ML18026A224

Forwards Responses to Questions 423.48-58 Re FSAR

ML18026A224
Person / Time
Site:	Susquehanna
Issue date:	03/30/1981
From:	Curtis N PENNSYLVANIA POWER & LIGHT CO.
To:	Youngblood B Office of Nuclear Reactor Regulation
References
ER100450, PLA-688, NUDOCS 8104020465
Download: ML18026A224 (81)

Text

REGULATOR ')FORMATION DISTR I BUT ION S EM (R I DS)

ACCESS IpN NBR: 810rl020rl65 DOC ~ DATE: 81/03/30 NOTARIZED: NO DOCKET FACIL:50-387 Susquehanna Steam Electr ic Stations Unit- 1~ Pennsyl va 05000387 50 388 Susquehanna Steam Electric Station< Unit 2s Pennsylva 05000388 AUTH, NAME- AUTHOR AFFILIATION CURTIS' ~ Ns Pennsylvania Power 8 Light Co ~

RECIP ~ NAMEI RECIPIENT AFFILIATION YOUNGBLOODiB.J, Licensing Branch 1 I

SUBJECT:

Forwards responses to Questions 423,48-58 re FSAR.

DISTRIBUTION CODE: BOOLS COPIES RECEIVEDlLTR TITLE: PSAR/FSAR AMDTS and Related Correspondence g ENCL L SIZE:

NOTES:Send Send I8E I8E 3

3 copies copies FSAR FSAR 8

8 all all amends'5000387 amends, 05000388 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME. LTTR ENCL ID CODE/NAME LTTR ENCL ACTION: A/D LICENSNG 1, 0 YOUNGBLOODrB 04 0

RUSHBROOKEM ~ 1 0 STARKgR ~ 1 1 INTERNAL: ACCID EVAL BR26 1 1 AUX SYS BR 07 . 1 1 CHEM ENG BR CORE PERF BR EMERG PREP 08 10 22 1

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a I' PP@IL TWO NORTH NINTH STREET, ALLENTOWN, PA. 18101 PHONEr (215) 770-5 Q) s NORMAN W. CURTIS ~~K o+ P Vice Presldenl-Engineering 8 Conslructlon-Nuclear 770.5381 March 30, 198 Mr. B.J. Youngblood Licensing Branch 1 U.S. Nuclear Regulatory Commission Washington, D AC. 20555 Susquehanna Steam Electric Station Docket Nos, 50-387 and 50-388 Responses to guestions 423.48 Thru 423.58 ER100450 File 841-2 PLA-688

Dear Mr. Youngblood:

Attached are the responses to guestions 423.48 Thru 423.58.

Very gcu N.W. Curtis

~~

truly yours, Vice President-Engineering and Construction-Nuclear cc: R.M. Stark bcc: N.W. Curtis W. E. Barberich B.A. Snapp C.T. Coddington R.J. Shovlin I/(

PENNSYLVANIA POWER 8L LIGHT COMPANY Sx o4629~46

SSES-FSAR with the exceptions taken on Regulatory Guide 1.52 in Section 3.13.

Maintenance, of Water Purity in Boiling Water Reactors (June 1973).

Initial Test Programs for Water-Cooled Reactors Power Plants (Revision 1, January 1977).

(1)

Reference:

Section C.1 of the Regulatory Guide.

Testing will be conducted on safety-related structures, systems, and components identified in Table 14.2-1 as required by 10CFR50.

(2)

Reference:

Section C.9 of the Regulatory Guide.

The requirements of Preoperational Test results documentation and reporting are satisfied by the format and content of the completed test procedures; generation of additional reports is not contemplated.

(3)

Reference:

Appendix A, Section l.h(10) of the Regulatory Guide.

Not applicable because SSES does not use containment recirculation fan for post accident containment heat removal.

(4)

Reference:

Appendix A, Section 5.1.1 of the Regulatory Guide. 'Ihe two pump trip is done at Test Condition 3 (approximately 100$ core flow and 75$ power).

1.68.1 Preoperational and Initial Startup Testing of Feedwater and Condensate systems for Boiling Water Reactor Power Plants (Revision 1, January 1977).

Testing may be limited by the availability of.

auxiliary steam.

1.68.2 Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants (January 1977).

1.70 Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants (September 1975).

I 8104020465 14.2" 17

~

t

SSES-FSAR Preoperational Testing of Instrument Air Systems (brune 1974).

The Instrument Air System is not safety related.

However; the various components in the Instrument, Ga's System will be tested to verify that they fail as designed per the statement in Section 3.13.

.The movement of affected. valves will be verified as part of the test associated with each respective valve's corresponding system test.

The action and flow of decay air is not an essential criteria of operation in relation to the affected valves. The valves are to fail with loss of gas to a safe position. Whether decaying pressure will hold some or all of the valves (except for those on the affected line) in normal operating positions is not of critical importance.

Overhead Crane handling Systems for Nuclear Power Plants (February, 1976).

Exceptions for testing of the cranes are outlined in Section 3.13.

Periodic Testing of Diesel Generators Used as Onsite Electric Power Systems at, Nuclear Power

. Plants (August 1977).

The testing of diesel generators will conform to RegulatoryGuide 1.108 per regulatory position 2 a(9).

Since sequence of events capability was not part of the design, testing will also take the same exceptions as outlined in, Section 3.13.

Design, Testing and maintenance criteria for normal ventilation exhaust system air filtration and absorption units of light-water-cooled nuclear power plants (Revision l).

Preoperational testing, will comply with regula'tory position C.5.

14. 2-18

SSES"FSAR 14.2.8 UTILIZATION OF REACTOR OPERATING AND TESTING EXPERIENCE IN THE DEVELOPMENT OF THE TEST PROGRAM The Manager-Nuclear Support is responsible for ensuring that reactor operating and testing=-experiences of similar power plants .

are made. known to the ISG and the Plant Staff during the Initial Test Program. The primary sources of experience information are NRC License Events and experiences of industry contacts. This

14. 2- 18a

SS ES-PSAB andT 14.2-4b)., The subsequent Startup Test Programs are scheduled for six months on each .unit..

+L The'Preoperational.Test Program seguential test schedules 20  :.presented on Figures 14 2-4a and 14.2-4b offer one possible plan for" an- orderly.and.,efficient progression of the program. -Phile these sequences. may be preferred,, numerous alternatives exist The schedule vill be, updated periodically at the jobsite to reflect construction-status manpower availability, and the required test prerequisites The safety-related structures, =systems, and components vill be preoperationally tested. The Preoperational Test Procedures are scheduled to be developed from September 1977 to January 1979 The schedule of Unit 1 and Unit 2 Startup Tests is presented in pigure 14.2-5. This schedule establishes the required testing as a function of test condition. The test conditions are described on 2'Figure 14.2-6 Startup testing vill be divided into three Hajor Test Phases, and within the Power Ascension Test Phase into three distinct test plateaus The testing included in each Hajor Test Phase and test plateau is described in Table 14 2-4 Even though this basic order of testing is required, there is still considerable flexibility in sequencing the startup testing specified to be conducted at each plateau. Detailed. startup testing schedules, commensurate with the reguirements of this schedule, will be developed at the job site

~ ft. W 1~42 1,-Z~lDZVZDDRL TRST DNSCRZPTZONS The individual preoperational tests to be conducted on safety-related structures, systems, and components are listed in Table 14.2-1 . The abstracts of these preoperational tests are contained in Subsection 14.2 12. 1 in numerical order The Startup Test Program procedures are listed in Table 14. 2 The abstracts of Startup Test procedures are contained in Subsection

14. 2. 12 2 in numerical order. The abstracts identify each test.

by title and number, describe the test objectives, specify the test prerequisites, provide a summary description of the test method, and establish the test acceptance criteria Testing of safety-related structures, systems and components will be in conformance with the regulatory guides as stated in Subsection 14.2.7.

REV. 20, 2/81 14 2-20

SSES-FSAR (P59.1) Prima Containment S stem Prep erational Test Test Ob ective .- To demonstrate the operability and isolation capability of the Primary Containment System. Containment isolation valve functional tests will be performed.

to perform this test and the system is turned over to the ISG.

Required instruments are calibrated and controls are operable.

The suppression pool is filled with demineralized water to the required level and the hotwell is available. The Containment Instrument Gas System, Instrument Air System and required electrical power supply systems are available. All prima'ry containment isolation valves are operable.

Test Method -. The suppression pool cleanup system will be tested

'ill for proper operation; the primary containment isolation system have signals. simulated with the valves in the non-isolation position, to verify the primary 'containment isolates when an isolation signal is received. Valve closure times are verified for those valves specified in the FSAR in the various system preoperational tests. The test method is described in the General Test Statement.

Acce tance Criteria - The system performance is in accordance with the applicable design documents.

TP 2.14 Nuclear Boiler System Ievel Instrumentation Verification Test Test Ob'ective - To demonstrate that the nuclear boiler level instruments. function as desired.

'to perform this test and the system is turned over to the ISG.

Required instruments are calibrated and controls are operable.

Required Electrical Power Supply Systems are available. A method to raise and lower the reactor vessel water level is available.

Test, Method - The actual reactor vessel water level will be changed to verify level switch trip points, indicating functions and alarms.

" Acce tance Criteria --The system performance parameters are in accordance with the applicable design documents.

14.2-43 I'p

SSES-FSAR (P59.2) Containment Inte rated Leak Rate Test containment does not exceed the maximum allowable, leakage .rate (La) at the calculated-peak containment internal pressure (Pa),

as defined in 10 CFR50, Appendix J.

including installation of all portions of mechanical, fluid, electrical, and instrumentation systems penetrating containment is complete. Type B and Type C local leakage rate is satisfactorily complete. Required test, equipment instruments and data acquisition systems are operable. Systems required to support the ILRT are operational.

Test Method '- The test shall be conducted in accordance with the requirements of Subsection 6.2;6 of the FSAR.

Acce tance Criteria - Acceptance criteria for this test are in accordance with .the requirements of Chapter 16 of the FSAR.

14: 2- 43a

SS ES-PS AR (P60 1) Containment atmosphere Circulation System Preoperational Atmosphere

-p Circulation'ystem to cool:and circulate air inside the Containment

~pre e u~isi es construction is complete to the extent 'necessary to perform this test and the system is turned over to the ISG.,

Required instruments are calibrated and controls are operable.,

Required electrical power supp3.y systems are'vailable.. The Reactor Building Chilled Water System or an alternate cooling water supply is available.

~rst Method 1he system operation is initiated annually, and flow for each fan is determined.. Required controls are operated or simulated signals are applied to verify; automatic start of standby units and other system interlocks and alarms., No heat loads are simulated during the test.

recce tance -C iteria. The system performance is in accordance with the applicable design documents e Ob ect ves To demonstrate the operability of the Reactor Water Cleanup and Pi3.ter Demineralizer System. In particular the following items are to be demonstrated=

1) The ability o individual components, instrumentations, alarms and interlocks to function properly

2) Verify proper'ystem performance by verifying all. flow paths, flow rates and component performance's to be in accordance with design specifications.

3) The ability of the system and filter to isolate by simulating each .sensor to its trip point

4) Verify the RWCU system containment isolation valves will respond properly to all control signals and closing times are within required specifications.

5) The ability of the filter/demineralizer valve and. pump operating sequence to operate properly, "ere uisites Construction is complete to the extent necessary to perform this test and the system. is turned over to the ZSG The Reactor vessel is fil3ed to provide enough .suction head to REV 17~ 9/80 . 14. 2-44

SSES-FSAR

3) That all warning signals are working per design intent.

4) The capability of the crane to operate in a designated area in accordance with design requirements.

over to the ISG. Requiied electrical power supply systems are available and controls are oparable. Required loads are available to perform load testing.-af this crane.

Test Method - The lighting system for the crane-~r energized and observed for proper operation. The bridge and the trolley ar~

speed-tested in both directions. Current and voltage readings are taken in both directions. The proximity switches are tested for both the bridge and the trolley including trolley movement restriction switches in zones A, B, and .C.

The main hoist and the auxiliary hoist are speed-tested traveling up and traveling down. Current and voltage readings are taken in both directions. All limit switches are tested. A loss of power situation is created for both hoists to check the brakes ability to hold without power. An overspeed test is simulated for each hoist. The main hoist load limit switch is also tested.

The above listed tests are run from the pendant pushbutton control system. Operability of the crane is also demonstrated from the 'cab and by radio control. The anticollision system is tested and the crane power source is verified.

Acce tance Criteria - The system performance parameters are in accordance with the applicable design documents.

TP2.23 REACTOR BULIDING CRANE TESTING OBJECTIVE:

'I To supplement load testing of the reactor building overhead crane.

Construction is complete to the extent required to perform the test, and the crane is available for service.

TEST METHOD:

1. Braking capability of the main and auxiliary hoist under rated load is verified (all -brakes operational).

a4:2-55

J SSES"CESAR

2. The ability of each individual main and auxiliary hoist brake to stop and hold rated, load while lowering at .rated speed is tested.

3. The capability of limiting movement'f the main hook to 1/32" and the auxiliary hook to 1/16" in both raise and lower direction at rated boa'd is tested from a compelte stan@till over an average of ten successive movements.

4. Voltage and current of all crane motors is recorded while running at rated load and rated speed.

>>i',

The capability of the main hoist to limit an uncontrolled drop at rated load and rated speed to less than 1/2" hook movement is verified.

6. Simultaneous bridge and trolley movement at rated load and the ability of the zone proximity switches .to restrict crane movement within safe limits is also verified.

ACCEPTANCE CRITERIA:

All crane parameters are'ithin design'imits.

(P100.1) Cold Functional Test Test Ob ective To demonstrate that the plant systems are capable of operating on .an integrated basis in normal and emergency modes, to demonstrate that adequate power supplies for the class IE equipment'will.exist, and to assure that optimum tap settings have been selected for transformerssupplying power from offsite sources to class IE busses.

completed and plant systems are ready for operation on an integrated basis.

Test Method - Emergency Core Cooling Systems (RHR 6 Core Spray) are lined up in their normal standby mode.. 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 loads are 14.2-55'

SS ES- FS AR adjusted, as practical, to simulate the anticipated ranges of variations. Proper response of the electrical distribution system, diesel qenerators, and ECCS pumps will be verified.

accordance with the applicable design documents

14. 2 12. 2 S~t~uT~gt P~og~a~mP oced~ue Abstracts 20 All those tests comprising the Startup 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 being explored.

Where applicable, a definition of the relevant acceptance criteria for the test, is qiven and is designated either Level or Level 2. A Level l criterion normally relates to the value of l

a process variable assigned in the design of the plant, component systems or associated equipment. If a Level l criterion is not satisfied, the plant will be placed in a suitable hold-condition until resolution i obtained. Tests compatible with this hold----

condition may be continued. Followinq resolution, applicable tests must be-repeated to verify that the requirements of the Level l criterion are now satisfied.

Level 2 criterion is associated with expectations relating to the performance of systems.

satisfied, operating and If testing a Level 2 criterion is not plans would not necessarily be .

altered. Investigations of the measurements and of the

'analyt'ic'al techniques used for the predictions would be started.

For transients involvinq 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 Rad~io hemical Test O~b'ectives 'The principal objectives of this test are a) to secure information on the chemistry and radiochemistry of the REV. 20, 2/81 14 2- 56

SSES-CESAR Test Method Starting at 45 to 65/ power, and continuing at progressively higher power levels, each turbine control,.main stop and .intermediate stop valve will be closed individually and the response of. the reactor will be observed. The margin to scram for reactor, pressure and neutron flux -and the margin to steam line isolation will be plotted for each tested power 'ain level. .These plots will be used to determine the maximum power level at which turbine valve surveillance testing can be performed. The test of the control, main stop, intermediate stop and bypass valves are performed at the predicted highest pow'er level to demonstrate that the Acceptance Criteria are satisfied.

Rate of valve stroking and timing of the close-open sequence will be such that minimum practical disturbance is introduced and that PCIOMR limits are not exceeded.

Acce tance Criteria - Ievel 1 - Not applicable.

Level 2 - Peak neutron flux must be at least 7.5$ below the scram trip setting. Peak vessel pressure must remain at least 10 psi below the high pressure scram se'tting. Peak steam flow in each line, must remain 10$ below the high flow isolation trip setting.

(ST-25) Main Steam Isolation Valves Test Ob ectives - The objectives of this test are (a) to functionally check the main steam isolation valves (MSIVs) for proper operation at selected power levels, (b) to determine

.reactor transient behavior during and following simultaneous full closure of all MSIVs, (c) to determine isolation valve closure time and (d) to determine the maximum power at which a single valve closure can-be made without a scram.

completed. Instrumentation has been checked or calibrated as appropriate.

Test Method - The Main Steam Isolation Valves (MSIVs) are operated during this test to veri'fy their functional performance and to determine closure times. While functionally testing the operation of the MSIVs, the time necessary for closing each individual valve will be noted. The fastest MSIV will then be tested to determine what power level an MSIV can experien'ce fast closure without causing'a scram. All MSIVs will later be used to

" demonstrate a .full isolation subsequently leading to a scram.

(Th'e Nuclear Steam Supply Shutoff System (NSSSS) logic will be used to initiate the f'ull isolation). 'The acceptability of the fast criteria (3 seconds) is determined by utilizing the full stroke time without dealy extrapolated from measured stroke times between 10$ closed and 90$ closed. The acceptability of the slow

14. 2- 81

SSES-FSAR criteria (5 seconds) is determined by utilizing the full stroke time with delay extrapolated for the final 10$ of stroke.

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.

The positive change in simulated heat flux following closure of all MSIVs shall not exceed predicted values by more than 2/ of rated value. Following the closure of'll MSIV's, the reactor must scram to limit'the severity of neutron flux and simulated heat flux transients.

The average of the closure times for the fastest MSIV in each steam line, exclusive of electrical delay, shall not be less than 3.0 seconds.

Closure time for any MSIV, exclusive of electrical delay, shall not be greater than 5.0 seconds.

Closure time for the fastest MSIV'shall be greater than or e'qual to 2.5 seconds.

Feedwater control settings must prevent flooding the main steam lines during the full isolation test.

The time 'delay betwen the close initiation signal and the extrapolated initial valve movement from l00'j open for any MSIV shall be less than or equal to 0.5 seconds.

Ievel 2 - The'ositive 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 referenced to actual test conditions of initial power level and dome pressure and will use beginning of life nuclear data.

The, change in simulated heat flux occurring within the first 30 seconds after the .closure of all MSIVs must not exceed the predicted values.'redicted values will be referenced to actual Test Conditions of initial power level and dome pressure and will use beginning of life nuclear data.

During the MSIV full closure test, the relief valves mu'st reclose properly (without leakage) following the pressure transient.

During .full closure of individual MSIVs, peak vessel dome pressure must gemain at least 10 psi below scram setting value.

During full closure of individual MSIVs, peak neutron flux must remain at least,7.5$ below sciam value.

14:2- 82

SSES-FSAR During full closure of individual MSIVs, steam flow in individual lines, must remain at aleast 10'j below the high flow isolation trip setting.

During full closure of individual HSIVs, the peak heat flux must remain at least 5/ less than its scram value; If water level. reaches the reactor ve'ssel low water level (Level

2) setpoint, RCIC shall automatically initiate and reach rated flow.

14.2- 83

SSES-FSAR (ST-26) Relief Valves Test Ob'ectives - 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.

completed. .Instrumentation has been checked or calibrated as approp'riate. Factory test results on SRV flow and operating times have been reviewed.

Test Method - Testing done at 250 psig reactor pressure consists of 'cycling each relief valve to verify proper operation. The transient monitoring sy'tem will be used to record the results of this test. The data collected will compare the operation of individual relief-valves against 'the operation of all relief valves. During relief valve operation, core power - .and therefore steam generation rate - is maintained constant. The pressure cont'rol system will close'the bypass valves an amount proportional to the relief valve steam flow to maintain constant reactor pressure. Th'is bypass valve. motion will be monitored and a comparison of the response for each relief valve operation will be made. If differences exist, it could suggest a partial obstruction of the relief valve or its tailpipe. Tailpipe temperatu're will be recorded to verify 'the relief valve has properly reseated. Reactor variables will also be'recorded to verify system stability during opening and closing each relief valve..

Testing done at rated reactor pressure consists of manually operating each relief valve at rated reactor pressure. The decrease in Main Generator output will be monitoried during the operation of each relief valve to provide an indication of relief valve flow. By 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 l0 seconds to allow for variables to stabilize.

Reactor variables will also be recorded to verify system, stability during opening and closing each relief valve.

Acce tance 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'esponse. In these cases, the decay ratio for each controlled mode of response must be less than or equal to 0.25.

14:2" 84

SSES-CESAR

.In the scoop tube reset, functions, the speed demand meter must agree with the speed meter within 6g of rated generator speed, along the 100$ rated rod line.

(ST-30) Reciruclation S stem Test Ob'ectives - The objectives of this test are:

a. Obtain recirculation system performance data during pump trip, flow coastdown, and pump 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 performance of the recirculat'ion two pump circuit trip sytem.

d. Verify the adequacy of the recirculation runback to mitigate a scram -upon loss of one feedwater pump.

e. Verify, that no recirculation system cavitation will

'ccur in the operable region of the power-flow map.

compl'eted. Inst'rumentation has been checked or calibrated as appropriate.

Test Method - Single recirculation pump trips will be made at Test Condition (TC)' and TC-6. These trips will be initiated by tripping the M-G Set Drive Motor Breaker from the control room.

Reactor parameters will be'ecorded during the transient and analyzed to verify non-divergence of oscillatory responses, adequate margins to RPS scram set points, and capability of the feedwater system to prevent a high level trip. The capability to restart the recirc. pump at a high power level will also be demonstrated. At TC-3, both recirculation pumps RPT breakers will be simult'aneously tripped using a temporarily. installed test switch. The data gathered will be used to demonstrate acceptable pump coastdown performance prior .to high power turbine trips and generator load rejects.

loss of,a reactor..feed pump will be simulated at TC-3 to

" Ademonstrate the proper operation of the recirculation pump runback circuits. This is done prior to an actual planned feed pump trip.

Both the jet pumps and the recirculation pumps will cavitate at conditions of high flow and low power where NPSH demands are high and little feedwater subcooling occurs. However, the

SSES"FSAR recirculation flow will automatically runback upon sensing a decrease in feedwater flow. The maximum recirculation 'flow is

.limited by, appropriate stops which will run back the recirculation flow from the possible cavitation region. At TC-3, it will be where'ecirculation verified that these limits are sufficient to prevent operation pump or jet pump cavitation occurs.

Acce tance Criteria - Ievel 1 - The response=of any level related variables during a pump trip must not diverge.

The two pump drive flow coastdown transient, during the first 3 seconds of an RFT trip, must be equal to or 'faster than specified.

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 water level margin to avoid a high level trip shall be at least 3.0 inches during the one pump trip.

The data from the two pump drive flow coastdown transient shall be analyzed by G.E. - San Jose to ensure compatability with the safety analysis.

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 Offsite Power Test Ob ectives - The objectives of this test are to demonstrate that the required safety systems will initiate and function .

properly without manual assistance, the electrical distribution and diesel generator systems will function properly, and the HPCI and/or RCIC systems will maintain water level -if necessary during a simultaneous loss of the main turbine-generator and offsite power.

completed. Instrumentation has been checked or calibrated as.

appropriate.

Test Method - With the unit synchronized to the grid at approximately 30$ power, the main turbine-generator will be manually tripped immediately followed by a manual trip of the unit's offsite power source breaker, both trips initiated from 14.2" 90

SSES-CESAR the control room. To ensure a full simulation of the loss of all offsite power to Unit 1 durin'g Unit 1 testing, all Unit 1 and 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 Un'it 2 sources. During Unit 2 testing, to ensure a full simulation of the loss of all offsite power to Unit 2 while minimizing the impact on Unit 1 operations, all Unit 2 loads willie. transferred to Unit 2 Auxiliary and Startup

,busses, all Unit 1 and common loads will be transferred to Unit 1 Auxiliary and 'Startup Busses, 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 will be monitored to verify that the acceptance criteria are satisfied.

The proper response of the electrical distribution system will be checked.

The loss of offsite power condition will be maintained for at least 30 minutes to demonstrate that necessary equipment, controls, and indication are available following station blackout to remove decay heat from the core using only emergency power supplies .and distribution system.

Acce tance Criteria - Level 1 - All safety systems, such as the Reactor Protection System, the diesel-generator, RCIC and HPCI must function properly 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.

Ievel 2 - Not applicable.

14.2- 91

SSES-FSAR (ST-32) 'ontainment Atmos here and Main Steam Tunnel Coolin ability of the drywell. coolers and circulation fans and the Zone I reactor building HVAC system and the main steam tunnel coolers to maintain design conditions in the drywell and reactor building portion of the mainsteam tunnel, respectively, during operating conditions and post scram conditions.

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$ power, data will be taken to ascertain that the containment atmospheric conditions are within design limits.

Acce tance Criteria - level 1 - not applicable Ievel 2 - The general drywell-area is maintained at an average temperature less than or equal to 135 F, with maximum local temperature not to exceed 150 F except for the areas beneath the reactor pressure vessel and around the recirculation pump motors which are specified in succeeding criteria.

The area beneath the reactor pressure vessel is maintained at an average temperature less than or equal to 1350F, maximum local temperature not the exceed 165 F, with minimum local temperature above lOO~F.

The area around the recirculation pump motors is maintained at an average temperature less than or equal to 128 F, with maximum local temperature not to exceed 135~F.

The drywell head area is maintained at temperatures between 1354F and 150~F.

The reactor building protion of the mainsteam. pipeway is maintained at or below 120~F.

14:2- 92

SSES-FSAR TABLE 14.2-2 ACCEPTANCE TEST PROCEDURES Page 1 Test Number 'Test Definition A-3.1 13.8 kV System A-7.1 Lighting System and Miscellaneous 120V Distribution A-8.1 Domestic Water Sytem A-9.1 River Water Makeup System A-9.2 Intake Structure Compressed Air System A-10.1 Screens & Screen Wash System A-ll.l Station Service Water System A-15.1 Turbine Building Closed Cooling Water System A-18.1 Instrument Air System A-19.1 Service Aii System A-20.1 Building Drains - Nonradioactive A-21.1 Water .Pretreatment System A-22.1 Makeup Demineralizer System A-27.1 Auxiliary Boiler System

A-28.2 River Intake Structure H&V System A-28.4 Chlorination Building H&V System A-28.5 Circulating Water Pump House H&V System A-29.1 Administration Building H&V System A"30.3 Control Structure Miscellaneous H&V System A-31.1 Computer A-31.2 Process Computer A-32.1 Security System 125 VDC A-32.2 Security UPS

SSES-FSAR TABLE 14.2-2 CONTINUED Page 2 A"32.3 Security 480 Volt A-32.4 . Security Backup Diesel

~ A-32.5 Security 480/120 Volt A-32.6 Security Bldgs. HGV A-32.7 Security Bldgs. Halon A-33.1 Turbine Building HSV System A-33.2 Turbine Building Chilled Water System A"35.1 Fuel Pool Cooling and Cleanup System A-37.1 Demineralized Water Transfer System A-38.1 Iow. Pressure Air System-A-39:1 Condensate Demineralizer System A-40.1 Lube Oil Transfer, Storage 8 Purification System A-41.1 Cooling Tower System A-42.1 Circulating Water System A-43.1 Main Condenser Air Removal System A-43.2 Condenser Tube Cleaning System A-44..1 Condensate System A-46.1 Extraction Steam System A-65.1 Radwaste Building Air Flow System A-65.2 Radwaste Building Chilled Water System A-68.1 Radwaste Solids Handling System-A-69.2 Liquid Radwaste Subsystems l

A-71.1 .Gaseous Radwaste Recombin er Closed Cooling Water A-72.1 Off-Gas Recombiner System A-74.1 Nitrogen Storage 8 Supply System A-74.2 Bulk Hydrogen System

SSES-FSAR TABLE 14.2-2 CONTINUED Page 3 A-76.2 Process. Sampling- System A-84. 1 Moisture Separators A-85.1 Cathodic Protection System A-85.2 Freeze'rotection System A-91.1 "A'nnunciator System A-92.1 Turbine Steam Seals 6 Drains A-93.1 Turbine Lube Oil Systems A-93.2 Turbine Valves, Valve Test, EHC and Supervisory Systems A-95.1 H2 Seal- Oil System A-97.1 Stator Cooling System A-9811 blain Generator 5 Excitation System "

A-99.2 Communications System

A-99.4 Radiation Area Doors A-99.6 Seismographical Monitoring System

SSES-FSAR TABLE 14.2"3 STARTUP TEST PROCEDURES Test Number Test Definition ST-1 -

Chemical and. Radiochemical ST-2 Radiation Measurements ST-3 Fuel Loading ST-4 Full Core Shutdown Margin ST-5 Control Rod Drive System ST-6 SRM Performance and Control Rod Sequence ST-7 Reactor'Water Cleanup System ST-8 Residual Heat Removal System ST-9 Water Level Measurement ST-10 IRM Performance ST-11 LPRM Calibration ST-12 .APRM Calibration ST-13 NSSS Process Computer ST-14 RCIC System ST"15 HPCI System ST-16 Selected Process Temperatures ST-17 System Expansion ST-18 TIP Uncertainty, ST-19 Core Performance ST-20 Steam Production Verification ST=21 Core Power - Void Mode Response ST-22 Pressure Regulator ST-23 Feedwater System ST-24 Turbine Valve Surveillance ST-25 Main Steam Isolation Valves

SSES-FSAR TABLE 14.2-3 (Cont)

STARTUP TEST PROCEDURES ST-26 Relief VaLves ST-27 Turbine Trip and Generator Load Rejection ST-28 Shutdown From Outside the Main Control Room ST-29 Recirculation Flow Control System ST-30 Recirculation System ST-31 Loss of Turbine Generator and Offsite Power ST-32 Containment Atmosphere and Main Steam Tunnel Cooling "ST-33 Drywell Piping Vibration ST-34 Control Rod Sequence Exchange ST-35 Recirculation System Flow Calibration

'ST-36 Cooling Water Systems ST"37 Gaseous Radwaste System ST-38 BOP Piping System Expansion ST-39 BOP Piping Dynamic Transients ST-40 BOP Piping Steady State Vibration

SSES-CESAR TABLE 14.2-4 MAJOR TEST PHASE-AND TEST PLATEAU SCHEDULE TEST CONDITION SE UENCE Test Test Phase Plateau 'est Condition Se uence Open Vessel Test Condition IV Heatup .Test Condition Test Condition 1 Testing between Test Conditions Test Condition 2 l and 2 Testing between Test Conditions 2 and 3 Test Condition 3 Testing between Test Conditions 2 and 5 Test Condition 5 Testing between Test Conditions 5 and 6 Test Condition 6 Test Condition 4 Warranty (not included in any Test Phase)

SSES-FSAR TABLE 14'.2-5 CONTROL ROD DRIVE SYSTEM STARTUP TESTS Reactor Pressure With Core Loaded Accumulator ps>g Action Pressure 0 600 . 800 Rated Position Indication all Normal Times all 4d Insert/Withdraw Coupling all Friction all all Scram Times Normal all 4- 4M all Scram Times Minimum Scram Times Zero 4M Scram Times Normal Refers to 4 CRDs selected for continuous monitoring based on slow normal accumulator. pressure scram times, or. unusual operating characteristics, at zero. reactor pressure or rated reactor pressure when this data is available; The 4 selected CRDs must be compatible with the rod worth minimizer, RSCS system', and CRD sequence requirements.

Scram times of the four slowest CRDs (based on scram data at rated pressure) will be determined at Test Conditions 2 6 6 during planned reactor scrams.

le Openeat Test Condition (1) No. Test. Name Vessel U 2 3 4 Warrant ST-1 Chemical 6 Radiochemical X X ST-2 Radiation }feasurements X X ST-3 Fuel Loading Full x(2) ST-4 ST-5 ST-6 SR}l Core Shutdown Margin Control Rod Drive Perf. 8 Control Rod Seq. x) X x( X x(3) x(') ST-7 Reactor Water Cleanup X ST-8 Residual Heat Removal X X(14) ST-9 Water Level, Measurements X X X ST-10 IRH Performance X X ST-11 LPRH Calibration X X X ST-12 APRH Calibration X X(4) X X ST-13 Process Computer X X X X ST-14 RCIC X X ST-15 }IPCI X ST-16 Selected Process Temps X (12) X(L2) x02) ST-17 System Expansion X X X X ST-18 TIP Uncertainty X X ST-19 Core Performance X X ST-20 Steam Production ST-21 Core Power-Void Hode Response X H( ST-22 Pressure Regulator H(8) H H>A X H H ST-23 Feedwater X X X X x(,) x ST-24 ST-25 Turbine Valve Surv.

     }ISIV's"                                     x(5) 7'9'1 x(')

X x ST-26 Relief Valves X ST-27 Turbine Stop Valve Trip (7 14) Generator Load Rejection H(13) H(7 >14) ST-28 Shutdown From Outside Control Room ST-29 Recirculation Flow Control H,A ST-30 Recirculation System H( 4) ST-31 Loss of T-G 6 Offsite Power ST-32 Containment Atmosphere and

     }lain Steam Tunnel Cooling ST-33 Drywell Piping Vibration                                                     X ST-34 Rod Sequence   Exchange                                                      X ST-35 Rccirc. Sys. Flow Calib.                                                          X ST-36 Cooling Water Systems                   X         X          X                    X ST-37 Gaseous   Radwaste System               X                    X               X    X ST-38 BOP  Piping System Expansion     X      X                    X                    X ST-39 BOP  Piping Dynamic Transients                                                    x ST-40 BOP  Piping Steady State Vibration                X          X               X    X SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINALSAFETY ANALYSIS REPORT INDLV DUAL STARTUP TEST SEQUENCE FIGURE   14>2-5> Sht>   1

escriptive Notes: Legend: (1) See Figure 14.2-6 for Test Condition region map H - Haster Hanual Flow Control Hode (2) Hay be done during Open Vessel Testing A - Automatic Flow Control Node (3) Refer to Table 14.2-5 X - Test independent of Flow Control (4) Perform the Dynamic System Test Case between Test Condition 1 and 2 (5) Between Test Conditions 1 and 3, 40-50X power, and 60-85$ power (6) Determine maximum power without scram (7) Take BOP piping data, ST-39, during this transient (8) Prior to synchronization (9) Between Test Condition 2 and 3 (10) Between Test Condition 5 and 6 (11) At minimum recirculation pump speed (12) Take data after recirculation pump trips, ST-30 (13) Within steam bypass capacity (14) Take Drywell piping vibration data'T-33, during this transient SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT INDIVIDUALSTARTUP TEST SEQUENCE FIQURE 14 ' 5 ~ She ~ 2

SSES"FSAR

 ,Regulatory Guide 1.68, Revision 1 (January 1977) is the

= applicable guide for your facility. However, Revision 2 (August 1978) which incorporates addit'ional industry and ACRS 'comments provides better guidance than Revision 1. Therefore, we request that you address Revision 2. Our review of your test program description disclosed that the operability of several of the systems and components listed in Regulatory Guide 1.68 (Revision. 2), Appendix A may not be demonstrated by your initial test program. Expand your FSAR to include 'appropriate test descriptions (or modify existing descriptions) to address the following items from Appendix A of the guide: (1) Preo erational Testin l.a(4) Pressure boundary integrity tests. 1.b(3) Standby liquid control system tests; verification of operability of heaters. Demonstration of redundancy, electrical, independence, coincidence, and safe failure on loss of power. 1.d( 1) Turbine bypass valves, 1.d. (3) Relief valves. 1.d. (4) Safety valves. 1.d. (9) Condensate storage system. 1.d. '(11) Cooling water system. 1 ' ~ (5) Steam extraction system. 1 ' ~ (6) Turbine stop, control, bypass, and intercept valves. l.e. (8) Condensate system. l.e (10)

            ~        Feedwater heater and drain systems.

1' ~ (11) Makeup water and chemical treatment systems. l.e. (12) Main condenser auxiliaries used for maintaining vacuum. (1) Circulating water system. (2) Cooling towers and associated auxiliaries'3) Raw water and service water cooling systems. (1) Normal ABC. power distribution system. (2) . Emergency A.C. power distribution system. Tests of structures and equipment (e.g., watertight hatches, walls,,floor drains) that protect engineered safety features from flooding (internal and external)., 1.h.(1)(d) Demonstration of operability of. interlocks and isolation valves provided for overpressure protection for low pressure cooling systems connected to the reactor coolant system. 1.h.(2) Auto depressurization system, including such items as operabilityusing'alternate power and pneumatic Containment post-accident heat removal system testing supplies'.h.(3) of the containment spray nozzles, spray headers; and 423.12-1

SSES-FSAR demonstration that.'piping is free of debris. 1.h(S) Tanks and other sources of water'used for ECCS (e.g., condensate storage tanks and suppression drool). 1.i(1) Containment design overpressure structural tests. l.i(2) Co'ntainmept isolation valve fuactional and closure timing tests. l.i(3) . Containment isolation valve leak rate tests. 1.i(4) Containment penetration leakage tests. 1.i(5) Containment airlock leak rate tests. 1.j. (6) Integrated containment leakage tests. 1.i(7) Main steam line leakage sealing systems. 1.i(g) Primary and secondary containment isolation initiation logic tests. 1.i(9) Containment purge system tests. 1.i(10) Containment vacuum-breaker tests (drywell/wetwell). I.i(13) Containment 'inerting system tests., 1.i(15) Containment penetration pressurization system tests. 1.i(17) Secondary containment system ventilation tests. l.i(19) Bypass leakage tests oa the pressure suppression 4 containment. 1.i(21) Containment penetration cooling system tests. . 1.3 (2) Feedwater control system. 1.3. (7) I,eak detection systems to detect failures in ECCS. 1.3. (9): Pressure control systems used to maintain design differential pressures to prevent leakage across boundaries (feedwater leakage control). 1.3. (10) Seismic instrumentation. 1.3 .(11) Traversing incore probe system. 1.3.(12) Failed fuel detection system. 1.j. (16) Hotwell level control system. 1:3. (17) Feedwater heater tempeiature, level, and bypass control systems.. l.j. (18) Auxiliary'tartup in'strument tests (neutron response checks). 1.3. (19) Instrumentation and controls used for shutdown from outside the control room. 1.3. (21) Reactor mode switch and associated functions. j

1. (22) Instrumentation that can be used to track the course of postulated accidents such as containment wide-range pressure indicators, reactor vessel water level monitors, pressure suppression. level monitors, high-range radiation detection devices, and humidity monitors.

1.3. (24) Annunciators for reactor control and engineered safety features. 1.j. (25) Process computers. X.k. (2) Personnel monitors and radiation survey instrument tests. 1.k.(3) Iaboratory equipment used to analyze or measure 423.12-2

SSES-FSAR radi'ation levels and radioactivity concentrations. High Efficiency Particulate Air (HEPA) filter and charcoal adsorber efficiency.and in-place leak tests. 1.1(2) Gaseous radioactive waste handling systems. 1.1(3) Solid waste handling systems. Solidification system tests should include verification that no free liquids are present in packaged wastes. 1.1.(5) Isolation features for condenser offgas systems. 1.1.(6) Isolation features for ventilation systems. 1.1. (7) Isolation features for,'iquid radwaste effluent systems. 1,1. (8) Plant sampling systems. i.m.(1) Spent fuel pit cooling system tests, including the testing of antisiphon devices, high radiation

          *alarms, and low water level alarms.

1.m. (3) Operability and leak tests of sectionalizing devices and drains and leak tests of gaskets or bellows in the refueling canal and fuel storage pool. 1.m. (4) Dynamic and static load testing of cranes, hoists, and associated lifting and rigging equipment, including the fuel cask handling crane'. Static testing at 125/ of rated load and full operational testing at 100$ of rated load. i.m.(5) Fuel transfer devices. 1.m.(6) Irradiated fuel pool or building ventilation system tests. i.n.(1) Service water cooling system. i.n.(2) Turbine building cooling water systems. i.n. (5) Sampling systems. i.n. (6) Chemistry control systems for the reactor coolant system (condensate demineralizers). i.n. (7) Fire'protection systems. i.n.(8) Seal water systems. i.n.(9) . Vent and drain systems for contaminated or potentially contaminated systems and areas and drain and pumping systems serving essential areas, e.g., spaces housing diesel generators, essential electrical equipment, and essential pumps. 1.n.(11) ~ Compressed gas systems. 1.n.(13) Communication systems. 1.n.(14) Heating, cooling, and ventilation systems serving the following: (a) Diesel generator buildings. (b) Turbine building and radioactive waste handling, building. Shield cooling systems. Heat tracing and freeze protection-systems. Dynamic and static load tests of cranes, hoists, and associated lifting and rigging equipment 423.12-3

SSES"FSAR (e.g., slings and 'strongbacks used during refueling or the preparation for refueling). Static testing at 125$ of rated load and full operational testing at 100% of rated load. 1.o.(2) .Demonstration of the operability of protective and inteilocks. 'evices 1.o.(3) Demonstration of the operability .of safety devices

               ~

on equipment. (2) Xnitial Fuel Loadin and Precritical Tests 2.d. Final test of the reactor coolant system to verify that system leak rates are within specified limits. 2.h. Mechanical and electrical tests of incore monitors, including traversing incore monitors, if installed. (4) Low Power Testin 4.d. Verification that proper operations of associated protective functions and .alarms provide for plant protection in the low-power range. 4.e. Flux distribution 4.g. of proper response of process and measurements'etermination effluent radiation monitors.

   -4.i; Demonstration of the operability of rod inhibit or block functions.

4.1. Demonstration of the operability, including stroke times, of branch steam line valves and bypass valves. 4.m. Demonstration of the operability of main steam line isolation valve leakage control system at hot standby conditions. 4.r. Demonstration of the operability of reactor condensate cleanup system. (5) Power-Ascension Tests 5.a. Demonstration that. power vs. flow characteristics are in accordance with design values. 5,c. Control rod pattern, the exchange demonstration. S.g. Demonstrate that control rod sequencers,. control rod worth minimizers, and rod withdiawal block functionsoperate in accordance with design. 5.1. Demonstrate design capability of turbine bypass valves. 5.m. Demonstrate that the reactor coolant system flows, pressure drops, and vibrations are in accordance with design for various operating modes. 5.o Calibration of instrumentation and demonstration of proper response of -reactor coolant leak detection systems. 5.t. Verify, as appropriate, response times and setp'oints for main steam line relief valves; turbine bypass 423.12-4

SSES-FSAR valves; and turbine stop, intercept, and control valves. 5.u. Uerify response times of'ranch s'team line isolation. 5.wo Demonstrate adequate performance margins for shielding

             -and penetration. cooling systems capable of maintaining temperatures of cooled components within design limits with the minimum design capability of cooling system components available (lOOQ 5.x. Demonstrate adequate beginning-of-life performance margins for auxiliary systems required to support the operation of engineered safety features or to maintain the environment in spaces that house engineered safety features. Engineered safety features will be capable of performing their design functions over the range of design capability of operable components in these auxiliary   systems  (50'j, 100'j).

5.z. Demonstrate that process and effluent radiation monitoring systems are responding correctly. 5.c.c. Demonstrate that gaseous and liquid radioactive waste processing, storage, and release systems operate in accordance with design.

5. f. f. Demonstrate that the ventilation system that serves the main steam line tunnel maintains temperature within the design limits.

5.h.h. Demonstrate that the dynamic response of the plant to the design load swings for,'the facility. 5 I.

       ~  ~ 3. ~     Demonstrate that the dynamic response of the plant is in accordance with design for closure of reactor coola'nt system flow control valves.

5.1.1. Demonstrate that the dynamic response of the plant is in accordance with design requirements for turbine trip.

RESPONSE

Preoperational tests of safety related systems are described. by the test abstracts provided in Subsection 14.2.l2.1. Specific detailed guidelines for testing such a loss of power, air, etc. are described in the startup administration manual Section 7.5. Ioss of power is tested if it causes an evolution to occur within the system such as switching automatically to a different power source. Xoss of air testing is performed .by placing the valve in its non-failed position by normal actuator operation, then isolating the actuator air supply, bleeding off air pressure and verifying valve movement to the failed position. Testing of pumps involves determining pump head and flow for comparison with design values. This testing is performed for all ESF pumps except HPCI, which requires nuclear steam for operation. Each automatic containment isolation valve is tested in the system pre-op test for proper operation and closure timing as required 423.12"5

SSES-FSAR by the design .sections of the FSAR. Leak detection systems such detection'system.'I as steam leak detection are tested in the system pre'-ops .affected by the Each item is answered as follows: l.a(4). - Hydro - All ANSI B31.1, ASME Boiler and Pressure Vessel Code Sections I, III and VIII, NFPA code, and plumbing code piping is hydrostatically tested. Two,primary hydrostatic tests will be conducted on the Reactor Pressure Vessel, recirculation system and main steam lines: ,A primary hydro at 125/ of generating pressure with the internals removed and an operational hydro at 100/ operating pressure with the internals installed. 'I

2. l.b(3) - Updated abstract P53.1 - Verification of chemical mixing and sampling will be done during the Start-up Test Program.

3. l.c - See abstract for F100

           -  See General    Test Statement 4,    l.d('1) -    See abstract for A93.2

5. l.d(3) - See abstract. for P83.1

6. l.d(4) - See abstract for P83.1

7. Should have a'test abstract - will be submitted later for A37.1, Demineralized Water Transfer.

8. 1;d(ll) - Service Water is not safety-related. It is tested by Acceptance Test All.l. The RHR Service Water System is the plant system which falls under section ld of Regulatory Guide 1:68. The RHR Service Water System is tested in Ply 1.

9. l.e(5) - Extraction Steam - See abstract A46.1

10. l.e(6) - Expansion monitoring will be done on'NSSS and the feedwater piping inside containment after fuel load. No other monitoring of BOP systems is anticipated.

ll. l.e(8) - See abstract for A44.1

12. 'l.e(10) - Feedwater Heaters 6 Drain Systems - See abstract A46.1

13. l.e(ll) With the condensate polisher under normal opera ing conditions, Bechtel Corp. will make a complete inspection of all piping and hangers to verify adequate 423.12-6

SSES-FSAR expansion and restraint capability. Test No. A22.1 will be performed to verify correct system operation.

14. l.e(12) - See abstract for A43.1

15. l.f(l) - See abstract for A42.1

16. l.f(2) - See abstract for A41.1

17. l.f(3) - Service Water is not safety-related. It is tested by Acceptance Test All.l.

18. l.g(l) - See abstracts for A3.1, P4.1, P5.1 and A7.1.

19. l.g(2) - See abstracts for A3.1, P4.1, P5.1 and A7.1.

20. l.h - These features are tested under 2 tests:

1) P69.1 - Liquid Radwaste Collection

2) P76 1

               ~
                    - Plant Leak Detection

21. =

l.h(l) (d) - Added to abstract P49.1

22. l.h(2) - See abstract for P83.1.

23. l.h(3) - Demonstrated during flush; not part of P.O. No change.

24. l.h(8) - Proper operation of valve sequencing for ECCS pump suction from the Condensate Storage Tank and 'suppression.

pool is tested in the system preop tests for those systems supplied by water from these systems. Alarms, etc., are tested in A37.1 for the CST and in P59.1 for the suppression.

25. l.i(1) - The containment design overpressure structural test is the Structural Integrity Test performed as a construction test.

26. l.i(2) -- See General Test Statement l.i(2) - Revised abstract for Reactor Water Cleanup .

l.i(2) - Added to abstract P59.1 l.i(2) See abstract for P59.1 27, 28,

29. l.i(3), (4), (5) - The tests covered by these portions of Reg. Guide 1.68 are Type B and Type C local leakage rate tests. The tests are conducted as part of the Component Inspection and Testing Phase. These local leakage rate tests are conducted prior to and as pre-requisites to the Containment Integrated Leak Rate Test. Each Type B and Type C test is conducted in accordance with the requirements of Subsection 6.2.6 of the FSAR. Acceptance criteria for the 423.12-7

SSES-CESAR Type B and Type C tests is in accordance with the requirements of Chapter 16 of the CESAR.

30. l.i(6) See abstract for P59.2

31. l.i(7) - See'bstract for P83.1

32. l.i(8)' Primary containment isolation initiation logic is tested in P59.1. Secondary containment isolation initiation logic is 'tested in P34.1.

33. l.i(9) - See revised abstract for P73.1 34 '.i(10) - See revised abstract for P73.1

35. l.i(13) - Containment inerting testing will be supplied at a later date.

36. l.i(15) - This is not applicable to Susquehanna since leakage surveillance by means of a permanently-installed system with provisions for continuous or intermittent pressurization of individual or groups of is not part of Susquehanna design.

containment'enetrations

37. l.i(17) - See abstract for P34.1

38. l.i(19)'"- See abstract for P59.1

39. l.i(21) -. Not applicable to Susquehanna SES design.

40. l.j(2) - See abstract for P45.2

41. l.j(7) - leak detection for the HPCI (ECCS) and RCIC systems is tested in their respective pre-operational tests. There is no leak detection system for core spray or the containm'ent spray mode of RHR. The leak detection and isolation of the RHR shutdown cooling mode is tested in the RHR pre-op. Overall steam leak detection logic is tested in one of the Main Stream Pre-op's.

42. l.j(10) - Not enough information available to prepare abstract.

43. l.j(ll) - See revised'abstract

- 44. l.j(12) - The off-gas pre-treatment system linear Wide Range Monitor will detect failed fuel and will be tested with other Process Radiation Monitors in P79.2.

45. l.j(16) - See abstract for A44.1 423.12-8

SSES-FSAR

46. l.j(17) -- Feedwater heater temp, level and by-pass control systems See abstract A46.1.

47. l.j(18) - Neutron response: checks, will be part of the Power Test Program. Preoperational testing i's addressed in abstracts P78.1, P78.72, P78.3 and P78.4.

48. l.j(19) - Not a separate system tested in each ECCS System l.j(19) -- See revised abstract for P54.1 l.j(19) Instrumentation and controls used for shutdown from outside the control room are tested under their respective system pre-operational tests.

49. l.j(21) - See abstract for P58.1

50. l.j(22) - Containment instrumentation is tested in the following pre-op tests:

Reactor Wide Range Pressure - P45.1. Feedwater Control Reactor I,evel - P45.1 Feedwater Control and P80.1 Reactor Non-Nuclear Instrumentation Suppression Pool Level - P59.1 Containment and Suppression Radiation Detection = P79.1 Area Radiation Monitoring and P79.2 Process Radiation Monitoring. Humidity Monitors - Not in present Susquehanna SES design.

51. l.j(24) - See'bstract for Annunciator System

52. l.j(25) - See abstract for Process Computer 53.. l.k(2) - See answer below for l.k(3)

54. l.k(3) - Laboratory equipment testing, calibration, etc., is discussed in Subsection 12.5.2 of the FSAR.

55. l.k(4) - Tests 'of HEPA filters and charcoa'1 efficiency are tested by factory representatives on-site prior to performing HVAC pre-op tests.

56. 1 1(2) - See abstract for A72.1

      ~

57. 1.1(3) - See .abstract for A68.1

                       'I

58. l. 1(5) - See abstract for A43. 1

59. 1.1(6) " See abstract for P34.1 and General Test Statement 423.12-9

SSES-FSAR

60. 1.1(7) - Liquid radwaste 'effluent discharge to the environment is tested "in Acceptance Test A69.2.

61. 1.1(8) - Plant Sampling System - Test A76.2 is Process Sampling Test, and tests all the Sample Stations on site.

Test P76.1 is Plant'eak Detection Test and verifies the operability of the leak detection.

62. l.m(l) " See revised abstract, part of TP1.9 for fuel pool

63. l.m(3) - Following erection of the liner plates for the spent fuel pool, dryer, separator pool and reactor basin cavity, the pools are filled with water and left to stand for 48 hours during which leakage is monitored. Helium leak testing is utilized to locate leaks. The pool gates are hydrostatically tested by filling the spent fuel pool 'and monitoring the leakage to the reactor cavity side of the gates.

64. l.m(4) 6 l.o(l) - Present information indicates that cranes and rigging used to .handle fuel and'eactor components will be load-tested (125$ rated load, static lift) at the manufacturer's plant. The pre-operational test will subject the cranes to lifting the heaviest item they are designed to lift (for example, a 700-lb. fuel assembly).

65. l.m(5) - See abstract for P81.1

66. l.m(6) - 'The refueling floor HVAC system is considered Zone 3 of the Reactor Bldg. HVAC system and is tested in P34.1.

67. 1'.n(1) - Service Water is not safety-related at Susquehanna SES. It 'is tested per All.l.

68. l.n(2) - See abstract for A15.1.

'69. l.n(5) -, Reactor Coolant and Secondary Sampling Systems-See abstract A76.2.

70. l.n(6) - System 39 Condensate Demineralizer and Regeneration System is tested under Acceptance Test'39.1 ~ See abstract A39.1.

71. l.n(7) - Fire Protection Systems are tested by Preoperational Tests P13.1 through P13.4.

72. l.n(8) The seal water for the reactor recirculation pumps is supplied by 'the CRD system. The seal water is tested in the Recirculation Pre-op Test P64.1.

73. l.n(9) - Abstract to be written for A20.1 at a later date.

423.12"10

SSES-FSAR

74. l.n(ll) Tested in P25 1

                                ~
.75. l.n(13) - Abstract to    be  written for A99.2, Communications, at a later date.

76. l.n(14),- See abstracts for P28.3, A33.1, A33.2, A65.1 and A65.2.

77. l.n(15) - Not applicable to Susquehanna SES design.

78. l.n(18) - Abstract to be written for A85.1, Freeze J

Protection, at a later date.

79. l.m(4) 8 l.o(1) - Present information indicates that cranes and rigging used to handle fuel and reactor components will be load-tested (125$ rated load, static lift) at the manufacturer's plant. The pre-operational test will OC. probably subject the cranes to lifting the heaviest item they are designed to lift (for example, a 700-lb. fuel assembly).

80. l.o(2) - See answer for 423,12.

81. 1.o(3) - See answer for 423.12.

82. 2.d - Reactor coolant leak. detection systems are placed in .

service and tested per Plant Technical Specifications. These systems are pre-op tested in the appropriate pre-op tests. In addition, an operational hydro of the reactor is performed per Plant Surveillance Tests. Ho change to the test description is required.

83. 2.h - Mechanical tests of the SRM, IRM and TIP drive mechanism are tested in P78.1, 78.2 and 78.4.

The APRM (including LPRM's) system is electrically tested in P78.3. FUrther, ST-6, ST-10, ST-ll, ST-12, and ST-18 demonstrate the overall operability of the nucleag instrumentation systems. As such, no change to the test description is required.

84. 4.d - SRM and IRM alarms are tested in their respective preop, P78.1 and P78.2. The SCRAM function is tested in P78.1 and P78.2, and the Reactor Protection System preop P58.1. No change to the test description is required.

85. 4.e - Flux distributions are not used to verify the items identified ia, section 4.e. Enrichment of the fuel rods and subsequently the fuel bundles is verified by the fuel manufacturer prior to shipping. The required location of the fuel assemblies is verified in ST-3. Proper control rod 423.12"11

SSES-FSAR positioning is verified, in P55.1 and control rod coupling is verified in ST-6.

   ,However, Uncertainty it  should  be noted   that during ST-18 (TIP is performed at Test Condition
                  .    ),  which                                    (TC) 3 and  6, the random'oise, geometric, and total uncertainty          =

of,the TIP trace are determined for an octant symmetrical core and rod pattern. ,Some of the factors which would cause excessive uncertainty are fuel enrichment and/or poisoning errors, improper fuel loading, and mispositioned fuel rods.

86. 4.g - Proper responses of the Area and Process Radiation Monitoring Systems are verified in P79.1 and P79.2 respectively, by using radioactive samples. ST-1 (Chemical and Radiochemical) provides for calibration of monitors in the liquid=waste system and liquid process lines. ST-37 (Gaseous Radwaste) provides for demonstrating proper operation of the Gaseous Radwaste System. .Further, Plant Tech Specs require periodic surveilance of the radiation monitoring systems to ensure proper operation during the appropriate plant conditions.

87. 4.i - The Operation of the Reactor Manual Control System, including RSCS and RWM, is verified in P56.1. These systems are required by Plant Tech Specs to be operable during startup and to demonstrate their operability prior to initiating startup. Therefore, there is not a dedicated startup test which demonstrates their operability. As such, no test description is required.

88. 4.1 - MSIV's are demonstrated operable including stroke times in ST-25 (MSIV) at TC 1,2,3,5 and 6.

Main Steam bypass valves are demonstrated operable, includ ng stroke times, in ST-24 (Turbine Valve Surveillance) at TC 3,5, and 6. Branch Steam Line isolation valves (HPCI, RCIC, and MSIV"LCS) are not tested in a Startup Test. However, they are demonstrated operable, including stroke times, in their surveilance procedures as required by Admin Procedure AD-000-75 (Station Inservice Inspection Progiams). No changes .to the test descript'ions need be made.

89. 4.m - The MSIV-LCS is. initially verified operable in P83.1.

Subsequently, the system is periodically verified operable per surveillance procedures as required by the Technical Specifications. There is no additional Startup Test deemed necessary.

90. 4.r - The RWCU system is partially tested in P61.1. The balance of testing required nuclear heating and is performed in ST-7 (RWCU). No change to 'test description is warranted.

423.12"12

SSES"FSAR

91. 5.a Demonstrations that power vs. flow characteristics are in accordance with design values are done in various Startup Tests (ST.',s) as described below. Refer to Figure 14.2-6 Sheet 1 .for definitions of terms used in the descriptions.

(1) ST-6 demonstrates line b from 0$ to ~25/ power, and line c from ~25$ to the intersection of line c with the 100/ rod line. (2) ST-21 demonstrates the intersect;.-'on of line b with the lOOX rod line. (3) ST-35 demonstrates 'line d atm 50/ and 100'/ power. (4) ST-30 demonstrates line d from ~50$ power to the intersection with the flow interlock line and also demonstrates that cavitation does not occur above and at the flow interlock line. (5). ST-29 performs testing along the 100/ rod line (with Xenon buildup). (6) ST-19 demonstrates that the plant operates below the 100'/ rod line at TC 4, 5, and 6.

92. 5.c Rod sequence exchange is performed in ST-34 (Control Rod Sequence Exchange). 'he test description has been modified.

93. 5.g - See response for 4.i.

94. 5.1 - The design capability of turbine bypass valves is demonstrated in ST-27 (Turbine Trip and Generator Load Rejection). The test description has been modified accordingly.

95. 5.m - The reactor recirculation system is initially and evaluated against desing performance.

tested,'alibrated, parameters in P64.1. During Startup Test Program, the system operating parma'eters are evaluated.,in ST-19 (Core Performance), ST-30 (Recirculation System) and ST-35 (Recirculation System Flow Calibration). Vibration levels for piping in the Recirculation System are evaluated in ST-33'Drywell Piping Vibration). No change to the respective test descriptions is deemed necessary.

96. 5.o - Calibration of instrumentation for reactor coolant leak detection systems is performed during the turnover/checkout phase prior to preoperational testing.

Demonstration of proper instrumentation response is 423.12-13

SSES-FSAR performed during the system's preoperational test. In a'ddition, the reactor coolant system leakage detection systems are periodically verified operated and calibrated as required by Tech Specs. No change to test descriptions is required.

97. 5.t - Main Steam Safety .Relief'Valves are factory tested to verify operability, response times, relieving capacities, setpoints and reseat pressures. Startup Test ST-26 verifies proper SRV.operation and relative relieving capacities.

Periodic surveillance operating tests are conducted to demonstrate SRV operability in accordance with the Technic'al Specifications. For all tests performed in the factory, the test method, test, results and methods of extrapolation (if required) of the data to actual plant conditions are reviewed, documented and retained. Turbine bypass valve operability, response 'time and relieving capacity is quali'tatively verified in the Generator I,oad Reject Within Bypass Valve Capacity test (part of ST-27). Operability is also verified in ST-24. Turbine Stop, Control and Combined Intermediate Valve operability is verified in ST-24. The response times of these valves is qualitatively verified in ST-27.

98. 5.u - Main Steam Isolation Valves are tested for operability and response time in ST-25. Periodic surveillance tests are also performed per Technical Specifications.

Reactor Feedwater Pump Turbine Steam isolation valve is tested for operability in P45.1 RCIC and HPCI steam line isolation valves are tested for operability, and response time in P50.1 and P52.1. Periodic surveillance testing is conducted to verify continued proper response times.

99. 5.w - Not applicable to Susquehanna SES design.

100. 5.x - RBCCW, TBCCW, and Service Water systems are tested in ST-36 (Cooling Water Systems) to verify their adequate performance. The tests are performed, at TC2, 3 and 6. The Containment Atmosphere Circulation System is tested in ST-32 at TC 2 and 6. The HGV systems, for the DG Building, ESSW Pumphouse, Reactor Building, and Control Structure are tested in P28.3, P28.1, P34.1 and P30.1 respectively. The RBCW system is tested in P34.2. The RHR Service Water System is tested in ST8 (RHR System) and is also verified operable in ST-28 (Shutdown From 423.12"14

SSES-CESAR Outside the Main Control Room) . These tests-are performed at TC6 (ST-8) and TCl (ST-28) Emergency Service Water will be tested in ST-36 (Cooling Water Systems) Appropriate test descriptions will be modified as required. 101. 5.z - See response to Item 4.g. J" 102. 5.c.c - Gaseous radwaste system is tested in ST-37 at TC 2, 3, '5, and 6. Liquid Radwaste Collection System is demonstrated operable in P69.1. Solid Radwaste Systems, Liquid Radwaste System, and Gaseous Radwaste Systems are demonstrated operabl ein A68.1, A69.2, and A72.1 respectively. Subsequent operation and surveilance of the Process Radiation Monitoring System verifies proper operation. There are no-. dedicated startup tests for liquid or solid radwaste systems. 103. 5.f.f - ST-32 (Containment Atmosphere and Main Steam Tunnel Cooling) demonstrates the operability of the systems (or portions of systems) which provide cooling for the primary containment and the main s'team tunnel. These tests are demonstrated at TC2 and 6. Refer to revised abstract. 104. 5.h.h - Load swings for the plant, both upward and downward step and ramp changes, are tested in ST-29 (Reciruclation Plow Control) at TC-l, 2, 3, and 5. Plant response to load swings are also demonstrated in ST-30 (Recirculation System) Refer to revised abstracts ST-29 and ST-30. 105. 106. 5.i.i - ST-30 (Recirculation System) tests one-pump trip at TC-3 and 6, and tests a two pump trip at TC-3 only. Reactor coolant flow control valve not applicable on SSES. 107. 5.1.1 - ST-27 (Turbine Trip and Generato'r Load Rejection) tests a turbine trip at TC 3 and tests a generator load rejection at TC6. No change to test description required. 423.12-15

SSES-FSAR 9 4>>. Provide a test description to provide for the integrated testing of reactor vessel isolation on low water level.

RESPONSE

Testing of reactor water level instrumentation will be done during the technical test program. The test will verify level instrument response and setpoints. The actual operation of the various isolation valves are tested in their respective systems and in. the containment system preoperational test P59.1. An abstract of preoperational test P59.1 is found in Section 14.2. A brief abstract of the level setpoint test TP2.14 is found following preoperational test p59.1. 423.23-1

SSES-CESAR g 4 .4 .Your response to item 423.22 states that P30.2 and P28.1 are only test that will be conducted on Unit l,,and not on Unit 2. Modify test descriptions for Pl'3.1 and ST-31 to indicate that testing will be accomplished on both units, or modify your response to item 423.22 to justify not cbnducting P13.1 and ST-31 on Unit .2.

RESPONSE

See the revised abstract for ST-31 for testing of Unit 1 and Unit 2 on loss of turbine-generator and offsite power. The abstract forIpl3..1 will be revised to discuss the reduced scope of testing to be performed on Unit 2 (,deluge systems; dry pipe, wet pipe and preaction systems; hoses in Unit 2 areas).

0 SSES-FSAR 2 NNTINN 423. 49: Your response'o several subitems of 423.12 are not acceptable. Provide the requested information: (1) 64.79 - Modify preoperational test descriptions P81.1 and P99.1 to demonstrate that the refueling grapple and reactor building crane are statically tested at 125$ rated load and dynamically tested at 100/o rated load. (2) 77.99 - Provide a startup test description that will demonstrate that concrete temperatures surrounding hot penetrations do not exceed design limits.

RESPONSE

(1) The reactor building crane was tested at 125/ of capacity by the vendor. Testing was performed on site by construction forces under the vendor's direction. Prerequisites to P99.1 require verification of the 1251. test documentation. Testing at 100% of rated capacity is accomplished during the preopera-tional test program by TP2.23. An abstract of TP2.23 follows P99.1. P99. 1 tests the reactor building crane at maximum critical load in accordance with NUREG-0554.

    =

The refueling bridge main hoist (1200 pound capacity) will be tested to 1254 of capacity utilizing a Technical Procedure. Preoperational Test P81. 1 provides for load limit interlock testing and functional testing utilizing a dummy fuel assembly. The weight of the dummy fuel assembly and the grapple is approxi-mately 950 pounds. (2) The design of hot penetrations includes insulation on the exterior of the process pipe and an air gap between the inside surface of the penetration and outer surface of the pipe insu-lation. Analytical calculations have been performed to provide assurance that the present Susquehanna SES design of the hot penetrations will be able to maintain the concrete temperatures around these penetrations below the design limit. ST-32, "containment atmos-phere and Main Steam Tunnel Cooling," demonstrates that the tempera-ture of the atmosphere inside the drywell is maintained within design limits. With the reactor at rated temperature during the drywell inspection (described in ST-17) a check will be made to estimate the concrete temperature surrounding one of the main steamline penetrations by measuring the temperature at several accessible points on the containment liner plate or containment concrete surface.

SSES-FSAR UESTION 423. 50 Provide testing to verify that containment spray nozzles and haders, are free of debris by testing. If this testing is not performed with worker in conjunction with testing the pumps, verify that the flow path for this testing overlaps the flow path used when testing the pumps.

RESPONSE

P49. 1 (RHR'ystem Preoperational Test) provides for testing of the containment spray nozzles. This test consists of connecting a streamer to each spray nozzle and connecting a source of service air to the system and verifying that the nozzles are not plugged by observing air flow and streamer movement. System flow through the containment spray header was verified during TP 3.25 (RHR System flush) by connecting hoses between the two loop spray headers and flushing from one loop into the other and back to the suppression pool. . Bench testing of a sample of spray nozzles will be accomplished in P49. 1 in accordance with FSAR Subsection 6.2.2.2.

SSES-FSAR QUESTION 423.51: Provide or modify test descriptions that will verify that the emergency ventilation systems are capable of maintaining all ESF equipment within'heir design temperature, range with the equipment operating in a manner that will produce 'the maximum heat load in'he compartment. If 'it is not practical to produce maximum heat loads in a compartment, describe the methods that will be used to verify design heat removal capability of the emergency ventilation systems. Note will be that it is not apparent that post-accident design heat loads produced in ESF equipment rooms during the power ascension test phase; therefore, simply assuring that area temperatures remain within design limits during this period may not, in itself, demonstrate the design heat removal -capability of ,these systems. It may be necessary to measure air and cooling water temperatures and flows and to extrapolate to verify that the ventilation systems can remove the postulated post-accident heat loads.

RESPONSE

ESF equipment room coolers were performance tested by the vendor to demonstrate conformance to design criteria. The preoperational test will provide for measurement of air and cooling water temperatures and flows to confirm room cooler performance. Comparison of cooler efficiency with vendor data will be made to assure that cooler performance at maximum heat load is acceptable.

SSES-FSAR g 4>>. .Modify ST-30 to indicate .that a simultaneous trip of both recirculation pumps will be performed at test condition 6 or provide technical justification, in .Subsection 14.2.7 for taking exception to .Regulatory Guide l.'68 (revision '1, 1/77), Appendix A, 5.1.1.

RESPONSE

On earlier plants, where MCHFR was used to determine reactor thermal margin, the two pump trip was performed since MCHFR was very sensitive to core flow. When GE developed the GEXL correlation, which establishes MCPR for determining reactor thermal margin for current plants, it was found that MCPR is relatively insensitive to core flow. When the effect of the two pump trip on the reactor thermal margin- was determined to be minor, the test was generically deleted from BWR Startup Test Programs. At Susquehanna, the two pump trip -is'done at Test Condition 3 (approximately 100/ core flow and 75$ power) not to determine the effects of core flow upon MCPR but to verify acceptable performance of the recirculation two pump circuit trip system and to demonstrate acceptable pump coastdown performance prior to high power turbine trips and generator load rejects.

~ t SSES-FSAR Modify ST-31'o provide as'surance that the loss of offsite p'ower condition will be maintained for at least 30 minutes to demonstrate that necessary equipment, controls, and indication are available fol'lowing .station blackout to remove decay heat from the core usi'ng only emergency power supplies and distribution ' systems.

RESPONSE

Test description for ST-31 has been modified to maintain the loss of offsite power condition for at least 30 minutes. See revised abstract for ST-31.

SSES-FSAR QUESTION 423.54:- Include the test description (TP2.14) provided as a response to item 423.23.in the FSAR, Subsection .14.2.12.

RESPONSE

Test description for TP2.14 has been removed from Question 423.23 and placed in FSAR Subsection 14.2.12 for the test abstract of Preoperational Test g59.1.

SSES-FSAR r Revise Subsection 14.2.12 to incorporate responses to items 423.37 and 423.38.

RESPONSE

Subsection 14.2. 12 has been revised to incorporate the responses to . questions 423.37 and 423:38

f f 1' g 4I 1'

 'N

} SSES-FSAR QUESTION UEU.US: Your responses to items 423.32 and 423.45 reference a revised response to item 423. 12. Provide this revised response, or revise your response to items 423.32 and 423.45 to provide the requested information.

RESPONSE

guestion 423. 12 has been revised to incorporate the responses to guestions 423.32 and 423.45.

~. V I

'v1 SSES-FSAR g 4 ST-25 to address the following: I'odify r (1) The 'present method for determining MSIV closure times 'is Modify the test method to measure the'ull 'naccurate. travel of the valves or provide technical justification for extrapolating the full closure 'time when only measuring 90 percent closed, plus the period from 10 percent closed to 90 percent closed times 1/8, or provide technical justification for the current method which "double-counts" delay '. (2) Provide a description of a test which demonstrates that the MSIV-ICS components operate properly when handling steam and that the system can handle the amount of leakage that is present when the.main steam system is at operating temperature.

RESPONSE

(1) ST-25 provides for determination of MSIV closure times as. described below:

       .MSIV    closure time must meet divergent criteria'he. valves
        .must   close, fast enought to limit the release of reactor coolant, and they must close slow enough so th'at simultaneous closure of all steamlines will not induce transients that exceed .the nuclear steam design closure time i.s calculated using limit switches which limits'SIV actuate when valve stem travel indicates 10/ and 90$ valve closure. Extrapolations using this data assumes linear valve closure.

r Two equations are necessary to accurately calculate elapsed, times'he slow criteria equation must include the delay time from solenoid deenergization to valve stem movement, whereas the fast criteria equation, which is concerned only with valve movement, does not include this delay. The two equations are: (1) for fast criteria T (T90 T10} + 0 ~ 25 (T T') (2) for slow criteria T = T + 0 1 T d 0 Tcwd " T90 + 0.125 (T90 T10)

~4 I'

     ~

where: T . = valve closure time, excluding delays Tcwdd = valve closure time, with delays T90

         = elapsed, time from solenoid deenergization to valve 90$ 'closed T>10
         = elapsed   time .from solenoid deenergization to valve 10$ closed.

(2) The MSIV-LSC is designed to control and minimize the release of fission products which could leak through closed MISV's fdllowing a LOCA. The MSIV-I,SC is initially verified operable in g83.1 using air and subsequently verified operable on a periodic basis in accordance with Technical Specification,. Complete system testing and isolation valve leak testing is performed only during'eactor"shutdown to preclude inadvertent steam discharge. Interlocks are, provided to preclude system operation at excessive MISV leak rates. No further Startup Testing is deemed necessary.

SSES-FSAR Update Table 14.2-3 {Startup Test Procedures) and Figure 14.2-5 (Individual Startup Test Sequence) to .reflect the current status-of Subsection 1'4.2.12.'2.

RESPONSE

Table 14.2-3 and Figure 14.2-5 has been revised to reflect the current status of Subsection 14.2.12.2.

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