system reliability. (" Cold" is defined to a minimum three days without any kind of HPCI operat'ion )
SSES-PSAR After the manual start portion of certain of the above. tests is completed and while the- system is disturbances in speed and flow still command.
operating, small step.
are input (in manual and automatic mode respectively) in ozder to demonstrate satisfactory stability This is to be done at both low (above minimum tuzbine speed) and near rated flow initial conditions to span the HPCZ operating range During testing at 150. psig this is done only near rated flow initial conditions.
A continuous running test is to be scheduled at a convenient time during the Startup Test. Program.. This demonstration of extended operation shou1d be for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> oz until steady turbine and pump conditions aze reached. oz unti1 limits on plant operation are encountered.
Pump flow testing will also be: verified since auxiliary boiler supply is insufficient to fully test the sytem during the Preoperational Test Program.
Acceptance Cci~tia Level 1 The average pump lischamge flow must .be. egual to or greater than; the .1'00% rated" value after 25 se'c'one's'-.'ave eIapded'from,"automatic Snit'tionat 'any 're'actor pressure between 150 psig (+15 -0) (10 5 kg/cm<)- and rated The HPCT'urbine shalX, not trip or isolate d'ning auto or manual, start tests revel g= Zn order to pzovide an overspeed and, isolation trf.p avoidance margin the transient. start first peak shall not come closer than 15'X (of rated speed) to the overspeed trip, and subsequent speed peaks shall not be greater than 5% above rated turbine speed The speed and flow control loops shall be adjusted so that the decay ratio of any HPCI system related variable is not greater than 0.25.
The turbine gland seal condenser system shall be capable of preventing steam leakage to the atmosphere w
The. d'elta;P: swxtch. for the HPCT steam supply line high..flow isolation-trip shaIX be calibrated. to actuate at no'reaterreactor than 300% of the markmum requirecU steady. state flow with the assumed to be near the: pressure for main relief valve actuation.
/ST-16$ Selected Process Temperatures Test. Objectives The ob jectives of this procedure aze a) t o establish the propez setting of the low speed limiter for the recizcula.tion pumps to avoid coolant temperature stratification in the reactor pressure. vessel bottom head, region~ b) to identify any reactor operating modes that cause temperature 1'4 2-ZZO
, h SSES-CESAR stratification,, and c) to familiarize the plant personnel vith, the temperature differential, limitations of the reactor system P~egeguisites "- The- required'reoperational tests have been completed., System instrumentation has been calibrated;.
Test Nethod During initial heatup vhile at hot standby conditions,. the bottom drain. line temperature, recirculation loop suction temperature and applicable reactor parameters are monitored. as the recirculation flow is slowly lowered to minimum stable flov Utilizing this data coolant temperature stratif ication it can be determined whether, occurs when the recirculation pumps a.re on and prevent, it if so what mi;nimum recirculation flow wi11 Honitorinq the- preceeding information during planned pump trips vill determine if temperature stratification occurs in the idle recirculation loops or in the lover plenum* when one or..more loops are inactive Acceptance'-Crite~r a . Level. 1-- The- reactor.- recirculation,,pumps; sha13;..'not .be.:strait'ed'r noh" flov in'cceaose'd. Qnletss: the."caolant.
temperatures betveen. the 'steam dome and. bottom head drain are within 105>P t
- i. t The recirculati;"on pump xn. an. idle loop= must not be: started'nless:
the loop" suction temperature is. vithrn. 50OF'f, the 3
active loop The recirculation pump in. an idle loop must not be started unless the operatinq loop flow rate is- less than or egual to 50% of rated'oop flow When both loops have been idle, an idle recirculation loop shall not be started unless the temperature differential betveen the reactor coolant vithin the idle loop to be started up and the coolant within the reactor pressure vessel is less than or equal to 50~P, L'evel 2- Not Applicable v 'o QS'Z 17)-'. ~S. stem.-'Bzpanszoa-,
i Test'O~bectiwes- The purposes of this test aze to demonstrate tha4 reactor recirculation main steam inside containment,. and those piping systems- identified'n, Table 3 9-33 respond. to therma1 expansion consistent with stress analysis results. {Note that this test nov includes piping previously contained in ST-
- 38. )
Prer~euisities- Instrumentation. has been installed and calibrated 10 2-22;1'=,.
SS ES-PS AH'Te t-'~de hod. Hanger positions and locations of piping in the Nuclear Steam Supply System an'd piping systems identified in Table 3 9-33'nside and outside the reactor drywell are recorded, prior to initial heatup and. after a planned cold shutdown..
During initial heatup,. visual inspections are made at intermediate reactor water. temperatures and at rated temperature to assure components are free to move as designed Adjustments are made as necessary. Devices for measuring continuous pipe deflections are mounted on main steam, recirculation and other selected lines Motion during heatup is compared with calculated values.
Acceptance Criteria Level 1 There shall, be no obstructions which will interfere with the thermal. expansion of the main steam and. recirculation piping systems Piping systems identified. in Table 3 9-33 will not be restrained against thermal expansion except by design intent.
Hangers on piping systems identified in Table 3..9-33 shall not be bottomed out or have the spring fully stretched Snubbers on
.pipinqp systems. identified in Table 3 9-33 shall not become extended'r compressed:'to..the:"limit+ o&.their:total.;trav'eI;.-."* "..'he measured. displacements. at: the established. transducer locations on the. main. steam. and recirculation systems shall not exceed; the allowable values: calculated far, the specific points II Level 2- The measured displacements at: the established transducer locations on the- main steam and. recirculation systems shall not exceed the expected values calculated for the specific points. The measured displacements at the established transducer locations on the piping systems identified in Table 3.9-33 shall be within the acceptable range calculated for the specific points Hangers on piping systems identified in Table 3 9-33 shall be in their operatinq ranqe
/ST-18} - TIP Once t int I Te t Ob ectives-- The: objective of this uncertainty- of the, TZP system. readings test z.s to determine the-Prer~euisites-. System installation is completed'nd'equired preoperational tests. are completed and. verified. Instrumentation has- been calibrated and installed Test Method The TIP uncertain ty consists of a random noise component and a geometric component, the qeometric component being due to variation in, the water gap geometry and TIP tube orientation from TIP location; to location. Measurement of these components is obtained. by taking repetitive TIP readings at a 14 2-222
SSES-'CESAR single TIP location,, and by analyzing. pairs of TIP.'eadings taken at TIP locations which are symmetrical about the= core diagonal of fuel loadinq and- control zod symmetry .
The random noise. uncertainty is determined. from successive TIP runs made at the common location (32.-33). with each. of the TIP machines makinq six runs at index, position 10 'Xhe TIP data wilg..
be obtained by simultaneous operation of the Process computer OD-2 program which provides 24 nodal TIP values foz each TIR traverse The standard deviation of the random noise is derived by. takinq the squaze root of the average of the variances at nodal levels 5'hrough 22, where the nodal variance, is obtained from. the fractional deviations. of the successive TIR valuer- about their noda,l mea.n; value The total TIP uncertainty is: determined by perfozminq a complete set of TIP traverses as required by Process Computer program OD-1'. The total. TIP uncertainty is obtained by dividing the standard deviation of the. symmetric TIP pair nodal, ratios by the square root of 2 The, nodal TIP ratio is defined as, the nodal B'A'SE'alue of'he- TIR'n- th'.'ower right'alf of the core divid'ed
'by'- 9'M symmetric; coun'perp'br'.=*iud 'the::upper"left'-haXf:. """'...--
The, qeometric: component; of; TIP'.uncertainty" is; obtained:
rand'om noise. component from. the by'tatistically'ubtracting'he.
total TIP- uncertainty.,
The TIP data. will. be taken, with. the reactor operating with an.
octant symmetric: rod pattern and'; at steady state conditions One set of TIP data will be taken at approximately 50% power and at least one other set at 75% power or. above. The acceptance criteria for this subtest uses the >>average uncertainties" for all data sets.. Therefore additiona1 performance of the subtest may be scheduled. and the previous values of uncertainty will be used in the averaqinq to determine the acceptability of the results Ancetance criteria- Level g Not applicable.,
~E~v 1 2'he --t-'otey. TXP uncertainty'including rand'om noise- andi.
- geometricat uncertainties); obtained" by:". averaging, the uncertainties for aZI;data;'sets 'must be less'han:-"6.0%'OTE'.
A'inimum- of two and'p= to six: data sets. may, be used to meet the. above criteria- ..
JST-19} -Core Performance Test Objectives. The objectives of this test are a) to evaluate the coze thermal power and b) to. evaluate the following core performance parameters 1). maximum linear heat generation rate 1'4-223.
SSE S-TSAR (MLHGR): . 2) minimum critical pover ratio,(MCPR) and 3) maximum average planar linear heat generation, rate {MAPEHGR)
P~e eguisites The required, preoperational, tests have been completed Test"-Method The core performance evaluation is employed to determine the principal thermal and hydraulic parameters associated with core behavior These parameters are:
Core flov. rate Core thermal'. power= level.
MT.HGR'CPR MAPLHGR Prior,to- th-verification of the Process Computer. in ST-1'3 "an
'"'i'nappe'nd(erat. m'ethdd: vms X3:. he- used"to" caI'cuXIte these '.par'ameters After the successful, completion. of ST-13, the process computer vill'e- used tunes- Criteria- Level' The Maximum. Linear Heat I
A'c e Generation Rate. (HLHGRb of. any rod'uring; steady-state conditions shall not exceed. the limit specified; by..the. Plant Technical.
Specifications=
The steady-state Minimum Critical Power Ratio (MCPR) shall not exceed the limits specified by the Plant Technical Specifications.
The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) shall not exceed the limits specified. by the Plant Technical Specifications; (
Steady-state reactor power shall be: limited, to the rated MWT and valuer- os or below the- licensed analytically: determined power; fl(ov .1zne
( ( 1 m Level 2.' Hot. applicable
/ST-20}y' Steam- Production -V'erification (This test deleted from the PSAR for Unit 2) .
/ST 21} Core Power-Void Node Response Test O~b'motives: The objective of this'est is to; verif y the stability of the core power-void dynamic response.
'10'?=220
SSES-PS'AR P~e~eg~sites: The required. pzeoperationaI tests have been completed Instrumenta.tion has. been. calibrated Test Method- The core power void loop mode that results fzom a combination of the neutron kinetics and core thermal hydraulic dynamics is least stable near. the natural circulation end of the rated 100 percent power rod line A fast change in. the reactivity balance is obtained by moving- a very high worth rod only 1 or 2 notches, by simulating a failure of the pressure requlator and by performinq pressure regulator setpoint changes Acceptance- Criteria Level 1" The transient response of any system related variable to any test input must not diverge
~evel. 2-- Nab applicable (ST-22) - Pressure Reaulator Test Objectives- The objectives of this test are to demonstrate the takeover capability of the backup pressure regulator upon failure of the, cqntrollincp pressure regulator and. to demonstrate
'smooth .Qr'essur'd-. co'ntzol. traipse.Mon betwe'en'the" control valves". 'ind.
~
bypass. valves when reactor steam generation exceeds steam flow, used' y the turbine Prer~euisf tes The required.'=preoperationaT test's have been completed instrumentation: has been, checked. oz calibra.ted. as:
appropriate Test Method The pressure set point will be decreased rapidly and later increased rapidly by about 10 psi and the response of the system will be measured in each case. Xt is desirable to accomplish the set point change in less than 1 second., At specified test conditions the load limit setpoint will be set so that the transient is handled by control valves, bypass valves and both. The backup regulator wi.ll be tested by simulating a failure of the operating pressure regulator so that the backup requlator takes over control. The response of the system will be measured. and
,k evaluated.,
I Acce a ce=C iteria---Level-O'-- The- transient response of any.-
pressure control system. related. variable- to any. test input must not diverqe Level-2 a) Pressure control system relate'd variables may contain oscillatory modes of response. In these cases, the decay ratio must be less than or equal to 0.25 when operating above the lower limit of the master manual controller.
2-225 e
SSES-FSAR.
~
b) When in. the recirculation manua1 mode,. the pressure response time- from initiation of, pressure setpoint step change to the turbine: inlet pressure peak shall be <10 seconds..
c) Pressure control system deadband;. delay etc, shall be small enough that steady state limit cycles (if any) shall produce steam flov variations no larger than +0 5 percent of rated steam flov d) The normal difference between regulator set points must be small enough that the peak neutron flux aad/or peak vessel pressure remain belov the scram settings by 7 5 percent and. 10 psi, respectively for the, Regulator Failure Test performed. at Test Condition. 6
/ST-2~3- Feedwa tenor. S stem Test-Objectives The objectives of this test are a) to demonstrate acceptable response to the feedvater control system for reactor water leveX:: control b) -to'. demonstrate stable. reactor response,to subcoolin'q changes.:,'"'i e "-loss'f feedvaMr heat'ing",
c) to demonstrate the capability of the. automatic core flov.
runback featuze. to prevent low water= level, scram following the trip: of one. feedwater pump and. d)~ to, demonstrate. the maximum feedpump runout: capability is. compatable'ith licensing assumptions Prereceuisites The required preoperational tests have been completed Instrumentation has been checked or calibrated as appropriate.
gest Method At Test Condition {TC) 1 vith the water level being automatically controlled using the lov load valve and the recirculation system in Manual +5 inch step changes in the water level setpoint will be made to demonstrate proper response and operability of the feedwater system at low reactor power At Test Conditions 2 3 and' vith one feedwater pump in manual and. the. othezs ia auto,,-small."and 1'arge*flov changes ia the manually'ontrolled'. feed 'pump viH .be .made The response of the feedvater system. to these- steps: vill, be anaIyzed: aad. compared to the applicable acceptance criteria The-zecirculation system vill be in manual for these tests., A't Test Conditions 1 2 3, 0,, 5. 6 6 with the recirculation system in. manual +5 inch chanqes in the vater level setpoint will be made to demonstrate proper response and stability of the feedvater system.
At approximately 80%%u to 90% power vith core flow near 100% of rated,, a simulated failure of the extraction steam valves to one of the feedwater heater trains is accomplished by closing the heatez. train steam inlet isolation valves vhich will isolate 1 4,2-226.
SSES-PSAR extraction steam from the last three stages of that train Recordings of the transient will be analyzed and compared to the predicted response and acceptance criteria At Test Condtion 6, one feedwater pump will be tripped to demonstrate the capability to avoid a scram and prevent a low reactor water level. trip due to the loss of one feedwater pump.
A maximum feedwater zunout capability test will he done to demonstrate that the actual capability is compatible with licensing assumptions Acceptance=Criteria Revel 1 -- The transient response of any level control system-related variable to any test input must not diverge Por the feedwater heater. loss test, the maximum feedwater temperature decrease due to a single "failure case must be less than or equal to lOO~P , The resultant MCPR must be greater than the fuel thermal safety limit Pov='the feed:water,heater loss,'test"the inczea'se in heat flux' '.
~ ~
cannot exceed the predicted Level 2 value by more than 2% The pred'ictecL value will be, based: on the. actual. test values: of feecLwatez temperature- cha.nge ancL power level<<
The feedwater flow. zunaut capability musC not exceed the assumed value in the PSAR I,evel g I.evel control system-related variables may contain oscillatory modes of response., In these cases, the decay ratio for each controlled mode of response must be less than or egual to 0 25.
The open loop dynamic- flow response of each feedwater actuator (turbine or valve) to small (<10%) step disturbances shall be:
(1) Maximum time to 10$ of a step disturbance 1 1'ec
{2) Maximum time from 10%'o 90% of a step disturbance 1. 9 sec .
(3) 2eak;. overshoot. (%: oC step disturbance)' <1'5%.
The. avezage rate of'esponse of," the feed'water actuator to lazge
(>20%'f pump flow) --step disturbances shall be between 10 percent and 25"-percent rated feed'water flow/second This average zesponse. rate will he assessed by detezmining the time reguired to pass linearly- through the 10 percent and 90 percent response points.
For the feedwater heater loss test the increase in heat flux cannot exceed the predicted value referenced to the actual Peedwater temperature change and. the initial power level.
T4 ~,2-227
SS ES-PS AR A scram. must be. avoided from lov vater level with at'east a 3 inch mazgin following a. trip of- one of the operating feedwater pumps With extrapolated reactor pressure equal to 1060 psig, the sum of the calculated maximum reactor feed pump flovs must be greater than 15 4x10~ lbs/hr.
With extrapolated reactor pressure equal to 1010 psig, the sum of the tvo smallest maximum reactor feed pump flows as calculated
. must be gzeater than 9 1x10~ lbs/hr iST'-2ai Tu~bin~eVa ve Surveillance j'~es objectives The objective of this test is to demonstrate acceptable procedures and maximum povez levels. for periodic surveillance testing of the main turbine contzol,. stop, intercept and bypass. valves vithout producing a reactor scram
~pre eguisites The reuuired preoperational tests have been completed';. Instrumentation. has been checked. or calibrated. qs
~Test- ethod-- The test of the: control main. stop intermediate stop and: bypass; valves: are pezf'ozmed near. the: pzedicted highest:
power level to d'emonstrate.-that the Acceptance Criteria. are.
satisfied Rate of'alve stroking; and timing oZ the close-open sequence vill be. such that introduced. and that PCIOMR minimum practical disturbance is.
limits are not exceeded Acc~eta nce Crit eria Level 1 Not applicable.
Level 2 Peak neutron flux must remain at least 7.,5% below the Neutron flux scram trip value Peak vessel pressure must remain at least 10 psi below the high pressure scram setting. Peak steam flov in each line must remain at least 10'g belov the high flov isolation trip setting., Peak simulated heat flux must remain at least 5% belov its scram trip point..
/ST-25)i--Ãain--Steam-Isolation-Valves.
h Test-Objectives-- The objectives- of this test a.ze (a) to functionally check the main steam isolation valves (BSIVs) for proper operation. at selected power Revels, (b} to determine reactor transient behavior. during and following'imultaneous full closure of all HSIVs, (c) to determine isolation valve closure time and (d) to determine the maximum power at which a single valve closure can be made vithout a scram.
Prereguisites The required preoperational tests have been completed. Instrumentation has been checked or calibrated as appropriate 14 2.-228 v
SSES-,ES'A'R Test-Hethof The Hain Steam. Isolation Va1ves (HSZYs) are operateCZuring: this test to verifV their functional performance and to determine, closure times, While. functionally testing the operation" of the MSIVs,. the time necessary for closing each individual valve- vill be noted., The fastest MSIV vill. then be tested to determine what power level an. MSIV can experience fast closure vithout causing a scram A'll MSIVs vill later be used to demonstrate a full isolation subsequently leading to a scram.
(The Nuclear Steam Supply Shutoff System (NSSSS) logic will be used to initiate the full isolation).. The acceptability of the fast criteria (3 seconds) is determined by utilizing the full stroke time vithout delay extrapolated from measured stroke times betveen 10% closed and. 90%. closed The acceptability of the slov criteria; (5'econds); is;, d'etermined: by utilizing the full stroke, time: wi.'th dela:y: Mextra'polated. for the final. 10% of stroke, Acc~tgce ~Citeria-- Kevel 1 The positive change in vessel dome pressure occurring within 30 seconds after closure of all MSIVs must not exceed, predicted values .by more than. 25 psi h rl The positive change in heat, flux f'olloving closure of all MSIVs; sha-11 not exceed! pred'icted. values: by more;.than 2%'f rarted value EoXIoving: the" closure oZ aEX. MSXV~s'he-. reactor must. scram I
Closure time for.. any MSIV'nc].uding delay. sha3Z, not be greater than 5 5 seconds.
Closure time for any MSIV- shall not be less than 3.0 seconds nor greater than 5.0 seconds.
Zeedwater control settings must prevent: flooding the main steam lines during the full isolation t'est..
I;eys g 2. The positive change in vessel dome pressure occurring.
within the first- 30 seconds, after the closure of all MSIVs must not exceed the. predicted'values Predicted. values-. will. be ref erencecU. to"actua1 test.;conditions.,o C i:nitia3'.. po ver; le v'el scram- timi.:ng'nd" dome;. pressure- and.r vi'1L'se. beginning of. li fe nucEear data The positive change in. heat flur occurring "vihthin* the first 30 seconds after. the..closure of. all. MSIVs must not exceed the predicted values. Predicted values will be referenced to actual test conditions of initial pover level, and dome pressure and vill use beginning of life nuclear data.
Ef water, level., reaches Revel 2 setpoint during the. MSIV full closure test, RCIC shall automa.tically initiate and. reach rated flov 1'4-, 2-229 I
7 r
7 7
SSES-PSAR During the MSIV full closure test the relief valves must reclose properLy {without any detectable leakage) following the pressure transient Durinq. full closure of individual MSIVs, peak vessel dome pressure must remain at least 10 psi belov the scram setpoint During full closure of individual its MSIVs,. peak neutron setpoint..
flux must remain at least 7.5% below scram During full closureat ofleast individual MSIVs, steam flow in individual isolation trip lines must remain 10$ below the high flow setpoint During. full closure of individual, MSIVs the simulated heat flux must remain at. least 5% less than its flow biased scram setpoint
~ST 2~6 pelief Valves.
Test o~b. ectives The objectives of this test are to verify that the 'relief valves function properly reseat pr'operly after
~
oped'ation. and: contain;:no ma jar-'lockages in 'the'Ire'li:ef valv'e' piping 'ischarge
~P:er~eu~isf. e The regutred preoperationaI.'ests. have been completed I'nstrumentation has- been checked. or calibrated as appropriate Pactory- test. results. on SRV'low and operating times have been re vie wed'est Snthod Testing done at low reactor pressure, in conjunction with plant surveillance testing, consists of cycling each relief valve to verify proper operation. The transient monitorinq system will be used to record the results of this test The data collected will compare the operation relief valves aqainst the operation of all relief of'ndividual valves. During relief valve operation, core power and therefore steam generation rate is maintained constant. The pressure control system will close the bypass valves an amount proportional to the relief valve steam flow to maintain constant reactor pressure.-.= This: bypass. valve motion. will. be monitored:, and a comparison .oZ- the-response-for each rel.ief valve operation will be male .Tf dif'fezences. exist, =it could suggest a partial obstruction of the" relief valve or its- tai.'lpipe Tailpipe temperature wilI be recorded to verify'he relief valve has.
properly reseated Reactor. variables will also he recorded to verify system stability luring opening and closing each relief val veee Testing done at rated reactor pressure consists of manually operatinq each'elief valve at rated, reactor pressure. The decrease- in Main Generator output will be monitored during, the operation of each relief valve to provide= an indication. of relief 14 2-230
SSES-FSAR va3ve f3,ow.'y comparison of the generator output response for each relief valve operation, any flow obstruction in the valve or its tailpipe can be identified. Each valve will be opened.for approximately 10 seconds to allow for variables to stabilize.
Reactor variables will also be recorded to verify system stabil'ity during opening and closinq each zelief valve.
Acceptance-Criteria Level 1 There should be a positive indication of steam discharge during the manual actuation of each valve.
Level-2 Pressure control system-related variables may contain oscillatory modes of response. In these cases, the decay ratio for each controlled mode of response must be less than or equal to 0 25.,
The temperature measured by thermocouples on the dischazge side of the valves shall return -to within 10~F of the temperature recorded before the valve was opened.
Durinq the low pressure functional tests, the change in bypass valve position for each SRV opening shall be greater than or equa1 to a. value corresponding to the average change minus 10%%u of one bypass valve., t During the rated pressure tests, the change in. HRe for each SRV opening shall be greater than or equal to a value corresponding to the average change minus 0.5% of rated Hle..
/ST-27) Turbine Trip and Generator Load Rejection Test Objectives - The objective of this test is to demonstrate the response of the reactor and its control systems to protective trips in the turbine and generator.
~eg~eu jsi tes The requ ire d pr cop erat iona 1 tests have been completed.. All instrumentation has been calibrated.
Test-Method-- At, Test Condition 3,. a turbine trip vill be manually initiated, by depressing. the Turbine Trip pushbutton in.
the main control zoom. At Test 'ondition 6 a generator load
~
rejection will be manually initiated by remotely opening the generator synchronizing breaker from the control room. During both transients, reactor water level,, pressure, neutron flux and simulated heat flux will be recorded and compared to predicted results and acceptance criteria.
At approximately 24%%u power, a generator load rejection within bypass capacity will be manually initiated as described above.
This will, demonstrate the ability to ride throuqh a load rejection within bypass capacity without a scram.
14.,2-231
SSZ S-'PSAR During; qll. 3'= transients,. main tuzbine stop, control and. bypass.
valve. positions 'and reactor. water: level will be recorded and compared to the acceptance criteria Ance tance Crite~ri T.e vel 1'oz Turbine and, Generator trips there should be. a. delay of no more than 0 .1 seconds following the beginning of control or stop valve closure before the beginning of bypass valve opening The bypass valves should be opened to a. point corresponding to greater than or equal, to 80" percent of fuLL. open within 0.3 seconds from. the beginning, of'ontrol or stop valve closure motron b)'- Peed'water system settings must prevent flooding. of the steam- line following these transients c) The positive change in vessel. dome pressure occurring within 30'econds after either generator. or turbine trip. must,.not.exceed. the, I;evel. 2. czitezia by" more than
'.5 The positive'hange psi.').
i;n; simu3.ated', heat flux; shall, not:
exceed! the: EeveI 2. criteria. by. more than; 2%'f zatecL va<Iue-The: two'ump =drive flow coastdown; transient during the.
first three seconds of an RPT'rz.p, must fall. within the specified Limits.
Level 2 a) There shall be no. NSIV closure in the first 3 minutes of the transient and operator action shall not be required in that period to avoid the HSIV trip.
The positive change in vessel dome pressure and in simulated: heat flux which occur. within the first 30
- seconds after;'.the. iaitiation, of either generator or:
'=.:.tuzbine..trip, mast: not. exceed.",the pred'xcted values.
I (PzedictehcL'values will be refezenced to actual, test conditions of inztial'ower; Level dome* pressure, scram.
timing and the time from the stazt of stop/control valve motion to start of. control rod motion, and will use beginning of life nuclear data.)
c) Por the Generator trip within the bypass valves capacity (initial thermal power values-'less than. or.
equal. to 25 percent of rated) the reactor shall. not scram.
1'4 2-232:
I' I
SSES'-PS AR d), The Total Delay from. the initiation of a. Turbine Stop Valve Closure or Turbine Control Va1ve- Past Closure 'to complete suppression of the Electric. Arc hetveen the fully open; contacts of the Recirculation Pump Tzip
{RPT) Breaker shall be less than 175 milliseconds.
e) Peedwater level control shall avoid the loss of feedwater flow due to a high, level {LS) trip.
f) Peed.vater level control shall maintain water level above the L2 leve1 trip, setpoint for HPCI, RCXC and ATWS RPT ST- 8 .-Shutdovn.-from utsic}e the ~M n Control Room Test Objective" The objective of this test is to demonstrate
~
that the reactor can be shutdovn maintained in a hot shutdovn condition ant cooled down from outside the main control room Also the adequacy of the Emergency Operating Procedures will be verified..
prese uis'ites The required preopezational tests have been:
completed Instrumentation. has been, checked,.or calibrated as appropriate Test Method'-- RhQ.'e. operati:ng at approximately 20%'ower synchronized.'o the. grid vitk normal electrxcaZ. system alignment the. reactor vill he scrammed. anc} the HSIV's vill be closed from inside the main control room The control room will then be-evacuated,. and. reactor level and pressure vill he controlled from outside the main control room. The Shutdovn Cooling mode of RHR will be placed into service with cooling water suppliedsupervisory from the ultimate heat sink. During this demonstration some and operatinq. personne1 vil1 remain in the control. room to protect non-safety-related equipment from unnecessary damage if if conditions conditions arise anc} to assume control of the plant varrant A, test vill he run. to demonstrate that control the react'oz can be scrammed and isolated from outside the room
~acc tadce-Criteria -- Leve1'::I'--,'ot applicable II Tevel'-g--. Duzing, a simulated'ontrol room; evacuation the reactor.
must be brought to the point where cooldown, is" initiated and.
under. control ancL the reactor vessel pressuze and water level are controllecL usinq eguipment anc} controls outside the control room. The test is deemec} successfu1 when reactor pressure is less than the permissive setpoiat and the RHR shutdown cooling mode has been put in operation.
The reactor must be capable of being scrammed and. isolated- from outside the control room..
14 2-"233
SSZS-PS'AR:
/ST'-29L- - Recirculati.on- Plow-Control S stem.
The objectives of: this test are.-.
a) To demonstrate the flow control capability of the plant over the entire pump speed range,. includ:ing individual local manual and combined Masker Manual Operation b) To determine that. all electrical compensators and controllers are set for. desired system performance and stability P~e~euisites . The requi'red; preoperational, tests; have been comp1eted A'lL instrumentation, has. been. calibrated Test-Method- At Test Conditions 2', 3' and 6 the stability- of the recirculation flow control system is; demonstrated by performing step changes in recirculation, pump speed, This testing is done in individual local manual ak Test Conditions 2 and 5 and'n combined" Halter Manual operation at: Test. Conditions
- 3. and. 6 to demonstrate operability and stability Testing.'ilL'lso: be:- performed,',to~ verify. that the-: Recizc." K-G'. set hi*gh speed'. mechanic'ca1. stops." are~, propert'.,set'I r
4 system-related. variable" to: any'est input'ust not diverge Level 2-- A scram shalL not occur due to recirculation flow control maneuvers.
The APRN'eutron, flux" tri.'p avoidance margin shall. be greater than or egual to T.5% and the simulated heat flux trip avoidance margin shall be greater than or equal to 5% when the power maneuver effects are extrapolated to those that would: occur along the 100%'ated rod line.
,1he decay.'ra4io: of-:, any..".osciI'amatory,;controIled,variable;.must be-.
- -less.-than or-=equai;.to-",0'5'- t' I
Steady stak~ .Limit -cycles"..{if a.ny)>'shal'I: not produce turbine:
steam: flow'ariations greater than +..5%". oK rated;-steam flow fST-30)= .Recirculation Svstem Tegt Objectives- The objectives of this test are:.
a Verify that.. the feedwaker:.control system can.
satisfactorily control water level, without a resulting turbine trip and associated'cram.
1
SSE S-F SAR b Record and verify acceptable perfcrmance of'he recirculation two pump circuit trip system c Verify the adequacy. of the recirculation. runback to mitigate a scram.
d, Verify that no recirculation system cavitation vill occur in the operable zegion of the power-flov map P~egeguisites The reguized preoperational tests have been completed Instzumentation has been checked or calibrated as appropriate
~est+ethod= Single recirculation pump trips vil'1. he made at Test Condition (TC) 3 and'. TC-' These trips will he initiated by trippinq the M-' Set Drive Motor Breaker from the control room Reactor parameters will be recozded during the transient and analyzed to verify non-diver'gence of oscillatory responses, adequate margins to RPS scram set points and capability o'f the feedwater system to prevent a. high level trip The capability to restazt the zecirc pump at a hiqh pover level vill also be
'emonstrated.. At.TC-3,. bath- recirculation. pumps. RPT breakers will be simultaneously. tripped using a. temporarily installed test switch 'Zhe datagathereL wi11 he used.':c demonstrate acceptable pump coastdown performance." prior to high, power.'urbine trips and generator" load rejects,,
cond'itions. vill. be simulated, at TC-3 to demonstrate
'ppropriate-the'roper operation of the. recircuIation: pump runback. cizcuits This is done prior to an actual planned feed pump trip at rated power..
Both the jet pumps and the recirculation pumps vill cavitate at conditions of hiqh flow and lov pover where NPSH demands are high and little feedwater subcooling occurs., However, the recirculation flow will automatically runback upon sensing a decrease in feedwater flow.. The maximum recirculation flov is limited'y appropriate stops which will zun back the recirculation flow fzom the possible cavitation. region. At TC-3,,
it- vill- be. verified. that, these-.limits. are suffer.'cient to. prevent, operation-where recizculation'pump or- get. pump cavitation. occurs.
~ \ ~ '
Acceptance-Criteria --Kevel-I:-- The~ response- of any level during a singlepump trip must not- diverge. related'ariables The two pump drive flov coastdovn transient, during the first 3 seconds of an RFT trip, must fall within the specified limits.
Level 2 The reactor shall not scram during the one pump trip.,
The APRM margin to avoid a. scram shall be at least 7'..5% during the one pump'zip recovery.
14 .2-235
SSES'-ESAR The reactor water. level..margin. to avoid a high: level. trip. shall be at least- 3 0 inches during the. one pump trip Peak simulated heat flux must remain at least 5% below its flow biased scram setpoint.
Runback logic shall have settings adequate to prevent.
recirculation pump operation in areas of potential cavitation The recirculation pumps shall runback upon a trip of the runback circuit
)ST-31~Loss of Turbine-Generator and Of fsite Power
~Ts t objectives:-- The objectives: of. this test are to. demonstrate that the required'afety"systems will 'initiate and function properly without manual assistance, the electrical distribution and diesel. generator systems will function properly,, and the HPCZ and/or RCIC systems- vill maintain. water level of Unit if necessary during turbine-generator a coincidental loss the 2 main and offsite power to Unit 2.
pr~ere nisites The regnired. preoperationa1 tests have been completed Instrumentation has-. been. checked, or calibrated as appropriate n 4
Test-Method' Kith; the unrt synchronized to the grid. at approximately" 30$ power, the mailer turbine-generator will be manually tripped coincident vith: a. manual trip of the unit's offsite pover souzce breaker both trips initiated fzom the.
control room During Unit 2 testing,, to ensure a full simulation of the loss of all offsite pover to Unit 2 while minimizing the impact on Unit 1 operations, all Unit 2 loads vill be transferred to Unit 2 Auxiliary and. Startup busses, all Unit 1 and common loads vill be transferred to Unit Auxiliary and Startup Busses, 1
and appropriate breakers will be racked out to prevent automatic transfer of Unit 2 loads to Unit 1 sources Reactor water level and. the operation of safety systems vill be monitored to~ verify~.that the acceptance'riteria are satisfied'.
The, proper response of'he* eI'.ectrical distzibution- system will he.
checkecP.
The loss of offsite power cond'ition vill he maintained for at least 30 mi.'nutes. to demonstrate that. necessary equipment controls, and indication are available to remove decay heat from the core using only emergency power supplies and distribution system Acceptance- Criteria Level 1 All safety systems, such as the Reactor Protection System, the diesel-generators RCIC and. HPCI must function properly wit'hout manual assistance, and HPCZ and/or 1Q- 2 236
SSE S-P S'AR RCIC system action if'ecessary shall keep the- reactor water-
!level a'hove the. initiation leveL of Core- Spray LPCI and ADS Level 2.- The temperature measured.'y the thermocouples on the discharge side of any SRV'hat actuated shall return. to within 10oP of the .temperature recorded before the valve opened Permanent instrumentation for reactor power, reactor pressure, water level,. control rod position,, suppression pool temperature, high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC3 shall, be demonstrated. operable following re-enerqization of the OkV busses by the diesel generators
/ST-32j -Containment A'tmosDhere
~ and Main- Steam Tnnnel Co~olin Test Objective The objective of this test is to verify the ability of the drywell coolers/recirculation fans and maintain the reactor building portion of the main steam tunnel coolers to design conditions in the drywell and reactor. building portion. of the mainsteam tunnel, respectively,, during operating- conditions and* post scram conditions., ~ ~ ~ ~ ~
~ >> ~
Pre~euisites The required preoperational, tests have been completed: ., Instrumentation, has been. checked. or calibra.ted
',, as'ppropriate
~Te t'Method: -'Dnrfng: heatnp>> at'est conditions'. 2; an!T 6'nd.
fo3Zoafn<r. a: planned'cram from100:.%'oserdata eilI be. taken to.
ascertain that. the contasnment atmospheric conditions are- within desiqn limits Acceptance Criteria L'evel 1 The area under the reactor vessel in the Control, Rod Drive are is maintained below 185oP.,
Level 2 The general drywell area is maintained at an average temperature less than or equal to 135~P,. with maximum local.
temperature not to exceed 150oP The. area. beneath..the reactor pressure: vessel is maintained at an average- temperature< less than o: r. equaL'. to~ 135.oZ . maximum. Local, temperatureot;-to.'xceed; '.167<F with.,minimum: local. temperature above. 100oP The area around: the; recirculation pump.. motors. is. maintained at "an=
average temperature less than. or eguaL to 128~P'with, maximum local temperature not to exceed 1350P; The inside base of the shield wall in the RPV skirt area is maintained at temperatures greater than 100~P.
The reactor building portion of the mainsteam pipeway is maintained at or below. 125oP.
14'.-', 23T
! h s>> ,I>> ~
"!, 1 p*. !
I!
SSES-'FSAR Thy area surrounding the drywell head shall have an average temperature- equal to or. greater than 135>7'ith maximum local temperature not to exceed 150~P..
The reactor pressure vessel support skirt flange shall be maintained. at or below 150oP The temperature of the concrete- surrounding the primary containment main steamline penetrations are maintained less than 200 oP
~S>-33] p~iincC Steady State Vibration Test. Objectives- The objectives- of this; test is to demonstrate.
that stead.y. state vibration: levels on reactor recircula.tion, main.
steam. inside containment and those piping systems identified in Table 3 9-33 are within acceptable limits (Note that this test now includes piping previously contained. in ST-00. Also note that dynamic transient vibration testing previously contained in this test have been merged into ST-39.)
PrOp~egjisites Instrumentation'as been 'installed and calibrated.
Test method', Devices, for measurinu continuous vibra.tion. a: re mounted. on: manu steam Tines recirculation" Zi;nes and lines of systems- identified in Table 3 9'-33. as applicable and vibration-Curing steady state operation xs, compared,- with calculated values Ance tan~ce C iteria revel i The measured amplitude (peak to peak) of each remotely monitored point on the main steam inside containment and reactor recirculation lines shall not exceed the allowable value for that point.
Level 2' The measured, amplitude (peak to peak) of each remotely monitored point on the main steam inside containment and reactor recirculation lines shall. not exceed the expected value for that point..
The vibratory .response of: non-, remotely. monitored..systems or portions. of'ystems. indentifieK; in: Table 3'-33 she?1. be QudgecL to be within acceptable: Hmits by'. guaHHed test. engineer 4
The maximum measured amplitude of the pi:ping response for each remotely monitored point. on systems identified. in Table 3.9-33 shall not exceed the acceptable value for that point QST~34 C~otr~o Rod Seceuence Rxcbange This test will not be performed. during the Unit 2 Startup Test Program 10 2-238
SSES-ZSAR Tegt-Method- The control rod. sequence exchange begins. on the design flov control line- vith core floe near minimum Control rods will be inserted as- necessary- to increase the margin to local core thermal limits Core. poser is maintained above the low power setpoint of the Rod Worth Minimizer and Rod Sequence Control System and belov the power vhich will..keep fuel assembly nodal power at the PCIOMR threshhold. The exchange is performed in accordance with the plant operating procedure RZ-TP-009. Data taken duzing the exchange will be revieved to verify that the Acceptance Criteria vere satisfied
~Acce ta~ce .Cri~te ia Level 1 Completion. of the exchange of one rod. pattern. for the complimentary pattern with continual.
satisfaction of all licensed. core 3.'imits- constitutes satisfaction of the requirements of this procedure Level 2 All. nodal powers sha11 remain belov their PCIOMR threshold limit during this test
/ST-35).= Recirculation System Plow Calibration Test"Ojb ectives'--'The'bjective'"of this test is to perform'a.
complete calibration of the installed recirculation system flow instrumentatioa frere*"sxtes The regained preoperationaI. tests have been.
completed Instrumentation. hms been checked; or calibratecL as appropriate
+~est ethod Daring the testing program at selected operating conditions which allow the recirculation system to be operated at speeds required for rated flov at rated power, the jet pump flov instrumentation vill be adjusted to provide correct flov indication based on the jet pump flow After the relationship between drive flov and core flow is established, the flow biased APRM/RBM system will be adjusted to match this relationship.
Acce t nc '- C ite ia:. .LeveI.-*1,-- Not applicable v~
~Le eI-2'=- Jet pnap floe instrnsentatiom shall be ad3nsted. sech that the jet pump total flov. record'er. vi11 provide a correct core flow indication at. rated"'cond'itions The APRM/RBM flov-bias instrumentation shall be ad justed to function properly at rated conditions.
/ST-36l -Cooling Sate~r S stems This test will'not be performed during the Unit 2 Startup Test Proqram.
14 2-.23.9
SSES-PS'AR
~ST'-37 - Gas~eous adwaste~Sstem Tegt Ojbectiyes The objective of this test is to demonstrate that the Gaseous Radwaste System operates within the Technical Specification durinq a full range of plant power operation and to demonstrate the proper operation of the offgas and, containment nitrogen inertingI systems during plant aperatian..
W Prereguisites The required preoperational tests have been completed.. Instrumentation has been checked or calibrated as appropriate.. In addition, the 100% power trip testing shall have been completed or 1'20 effective full power days shall not. have elapsed price to performing the nitzogen inerting test Test method The. test wilZ consist of collecting data and necfoxminq. quantitative analysis of the off gas system effluent to determine if the performance is acceptable per the Technical Specification.. For the nitrogen inezting system, the proper nitroqen concen tration will, be verified. b y the as installed plant oxygen detectors/instruments in the two ma]or volumes of the primary containment Proper operation of. the offgas system. will also be verified
~Acce tance~Cxite ia. Level 1t. The nlees a'eof radioactive gaseous- and;= partxculate effluentr, must not. exceed.; the limits specified'n. the si'te technical specifications TeveX. 2-The system flow. pressure temperature and. reIative humidity. shall comply with design, specifications The catalytic recombiner AESOP the hydrogen analyzer the activated carbon beds and the filters shall be performing their required function., There shall be no less than 8000. lb/hr. of dilution steam flow when the steam jet aiz ejectozs are pumping The containment nitrogen inerting system shall be capable of inerting the primary containment free volume within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> f rom the start of the test and, the resulting oxygen concentration shall be les's than or equal to 4Ã
/ST-38$ " fiEing. Sfstem ExEansion i w
/The.system: expansion. testing previously'ontained: in. this:: test has been merged'to ST. '1T ).,-
L-:n.-
Test-Q~b'ective The objective of this test is to demonstrate that vibration levels on main steam inside containment, reactor recirculation, and system piping identified in Table 3.9-33 meet acceptable limits during selected dynamic transients.
Prereguisites: Instrumentation. has been installed and calibration.
1'4 ? 240 w
SSES-ZS'AR
~Te t Set.had Devices Xor measuring continuous disn1acements 'ccelerations: and pressures are loads .
mounted on piping systems and responses during transients are compared with calculated values. Those'ortions of the systems which are non-safety related are visually inspected prior to, during and subsequent ta the transient loading condition.
Acceg~ta ce Criteria Level I The measured vibration amplitude (peak to peak) for each remotely monitored point of main steam inside drywell and/or reactor recirculation piping shall not exceed the allowable value for each specific point Level The maximum measured accelerations on those systems listed. in Table. 3 9'-33 shall not exceed. the design maximum expected values at each. specific point The vibratory response of non-remotely monitored systems identified in Table 3 9-33 shall be judged to be within acceptable limits by a qualified. test engineer Based on visual inspection during a post transient walLdown, there shall be no 'signs of excessive'piping. response (such as damaged insulation markings: oz piping,. structural oz hanger steel or walls. damaged'ipe- supports, etc ), on: systems listed.'n TabI:e 3 9'-33' j
The measured'ibration amplitude (peak to peak), for each, remotely monitored- point of mai'n steam inside. drywell, and/or reactor.
recirculation piping shaI1 not. exceed. the- expected. value for each specific point r ~
/ST-40) BQP Piping Steady State Vibration (The steady state vibration testing previously contained in this test has been merged into ST33.)
v 14'-241
SSES-CESAR TABLE T4 2
~ 3:
STARTUP'EST'ROCEDURES'est-Number- Test Definition ST-1 Chemical and Radiochemical ST-2 Radiation Measurements ST-3 Fuel'L'oading ST-O. Full Coze Shutdown Mazgin; ST-'5. Control. Rot Drive. System:
ST-'6= SRM Performance and: Control Rod Seguence (Unit 1'nly)
ST-7 Reactor Water. Cleanup System.
ST-8 Residual Heat Removal System ST-9 . Water; Level: Measuz'ement "..
ST-TQ ZRM'. Performance- (Unit:.
h and', XHM'erformance ancL'ontro1 Rod'equence 1)'RM>>
(Unit 2).
h ST h
e 1 1" ST-12.
EPRM-APRM.
Calf;brati'oIz h
Calibration
(
(
- h ST-1'3 VASSS Process Compu ter ST-14 RCXC System ST-15 HPCZ System ST-16 Selected Process Temperatures ST-17'. 'ystem Expansion h
ST=1'F;; - .'TXP'n(certai;nty =
ST.19-4 h
Core- Per ormance f ST-20 Steam.- Pzoducti;on Verification. (Unit 1'n1y)
ST-21 Core Power Void Node Response ST-22 Pressure Regulator ST-23 Peedwatez'ystem ST-2Q Turbine Valve Surveillance ST'-25 Main Steam Isolation. Valves l(
l(,
Q h
( >, i fh
SSES-PSAR I
14= 2-'3- -/con&~
PROCEDURES'ABLE STARTUP TEST:
ST-26 Relief Valves'T-27 Turbine Trip and Generator Load Rejection ST-28 Shutdown Prom Outside the Hain Control Room ST-'9 Recirculation FIow Control System ST-30'ecirculation: Syst'm ST-31 Loss of Turbine Generator. and Offsite Power ST-32 Containment Atmosphere and, Main Steam Tunnel C'ooling ST-33 . Piping Steadv State Vibration ST Control, Rod', Sequence Exchange (Unit.
ST.-35.'. System Zloty'alibration., 1'nly)ecircuXatzon ST '36, Coofi:ng,Rate+" Systems; (Uni;t".1." only),
ST '37 Rad'waste- System 'aseous ST-38 BOP Piping System Expansion ST-39 Piping Vibration During. Dynamic Transients-ST-'40 BOP Piping- Steady. State V'ibration 4 'h A
a 1
SSES-PSAR TABLE'74=2 M'A'JOR TEST PHA'SE'ND'EST PLATEAU'CHEDULE TEST'ONDITION SEQUENCE Test Test Phase Plateau Test Condition S'eguence Open Vessel. Test Condition XV Heatup. Test Condition.
Test Condition during approach. to Test Condition I'esting 2
Test Condition 2.
Testing during approach to Test Condition 3 Test Condition 3 << C Testing duri.ng approach to Test Condition 5 Test" Cond'iti.on 5.
T'esti'ng: dure,ng: approach to. Test Cond'ition;- 6
,"." ': Te<<st'Conation:
6'*
.-'.',, Test Conditi.on Because oF the transitory nature of testing performed'long, the 'f00% rod line during Test Phase V'est Plateau D, all testing assigned to Test Condition 6 may not be completed prior to entering Test Condition 4.,
h A<<<< ~
~ '
TA'BEE 't4..2-5 CONTROL ROD DRXVE SYSTEM'TA'RTOP TESTS Reactor Pressure With Core Loaded Accumulator
-Rated psig00 Action Pressure 800 Position Indicati'on all Normal Times all Insert/Withdraw Cou ply.ng all Friction all Scram. Ti'mes all 4~ all Normal'inimum Scram Times Scram Times. Zero Scram, TZme's: Norma'X'.-
rh ~
Refers. to 4- CRDs selected for continuous monitoring based on slow normal accumulator pressure scram times, or. unusual operatinq,rcharacteristics, at zero reactor pressure or rated reactor pressure..when,.this data is available., The 4 selected CRDs~-mustbe,-::compatible;vwith;.the. rod'-.worth:.minx'mizer,,RSCS
, system"~;and CRDt sequence, requirements Scram'rmes'of the f'our, slowest CRDs. (based'n scram data at A T "rated pressure): will:-be determined't Test Conditions', 3 -6' Caring, planned reactor scrams
ST.. 0 MAIN BODY ST. 0 .3 OBJECTIVES ST.0.2 TEST DESCRIPTION ST.O 3 ACCEPTANCE CRITERIA ST'. 0-4 REFERENCES ST'.0 5 PREREQUISITES ST'..6 PRECAUTIONS ST'0 7, TEST" EQUIPMENT ST.O ..8 PROCEDURE ST 0-A GENERAL APPENDICES S'2' SUBTEST:
I, ST,' D ISCUSS ION:
INITI'AI.
1'T,;Z STATUS'EST'NSTRUCTIONS 2,'T'..X-3'T.X.4 ANALYSIS'PECIFIC ST.X.-A APPENDICES Legend: ST':.Startup Test Number K' 'ubtest. Number A. A'ppendix Designator.
SUSQUEHANNA STEAM ELECTRIC STATION UNITS T AND 2 FINALSAFETY ANALYSIS REPORT'
- STARTUP.'EST,'ROCEDURE, 2
~
', ":STANDARD FORMAT-UNIT,'
FIGURE3'4 ~ 2 2B,
ST.O MAIN BODY ST.O 1 OBJECTIVES ST.0.2 TEST DESCRIPTION ST.O ..3 ACCEPTANCE CRITERIA ST' 4. REFERENCES ST'..0. 5 PROCEDURE ST 0-A GENERAL APPENDICES ST.X ~
.SUBTEST ST X 1 DISCUSSION, ST X'2; PREREQUIS'ITES ST'X 3 INITIAL STATUS STX4 TEST INSTRUCTIONS ST.X.5 SUBSEQUENT ACTIONS ST X.6 GROUP A ANALYSIS ST.X.7 GROUP B ANALYSIS ST X-A SPECIFIC APPENDICES Iegend." ST Startup Test Number X: Subtest Number.
A Appendix Designator
, SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINAL'SAFETYANALYSIS REPORT'TARTUP'EST'ROCEDURE
"., '. STANDARD NORMA.T-UNIT 2 FIGURE. 14. 2-2C.
FIGURE .14 '-5, tr 1
~ ~
Vest ~n )Ieat Vest Cordi/ion (ll Vest Nane Vessel Up 2 3 4 )(arrantly SP-I Chemical S Radioc)mni~l-x(" X X ~ X (2)
ST-2 )Iadiation Heasurpnents ~. X X X 3 Fuel Loading x (2)
SP-4 Full Core Shutdrxrn furr)in x (6)
SP-5 Control Rod Drive (2,3)
~
(3) x (3) x t3~ x (3)
ST-6 SRH Perf. & Control ppd x('~ . x x(
x -..:
SP-7 Reactor Water Cleanup pg, x(7) .. X
~
X(9,13 ST-8 Residual IIeat RerIpvql X SP-9 Pater Level )4easurernet)ts SP-10 IRH Performance ' x x (8)
. ST"ll LPRH Calibratio)r ~(7) X X X SP" 12 APM Calibration X X X X ST-13 Process Car@uter X (9)
'- x(7) X(8'9 ST-15 IIPCI X X X(8I9) X(8)
SP-16 Selected Process ~s X x(") X (14)
(9r 13) x<<',
x, ST-17 System Expansion ST-18 .Tip Uncertainty X X
@~19 Core Performance x X X ST-20 Steam ProductioII (17)
ST-21 ST-22 Core Power-avoid Pressure Regulator
~e pespopse X X X X
X ~ X X
15)
SP-23 Feedwa ter X X X X X (Sr Turbine Valve Survr X X X(8I16)
SZ 24 16)
SP-25 +IVs - ~ '. ' X X X (8, ST-26 Relief Valves II X
PP-27 Turbine Stop Valye Yrip Generator Load Re3ecticr(I x(10)
ST-28 Shutdown Fam Outside Co(Idol poem SP-29 Recirculation FloW Contro) X SI'-30 Recirculation System (1$ ) X ST-31 Loss of T-G & Offsite /carr X ST-32 Containment Atrrpsphere and (13)
Hain Steam Vunnel Cooling X X X(8,13I9)
ST-33 Piping Steady State pibiation X ST"34 Rod Sequence Exc)range x (17)
ST-35 Recirculation System Flow Calibration X X SP-36 Cooling Hater Systems x x SP"37 Gaseous Radwaste System X X X SP-38 BOP Piping System Expansion I4)
ST.-39 Piping Vibration Durirg Spamic Transients ST"$ 0 BOP Piping Steady State Vibration sUsQUEJANNA sTPAIIII ELEG/R(P sfpT(op UNITS 1 ANP 2 flNALSAFETY ANALYSIS REPORT INDIVIDUALSTARJUP TEST SEQUENCE - UNIT 1 FIGURE 14-2-5, Sheet 1
FIGURE 14. 2-5, Sht. 2 Pescriptive Notes:
Figure 14.2-6 for Vest Condition (VC) region map.
(1)
(2)
(3)
See Scae Suhtests reqqi~ to be carpleted prior to fuel load Refer to Table lI),2-$ ,
~ peg p performed during Phase II, (4) Testing merged into PP-l/.
(5) 'esting nerged into ST-33 ~
(6) Hay be done during Open'Vessel Vesting.
(7) Hay be done duriqg )(eatup, Sane Subtests done during approach.to Test (8)
(9) Hay be done during eailier Vest Gondition Done withip steam /@pass capacity.
ifCondition, oorditiops yqrraqg, (10)
(ll) 'dm svmltaneous ttip ojtwo Reactor recirculation pmps is dope af. 100% core flow on the /5% rod line.
(12) Started during approach to Vest Condition 5, continu+ dur~ approach ' to Vest Condition 6.
(13) Sane Subtests done after planned major trips fran 100$ pxw) >
(14) Started duping Test Cagdition 6 and continued during Test Condigcq $ ,
Loss of feedwater heatiqg test done at 80$ power.
(15)
(16)
(17)
Determine maxinaxn Done on 100% rod
~r l~ Subtest can be performed without causipg reactor. scram.
near minimum core flow with recirc pips o)) ~
EUsQUEQAhJQA sTEAM gLEGT()(c (I$$ (og UN+s ) AND 2 FINAL SAFETY ANALyslS ()Epo()T INDIVIDUALSTARTUP TEST SEQUENCE UNIT 1 F(QURE 14.2-5, Sheet 2
E FIGURE 14.2-5, Sht. 3
'gast Heat Vest ConditioII( )
?faI Vest Name pessel Up 3 gxmical S RadiodIemical (2)
X )( X Radiation HeasuzyJaa))+ $ (2) X Puel Loading Bgl Core Shutdo)fn X")
Contml Pod Drily X(2' X(3) pP) X(3> X(3)
ST S+ Performance
~upf Ctml Rod peq. ~~
Reactor Pater Residua) Peat $
+ter
~1 Level Heasuzesjn+
gl7)
X(9i13)
ST-10 S~ s IRH Perfozpance PR4 Calihratioj)
('ontZz)) ~ Seq. X(')
xP) f' X X AP+ Calibratioq ST-1$
Pmcess Ccapute RCIC j X X .
5 "A)7).
(9)
X (8i9)
X(8f9)
Sg-.15 HPCI X X X "4)
'C
~
P7-l(f Selected grooms Tepps
@- 17 System Expansion X(6) 1
~ I X(9f13)
'4~ j, ST-l(I Tip tincertainty N X' ST-19 Core Per fozmaflce I s~ X ST-20 ST-21 PT-22 Sg=23 Core Power-Void Pressuze Regulator Feedwater
~ "
Pespopse-4
~ .)('. -. X X
X X
X'"'(8f15)
SP-24 Turbine Palve Surd; X(8f16)
ST.-.25 HSIVs (8<16)
ST.-26 Relief /elves SZ-.27 Sg-28 Sg-29 Generator ~
Turbine Stop V+ve +ip pe]ect(o)I Shutdown Fran Oqts jde Contmg. palp Recirculation P~ /antral y(P) X 0:
X
~30 Recirculation Syste(a f kf'l,r X (ll) X
~31 Loss of T-G S Offsite Power
~32 Containment Atazf~eze and X(13)
Hain Steam gunnel Coolinq X SZ"33 Piping Steady S+te Vg~tion X X(')
ST=34 hxl Sequence ExdIange 35 Recirculation Systes) F~ Crib).ation SX.36 Cooling )iater Syj~
l)Pp'i: ST-37 ST-38 Gaseous Radwaste System HOP pip~ Syst'aq i~<<l ST-39 Piping Vihratio)I Duripg Synamic Transients X SIt"40 BOP Piping Steady /tate VibratioII8
%USQUE)IANNA STEAM ELECT() (C GATI)flD)t UNCS $ ANDS FINALSAFETY ANAL'YSIS IIEFDIIT INDIVIDUAL STARTUPDIES'g SEQUEI(CE UNIT 2 FIGUAE 14.2-5, Sheet 3
$+s f'IGURE 14.2-5,
~ Sht.~ ()
~ }'.
t3g '( ~: Descriptive Nofes<
(1) See Figure l)I2-(I for est popKtion (VC} region map.
(2)
(3}
(5)
(6)
Refer to Vesting Vesting
~~
SarIe Sub~ts May be dope
~le quired I$ .2-5.
@ Q ccppleted prior to fuel load into SP-17p i})to ST-3).
s Opeq Vessel. Vestry@.
and pay Q perfoppd during Phase l1~
(7}
(8)
(9)
($ 0)
May g dur~
doge May be doJ}e Done
~ wi~ ~
Sane Sub~tS done during HeatupI a~c))
earlier Vest Pediticq steaiq bypass'apacity.
to ~p ifCandition.
conditions warrant)
~;j$
(11)
(12) sinaltaneoI}S trip pone on )(I0$ ~ line ~
of tlap Reactor recixculatio}I pu(rps js piniiialq cori paj with recirc puapj g,p+ImI}I) yoked.
(f3) Scna Subtests pope after plan}Ie4 ma)or trips frcuI 100% po(((er>
$ 00) ppRI /leap on the 75) ~.li}Ie (14) Started dujir)g '(Iest Ca+/i~ (p and coqtinuid during Vest
($ 5) loss of feahaIter heating test done at 80% to 90% power.
~gjj I},
(18) Vesgng ~
(16) petezmine jmdrain power Subtest can be jgfozmd without causgg~c}or scram.
(17( this tsist <<tltTnot he perfornied dnriflg the Unit 2 dt~ertn into sr=10, .
(19) Deleted from Unit 2 FSAR s
~ '~*
Test p~ojree.
,i 4
i so 'i s
's
.sI}gt+e- p p
- s, s
J)/vs f
- 4;,j",...
i ki.- '
srp s Pp s(
Xf,,o
~
sdi'-
PUSOUPI}ANNA fTEAI(}El/CT(Cf/'}}))j s (}IN}TS1 ANQ g FINALSAFFTY ANALYSIS IIPO}IT INDIVIDUAL STARTUT'ES'f SEQUENCE - UNIT 2 Y F}OURg 14.2-5, Sheet 4
400 300 O UNSAFE 200 a 100 SAFE UP TO LIMITED POWER 0 25 50 75 100 PERCENT RATED, POWER 100 SAFE H rn 5 A O H UNSAFE 0 0 25 50 75 100 PERCENT RATED POWER SUSQUEHANNA STEAM'ELECTRIC STATION UNITS. 1 ANO 2. FINAL SAFETY'ANALYSIS'REPORT RCIC ACCEPTANCE CRITERIA,
; CURVES: FOR CAPAC'IVY'ND ACTUATION',TIME FICuaE 14.2-7}}