Text

1 to Updated Final Safety Analysis Report, Questions and Responses 423.1 Through 423.58
	ML23292A215
Person / Time
Site:	Susquehanna
Issue date:	10/12/2023
From:	; Susquehanna
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML23291A105	List: ML23291A113; ML23291A342; ML23291A343; ML23291A344; ML23291A369; ML23291A370; ML23291A371; ML23291A372; ML23291A373; ML23291A374; ML23291A375; ML23291A376; ML23291A377; ML23291A378; ML23291A379; ML23291A380; ML23291A381; ML23291A382; ML23291A383; ML23291A384; ML23291A385; ML23291A386; ML23291A387; ML23291A388; ML23291A389; ML23291A390; ML23291A391; ML23291A392; ML23291A393; ML23291A394; ML23291A395; ML23291A396; ML23291A397; ML23291A398; ... further results
References
	PLA-8081
	Download: ML23292A215 (1)
	v • d • e

SSES-FSAR QUESTION 423.1 Provide minimum education and experience requirements, at the time of assignment to the function, for: (1) Personnel assigned to conduct preoperational tests (test directors); (2) personnel assigned to conduct startup tests (test directors}; (3) personnel assigned to the group responsible for review of preoperational test procedures and results (Test Review Board members), and (4) personnel assigned to the group responsible for review of startup test procedures and results (Plant Operating Review Committee}.

RESPONSE

(1) The minimum qualifications for Preoperational Test Directors at the time of test performance are:

Bachelor's degree in engineering or the physical sciences or

High School graduate and four years experience in related testing or operations (or both) of power plants, nuclear facilities or similar industrial installation. Up to two years of this experience may be replaced on a one-for-one basis by successfully completed technical training time in a recognized associated degree program 4 and

One year of applicable nuclear power plant experience consisting of:

Test procedure preparation

Component initial checkout and testing during Technical Test Phase

Initial system operation

System flushing and initial integrated system operation

Documentation of the above applicable activities per approved Technical Test Procedures

Attendance at any of the following courses as determined by the ISG supervisor

Susquehanna Technology (General Physics)

Rev. 46, 06/93 423.1-1

~** - * - * - * - * * - * * -.

SSES-FSAR

BWR Fundamentals (General Electric)

BWR Design Orientation (General Electric)

BWR Technology (General Electric)

Training seminars on Quality Note that while the Test Director is responsible for directing the preoperational tests, the test procedures and any changes thereto are reviewed by the Test Review Board.

Test results are also reviewed and approved by the Test Review Board.

(2) The minimum qualifications of Startup Test directors at the time of test performance are:

Bachelors degree in engineering or the physical sciences or the equivalent and

Two years of applicable power plant experience of which at least one year shall be applicable nuclear power plant experience.

Note that startup test performance is coordinated by the Test Director; however, the actual manipulation of plant equipment is done by or under the direct supervision of the PP&L Shift Supervisor, Assistant Shift Supervisor or Plant Control Operator all of whom will be NRC licensed individuals. The test procedure and any revisions thereto and test results must be approved by the Plant Operations Review Committee.

(3) The minimum qualifications of Test Review Board (TRB) members are as follows:

Personnel assigned to the Test Review Board shall possess that combination of education and experience recommended in ANSI NlS.1-1971 for the position of Operations Manager (Section 4.2.2) or that combination specified in Subsection 14.2.2.2.1, ISG Supervisor.

(4) The qualifications for Plant Operating Review Committee (PORC) members are listed in FSAR Subsection 13.1.3 as referenced in Subsection 13. 4 .1.1 since this is a permanent plant committee.

Rev. 46, 06/93 423.1-2

SSES-FSAR QUESTION 423.2 The description of your planned degree of conformance with certain regulatory positions contained in Regulatory Guide 1. 68 requires clarification and modification. Regulatory position C.1. describes criteria for selection of plant structures, systems, and components to be tested. Further, Appendix A to the guide provides a representative list of such structures, systems, and components that should be considered for preoperational testing and startup testing. The regulatory cases for this testing includes both Criterion I of Appendix A to 10 CFR 50 and Criterion XI of Appendix B to 10 CFR 50. Your reference to Table 3.2-2 in the FSAR is not acceptable and, therefore, your response should be modified. Further, your categorization of "acceptance tests" should be modified to identify these tests as preoperational tests and the specific controls that will govern the review and approval of test procedures and test results and the conduct of tests for this category should be described. Also, state your plans for review of the results of the tests currently listed as "acceptance tests" prior to fuel loading and provide teit abstracts for each test that will identify the test objectives, test methods, and acceptance criteria. Your classification of "power tests" should be modified to "startup tests" to achieve consistency with the terminology used in Regulatory Guide 1.68 and to avoid anticipated interpretation problems between your plant staff and the I&E inspection staff.

RESPONSE

Testing of safety-related structures, systems and components will be done according to Table 14.2-1, Preoperational Test List. Subsection 14.2.7 of the FSAR has been revised to reference the correct table.

The Preoperational Tests are performed on safety-related equipment. The Acceptance Tests are similar to the Preoperational tests in format, preparation, review and approval. The only difference is that Acceptance tests are done on equipment other than that which is safety-related.

While it is our intention to perform all the Acceptance Tests identified on Table 14.2-2, it is not a requirement that they be performed.

The term "power tests" was adopted by PP&L at the beginning of the startup program development; however, we will change the term "power tests" to "startup tests" as defined in Reg. Guide 1.68. Existing PP&L documents will be modified by Dec. 31, 1979 to reflect this change.

Rev. 46, 06/93 423.2-1

SSES-FSAR QUESTION 423.3 State the approximate number of test personnel that will be assigned to augment the plant staff (e.g., the integrated startup group) and the approximate schedules (relative to fuel loading} for assignment.

RESPONSES:

The Integrated Startup Group plans to have approximately forty-nine (49) engineers for the Initial Test Program in addition to the Plant Staff people. A schedule for this estimated manpower loading is attached.

Rev. 46, 06/93 423.3-1

-t-+++-+++-1 -!lU~B-..::t::t:t:+t=t:t=t=n+H

-~t-++-+MANP-t-t--rt---+-tOWER CUR~- FOR SU5?~~At:t:-++tu: _-++i::s::-t--t-::t:t

~ ~

-l-+++-H-44 - - ,.._

I- * - - - -

UNIT 244-f-.-.t+t++-H-N

-+/-; 1-

-- *- --- ,. _ ~~ ---.l1 l::1~

1-t -

~; *-

~

.TOT~ SfAR11.JP EFFORT4-I

~- - * -

.. . -+++-+++-l_l-+-+-+-rl PP&L STAR11JP 11-+-+-o-+-1-<-+++-

BECHTEL SI'AR1lJP ENGRS.

- -:_f H U U i U f hUU, l IUi::Uil~

ii~ O ! . i I

u~ i I l u~ nu* Uii i *I S U ~ i 1'-1:l 19-1lL 19LL I 19.J ~ 19~ lt.tiU.. . l9BL 111~ I FSAR REV.65 SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 & 2 FINAL SAFETY ANALYSIS REPORT INTEGRATED STARTUP GROUP ESTIMATED MANPOWER LOADING CURVE FIGURE 423.3-1, Rev 47 AutoCAD: Figure Fsar 423_3_1.dwg

SSES-FSAR QUESTION 423.4 There appears to be a discrepancy between FSAR Sections 14.2.2 and 14.2.3. To clarify this issue provide a clear statement regarding the Test Review Board responsibility for review of startup test procedures (Phases III, IV, & V).

RESPONSE

To clarify the duties of the TRB, their participation in the Startup Test Program is to respond to any review requests made by the Plant Operations Review Committee (PORC).

The TRB will not review startup test procedures or results unless requested by PORC whose responsibility it is to recommend approval of these procedures and results to the Superintendent of Plant.

Rev. 46, 06/93 423.4-1

SSES-FSAR QUESTION 423.5 Describe your controls to assure that plant modifications and repairs identified as a result of plant testing are reviewed, approved, and completed and to assure retesting following such work is completed.

RESPONSE

Historically, any plant modification or repair done on a system was done under a Start-up Work Authorization (SWA)--Start-up Administration Manual Procedure AD6.4, or a Work Authorization

{WA), Plant Administration Procedures Manual Procedure AD-00-046. The current Work Authorization Procedure is NDAP-QA-0502. All these procedures address review, approval, completion and post-work testing involved with modifications and rework. These procedures are available on site for NRC review.

Rev. 53, 04/99 423.5-1

SSES-FSAR QUESTION 423,6 Describe your provisions for the retention of test records (note: N45.2.9).

RESPONSE

Retention of test records is addressed in Start-up Administrative Manual Procedures AD 7.6--

Preoperational/Acceptance Test Procedure Control and AD 3.3--

Start-up Filing Control.

Material filed in any startup file is not indiscriminately removed. A standard nout" card system is used to indicate removal of material from any startup file. The documents will be kept in fire-resistant, lockable cabinets and are considered QA records. These records will be transferred to the permanent plant filing system at some time at the completion of the Initial Test Program.

Rev. 46, 06/93 423.6-1

SSES-FSAR QUESTION 423.7 You state in response to Question 423 . 2 that testing of safety-related structures, systems, and components will be done according to Table 14.2-1 and that Subsection 14.2.7 of the FSAR has been revised to reference the correct table. However, you state in 14.2.7 regarding conformance to Regulatory Guide 1.68 that "testing will be conducted . . . identified in Table 3.2-2." Please correct this inconsistency to conform to the position stated in Question 423.2.

RESPONSE

Subsection 14. 2. 7 has been revised to reference the proper table (Table 14.2-1).

Rev. 46, 06/93 423.7-1

SSES-FSAR QUESTION 423.8 Your position to have approved test procedures available for NRC review at least 30 days prior to intended use is not acceptable. Revision 1 to the Standard Review Plan has been revised to, among other items, state that these procedures should be suitable for review at least 60 days prior to their intended use. Revise Section 14.2.7 to be consistent with this position.

RESPONSE

See revised Subsection 14.2.3 Rev. 46, 06/93 423.8-1

SSES-FSAR QUESTION 423.9 The information provided in Section 14.2.4.3 states that "if necessary, procedures may be modified to complete testing."

This implies that the tests may not be conducted in a manner consistent with that described in your FSAR. Your application should be modified to provide a clear commitment that the tests will be conducted as described or that the FSAR will be modified to reflect identified changes.

RESPONSE

Subsection 14.2.4 addresses administrative procedures.

Subsection 14.2.4.3 addresses test procedures. There is no discrepancy with respect to test changes. Changes to test procedures will not change the intent of the procedure. Any change in FSAR commitment will be proceeded by an FSAR revision.

Rev. 46, 06/93 423.9-1

SSES-FSAR QUESTION 423.10 Provide test abstracts for the acceptance tests shown in Table 14.2-2 except for the following:

A-3.2 Station Ground System; A-8.1 Domestic Water System; A-9.1 River Water Makeup System; A-9.2 Intake Structure Compressed Air System; A-10 .1 Screens and Screen Waste System except for the Emergency Service Water System; A-20.1 Building Drains-Nonradioactive except for those in the ESF equipment rooms; A- 21. 1 Water Pretreatment System; A-27.1 Auxiliary Boiler System; A-28. 2 River Intake Structure H&V System; A-28. 4 Chlorination Bldg. H&V System; A-28.5 Circulating Water Pump House H&V System; A-29.1 Administration Bldg. H&V System; A-29.2 Administration Bldg. Chilled Water System; A-37.1 Demineralized Water Transfer System; A-43.2 Condenser Tube Cleaning System; A-74.2 Bulk Hydrogen System; A-85.1 Cathodic Protection System; A-95.1 H Seal Oil System; A-97.1 Stator Cooling System; A-98.1 Main Generator and Excitation System; and A-99.4 Personnel Access Monitors.

Note: We consider that the acceptance tests, except as noted above, should require the same reviews and approvals as your Phase II tests. Modify your FSAR to include these administrative controls.

RESPONSE

For response see Subsection 14.2.12.3.

Rev. 46, 06/93 423.10-1

SSES-FSAR QUESTION 423.11 You state that in your testing of containment recirculation fans that it may be possible that they will not be tested to verify that fan motor current is within design. Provide a description of how you plan to verify fan motor currents at conditions representative of accident conditions or provide technical justification for not conforming to the regulatory guide position.

RESPONSE

See revised discussion on Regulatory Guide 1.68, Appendix A, Section 1.h(l0) given in Subsection 14.2.7.

Rev. 46, 06/93 423.11-1

SSES-FSAR QUESTION 423.12 Regulatory Guide 1. 68, Revision 1 (January 1977) is the applicable guide for your facility. However, Revision 2 (August 1978) which incorporates additional industry and ACRS comments provides better guidance than Revision 1. Therefore, we request that you address Revision 2. Our review of your test program d~scription disclo~ed 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 appropria~e. test descriptions (or modify existing descriptions) to *address the following items from Appendix A of the guide; (1) Preoperational Testing l.a(4) Pressure boundary integrity tests.

l.b(3) Standby liquid control system tests; verification of operability of heaters.

Le Demonstration of redundancy, electrical, independence, coincidence, and safe failure on loss of power.

l.d(l) Turbine bypass valves.

1.d(3) Relief valves.

1.d(4) Safety valves.

l.d(9) Condensate storage system.

l.d(ll) Cooling water system.

l.e(S) Steam extraction system.

l.e(6) Turbine stop, control, bypass, and intercept valves.

1.e (8) Condensate system.

1.e(l0) Feedwater heater and drain systems.

1.e{ll) Makeup water..and chemical treatment systems.

l.e(12) Main condenser auxiliaries used for maintaining vacuum.

1. f {1) Circulating water system.

Rev. 46, 06/94 423.12-1

SSES-FSAR l . f (2) Cooling towers and associated auxiliaries.

1. f (3) Raw water and service water cooling systems.

l.g(l) Normal A.C. power distribution system.

1. g (2) Emerge_ncy A.C. power distribution system.

1.h Tests of structures and equipment (e.g.,

watertight hatches, walls, floor drains) that protect.engineered safety features from flooding (internal and external).

1.h(l) (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 operability using alternate power and pneumatic supplies.

1.h(3) Containment post-accident heat removal system testing of the containment spray nozzles, spray headers; and demonstration that piping is free of debris.

1.h{8) Tanks and other sources of water used for ECCS (e.g., condensate storage tanks and suppression pool).

1.i(l) Containment design overpressure structural tests.

1. i (2) Containment isolation valve functional 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.i(6) Integrated containment leakage tests.

1. i (7) Main steam line leakage sealing systems.

1.i(8) Primary and secondary containment isolation initiation logic tests.

Rev. 46, 06/94 423.12-2

SSES-FSAR l.i(9) Containment purge system tests.

1.i(lO) Containment vacuum-breaker tests (drywell/wetwell).

Li(l3} 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 on the pressure suppression containment.

1.i(21) Containment penetration cooling system tests.

1.j (2) Feedwater control system.

1. j (7) Leak detection systems to detect failures in ECCS.

1.j(9) Pressure control systems used to maintain design differential pressures to prevent leakage across boundaries (feedwater*leakage control}.

1.j(lO) Seismic instrumentation.

1. j (11) Traversing incore probe system.

1. j (12) Failed fuel detection system.

1. j (16) Hotwell level control system.

1.j (17) Feedwater heater temperature, level, and bypass control systems.

1. j (18) Auxiliary startup instrument tests (neutron response checks).

l.j(19) Instrumentation and controls used for shutdown from outside the control room.

1.j (21) Reactor mode switch and associated functions.

Rev. 46, 06/94 423 _1.2-3

SSES-FSAR 1.j(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.j(24) Annunciators for reactor control and engineered safety features.

1. j (25) Process computers.

1.k(2) Personnel monitors and radiation survey instrument tests.

1.k(3) Laboratory equipment used to analyze or measure radiation levels and radioactivity concentrations.

1.k(4) High Efficiency Particulate Air (HEPA) filter and charcoal absorber 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 offgaa systems.

1. 1 (6) Isolation features for ventilation systems.

1.1(7) Isolation features for liquid radwaste effluent systems.

1.1(8) Plant sampling systems.

1.m(l) 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 ..

Rev. 46, 06/94 ** 423 .12-4

SSES-FSAR l..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.

l.m(S) Fuel transfer devices.

1.m(6) Irradiated fuel pool or building ventilation system tests.

l.n(l) Service water cooling system.

1. n (2) Turbine building cooling water systems.

l.n(S) Sampling systems.

1.n(6) Chemistry control systems for the reactor coolant system (condensate demineralizers).

l.n(7) Fire protection systems.

l.n(8) Seal water systems.

1.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.

l.n(ll) Compressed gas systems.

l . 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.

l.n (15) Shield cooling systems.

1.n (18) Heat tracing and freeze protection systems.

Rev. 46, 06/-g4 423.12-5

SSES-FSAR 1.o(l) Dynamic and static load tests of cranes, hoists, and associated lifting and rigging equipment (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 devices and interlocks.

l.o{3) Demonstration of the operability of safety devices on equipment.

(2) Initial Fuel Loading 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 Testing 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 measurements.

4.g Determination of proper response of process and 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.

Rev. 46, 06/94 423.12-6

SSES-FSAR (5) Power-Ascension Tests 5.a Demonstration that power vs. flow characteristics are in accordance with design values.

Control rod pattern, the exchange -

demonstration.

5. g Demonstrate that control rod sequencers, control rod worth minimizers, and rod withdrawal block functions operate 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.

s.o Calibration of instrumentation and demonstration of proper response of reactor coolant leak detection systems.

5.t Verify, as appropriate, response times and set points for main steam line relief valves; turbine bypass valves; and turbine stop, intercept, and control valves.

5.u Verify response times of branch steam line isolation.

5.w Demonstrate adequate performance margins for shielding and penetration cooling systems capable of maintaining temperatures o-f cooled components within design limits with the minimum design capability of cooling system components available (100%).

5 .x .. Demonstrate adequate beginning-of-life performance margins for auxiliary systems requi:r:ed 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 Rev. 46, 06/94 423.12-7

SSES-FSAR components in these auxiliary systems (50%,

100%) .

5.z Demonstrate that process and effluent radiation monitoring systems are responding correctly.

S.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 des-ign load swings for the facility.

5.Li Demonstrate that the dynamic response of the plant is in accordance with design for closure of reactor coolant system flow control valves.

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

RESPONSE

Preoperational teats of safety related systems are described by the test abstracts provided in Subsection 14.2.12.1. Specific detailed guidelines for testing such a loss of power, air, etc.

are described in the startup administration manual Section 7. 5.

Loss of power is tested if it causes an evolution to occur within the system such as switching automatically to a different power source. Loss 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. Each automatic containment isolation valve is tested in the system pre-op test for proper operation and closure timing as required by the design sections of the FSAR. Leak detection systems such as steam leak detection are tested in the system pre-ops affected by the detection system.

Each item is answered as follows:

Rev. 46, 06/94 423.12-8

SSES-FSAR

1. 1. a (4) - Hydro - All ANSI B31. l, ASME Boiler and .

Pressure Vessel Code Sections I, III and VIII, NFPA code, and plumbing code piping is hydrostatically tested. Two primary hydrostatic tests are 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.

2. 1.b(3) - Verification of chemical mixing and sampling will be covered by the Technical Specification Surveillance requirements per 4.1.S.

3. l.c - See abstract for PlOO

- See General Test Statement

4. 1.d{l) - See abstract for A93.2

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

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

7. 1.d(9) - See abstract for A37.1

8. l.d(ll) - Service Water is not safety-related. It is tested by Acceptance Test All. 1. The RHR Service Water_ .

System is the plant system which falls under section 1.d of_ Regulatory Guide 1. 68. The rum. Service Water System is tested in P16.1.

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

10. 1.e(6) - Expansion monitoring is done on NSSS and the feedwater piping inside containment after fuel load. No other monitoring of BOP systems is anticipated. {ST-17)

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

12. 1. e ( 10) - Feedwater Heaters & Drain Sys terns - See abstract A46.l

13. l.e(11) - With the condensate polisher under normal operating conditions, Bechtel Corp. will make a complete inspection of all piping and hangers to verify adequate expansion and restraint capability. Test No.

A22 .1 will be performed to verify correct system operation.

Rev. *46, 06/94 423,.12-9

SSES-FSAR

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

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

16. 1.£(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. 1.g(l) - See abstracts for A3.1, P4.1, PS.1 and A7.l.

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

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

1} P69.1 - Liquid Radwaste Collection

2) P76.1 - Plant Leak Detection

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

22. l.h(2) - See abstract for P83.l 23.

23. 1.h{3) - Demonstrated during flush; not part of P.O.

No change.

24. 1.h (8) - Proper operation of valve sequencing for ECCS pump suction from the _ Condensate Storage TanJc 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. 1. i (l} - The containment design overpressure structural test is the Structural Integrity Test performed as a construction test.

26. 1.i(2) - See General Test Statement 1.i(2) - Revised abstract for Reactor Water Cleanup 1.i(2) - Added to abstract PS9.l 1.i(2) - See abstract for P59.l Rev. 46, 0-6/94 423 .12-10*

SSES-FSAR 27, 28,

29. l.i(3), (4), (5) - The tests covered by these portions of Reg. Guide 1.68 are Type Band 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 Band Type C test is conducted in accordance with the requirements of Subsection 6. 2. 6 of the FSAR. Acceptance criteria for the Type B and Type C tests is in accordance with the requirements of Chapter 16 of the FSAR.

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

31. 1.i(7) - See abstract for P83.1

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

33. 14i(9) - See revised abstract for P73.l

34. l.i{l0) - See revised abstract for P73.1

35. 1.i(13) - See revised abstract for ST-37.

36. l.i(lS) 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 containment penetrations is not part of Susquehanna design.

37. 1.i(l7) - See abstract for P34.1

38. i.i(l9) - See abstract for P59.l

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 containment 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.

Rev. 46, 06/94 423.12-11

SSES-FSAR

42. 1.j(lO) - See abstract for A99.6.

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

44. 1.j (12) - The off-gas pre-treatment system linear Wide Range Monitor detects failed fuel and is tested with other Process Radiation Monitors in P79.20.

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

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

47. 1. j (18) - Neutron response checks are part of the Power Test Program (STs 6, 10, 11, 12, & 18). Preoperational testing is addressed in abstracts P7B.1, P78.2, P78.3 and P78.4.

48. 1.j(l9) - Not a separate system tested in each ECCS System 1.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 PSB.1

50. 1. j (22) - Containment instrumentation is tested in the .

following pre-op tests:

Reactor Wide Range Pressure - P45.1 Feedwater Control Reactor Level - P45.1 Feedwater Control and P80.l Reactor Non-Nuclear Instrumentation Suppression Pool Level - P59.1 Containment and Suppression Radiation Detection - P79.l Area Radiation Monitoring and P79.2 Process Radiation Monitoring.

Humidity Monitors - Not in present Susquehanna SES

. design.

51. 1.j(24) - See abstract for Annunciator System

52. 1.j(25) - See abstract for Process Computer Rev. 46, 06/94 423.12-12

SSES-FSAR

53. 1.k(2) - See answer below for l.k(3)

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

55. 1.k(4) - HEPA filters and charcoal efficiency were tested by factory representatives on-site but not prior performing INAC pre-op tests. The pre-op test was reviewed and it was verified that pretesting of the HEPA filters and charcoal efficiency was not required.

56. 1.1(2) - See abstract for A72.1

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

58. 1.1(5) - See abstract for A43.l

59. 1. -1 (6) - See abstract for P34 .1 and Generai Test Statement

60. 1.1 (7) - Liquid radwa.ste effluent discharge to the environment is tested in Acceptance Test A69.2.l.

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 Leak Detection Test and verifies the operability of the leak detection.

62. l.m{l) - See revised abstract, part of TPl.9 for fuel pool

63. 1. 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 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 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 oL.the gates.

64. l.m(4) & 1.o(l) - For testing of the fuel handling system, see the abstract for P81.1. For testing of the reactor building crane see the abstract of P99.1.

65. l.m(S) - See. abstract for P81~1 66.

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

Rev. 46, 06/94 423 ~ 12-13 .

SSES-FSAR

67. l.n(l) 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. 1. n .(5) - Reactor Coolant and Secondary Sampling systems

- See abstract A76.2.

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

71. l.n{7) Fire Protection Systems are tested by Preoperational Tests Pl3.1 through Pl3.4.

72. Ln(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.l.

73. 1.n(9) - See abstract for A20.1

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

75. 1.n(13) - See abstract A99.2.

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

77. 1.n(lS) - Not applicable to Susquehanna SES design.

78. 1.n(18) - See abstract ABS.2.

79. l.m(4) & Lo(l) - For testing of the fuel handling system, see the abstract for P81.l. For testing of the rea*ctor building crane, see the abstract for P99 .1.

80. 1.o(2) For testing of protective devices and interlocks on the fuel handling system and reactor building crane, see the abstracts for P8.1 and P99.1 .

81. 1. o (3) - For testing of safety devices on the fuel*

handling system and reactor building crane, see the abstracts for P81.l and P99.l.

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 Rev. 46, 06/94 423.12-,.J.4

SSES-FSAR hydro of the reactor is performed per Plant Surveillance Tests. No 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 teated in P78.3. Further, ST-6, ST-10, ST-11, ST-12, and ST-18 demonstrate the overall operability of the nuclear 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.l and P78.2. The SCRAM function is tested in .P78.l 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 in sec.t:ion 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 positioning is verified in P55 .1 and control rod coupling is verified in ST-6.

However, it should be* noted that during ST-18 (TIP Uncertainty), which is performed at Test Condition (TC) .

3 and 6, the random noise, geometric, and total uncertainty of the TIE. 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.l and P7 9. 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 surveillance 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.l.

These systems are required by Plant Tech Specs to be Rev. 47, 06/94 423.12-15

SSES-FSAR 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, including 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 surveillance procedures as required by Administrative Procedure AD-000-75 {Station Inservice Inspection .Programs). No changes to the test descriptions need be made.

NQ.'.rn.: MSIV-LCS information maintained here for historical purposes. The MSIV-LCS has been deleted.

The function is now performed by the Isolated Condenser Treatment Method (Section 6.7).

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.

91. S.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 14.2-6-1, 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 lDOt rod line.

(2) ST-21 demonstrates the intersection of line b with the 100% rod line.

(3) ST-35 demonstrates lined at~ 50% and 100\ power.

Rev. 51, 02/97 423.12-16

SSES-FSAR (4) ST-30 demonstrates lined 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). The 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 mooified accordingly.

95. 5. m - The reactor recirculation system is initially tested, calibrated, and evaluated against design performance parameters in P64.l. During Startup Test Program, the system operating parameters are evaluated in ST-19 (Core Performance), ST-3O (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. Demonstratio~ of proper instrumentation response is performed during the system's preoperational test. In addition, 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 Technical Specifications. For all Rev. 51, 02/97 423.12-17

SSES-FSAR 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 qualitatively verified in the Generator Load 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 PS0.1 and P52.l.

Periodic surveillance testing is conducted to verify continued proper response times.

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

100. S.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 H&V systems for the DG Building, ESSW Pumphouse, Reactor Building, and Control Structure are teated in P28. 3, P28 .1, P34 .1 and P3 O. 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 Outside the Main Control Room) These tests are performed at TC6 {ST-8) and TCl (ST-28)

Emergency Service Water is tested in ST-36 {Cooling Water Systems) 101. 5.z - See response to Item 4.g.

102. 5.c.c - Gaseous radwaate system is tested in ST-37 at TC 1, 3, 5, and 6. Liquid Radwaste Collection System is demonstrated operable in P69.1. Solid Radwaste ..

Rev. 51, 02/97 . 423.12-18

SSES-FSAR Systems, Liquid Radwaste System, and Gaseous Radwaste Systems are demonstrated operable in A68.1, A69.2, and A72.1 respectively.

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 steam 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

( Recirculation Flow Control) at TC-1, 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 Generator Load Rejection) tests a turbine trip at TC 3 and tests a generator load rejection at TC6. No change to test description required.

Rev. 51, 02/97 423.12-19

SSES-FSAR QUESTION 423.13 Expand your test abstracts of Section 14.2.12 and those provided in answer to Questions 423.10 and 423.12 to describe in more detail the test objectives, prerequisites, test method, and acceptance criteria in regard to applicable parameters and functions (e.g., pressure, temperature, flow, valve operability, valve opening and closure times, controls, logics, and interlocks).

RESPONSE

Abstracts for the following preops have been revised: P2.l, 4.1, 5.1, 16.1, 17.l, 24.1, 30.1, 34.1, 45.1, 49.1, 51.1, 52.1, 53.1, 54.1, 55.1, 57.1, 58.1, 59.1, 59.2, 61.1, 73.1, 75.1, 76.1, 78.4, 83.1, 99.1, and 100.1. These abstracts are consistent with other licensing applications. The NRC will receive a draft copy of each test as it is developed and a copy of the approved test 60 days prior to its run date.

Rev. 46, 06/93 423.13-1

SSES-FSAR QUESTION 423.14 We note your position relative to Regulatory Guide 1.80 contained in Section 14.2.7 of the FSAR and disagree with your position. This guide is applicable since the instrument air system is used for a source of air for systems and components that provide a safety function. Modify your application to show that your test program will be consistent with the guide or show that you will conduct equivalent testing for the air system and supplied loads.

RESPONSE

The primary containment instrument gas system will be tested in accordance with the requirements of Regulatory Guide l. 80 Sections C.1 through C.6. The portions of the instrument air system which supply safety related equipment will also be tested in accordance with sections C.1 through C.6 of Regulatory Guide 1.80 {June, 1974).

The various components fed by the instrument air and instrument gas system will be tested to ensure proper operation on loss of air/gas. This testing will be done as part of the various systems preoperation testing in which the components are located.

The action and flow of decay air is not an essential criteria of operation in relation to the affected components. The components are to fail with loss of air/gas to a safe position.

Whether decaying pressure will hold some or all of the valves in normal operating positions is not of critical importance.

In addition to the above testing, the systems/components whi.ch have separate accumulators (MSIV's, safety relief valves} will be tested with a loss of air/gas to ensure that the accumulators function in accordance with design.

Loss of air testing as described above will be done in the various system preoperational tests. Therefore testing described in Regulatory Guide 1.80 Sections C.7 through C.10 will not be done in the instrument air system or primary containment instrument gas system tests.

See revised Section 3.13.

Rev. 46, 06/93 423.14-1

SSES-FSAR QUESTION 423,15 We could not conclude from our review of the preoperational test phase and the test abstracts provided in Table 14.2 that comprehensive testing is scheduled for several of the described tests. Therefore, clarify or expand the description of the preoperational test phase to address the following:

(1) Modify the individual A.C. and D.C. distribution system test descriptions or provide an integrated test description to verify proper load group assignments (reference Regulatory Guide 1.41).

(2) Class lE 125 Volt D.C. System Preoperational Tests -

State your plans for demonstrating the following: (a) that emergency loads are in accordance with battery sizing assumptions; and (b) that each emergency load can operate at the minimum voltage level at which it can be postulated to operate.

(3) State how operability of emergency loads using offsite power will be demonstrated during A.C. and D.C. system tests.

(4) Identify testing that will be accomplished to verify drywell floor bypass leakage and provide quantitative acceptance criteria.

(5) State your plans for assuring that the effects of interfacing hardware (e.g., snubbers, pulse dampers) located between measured variables and the input to the sensors for the Reactor Protection System do not compromise the channel response time requirements.

(6) Control Room HVAC System Preoperational Test - Expand the test description to include a demonstration that outleakage from the control room is in accordance with design assumptions when the system is on the emergency outside air supply.

RESPONSE

(1) See abstracts for P2.l, A3.l, P4.1 and PS.1.

(2) See revised abstract for P2.1.

(3) See abstract for Pl00.1.

(4) Testing will be done per P69.1.

Rev. 46, 06/93 423.15-1

SSES-FSAR (5) In our tests, we do not address the effects of interfacing hardware located between measured variables and the input to the sensors on the channel response time for the Reactor Protection System.

(6) See revised abstract for P30.1.

Rev. 46, 06/93 423.15-2

SSES-FSAR QUESTION 423.16 Describe your tests to demonstrate that the core spray flow distribution header provides adequate cooling flow to each fuel assembly.

RESPONSE

Preoperational test PSl. lA, Core Spray System Pattern Test, describes the demonstration of adequate core spray cooling flow to each fuel assembly.

Rev. 46, 06/93 423.16-1

SSES-FSAR QUESTION 423.17 Provide a description of the electrical lineup for Unit No. 2 during preoperational tests that will be conducted to satisfy regulatory positions in Regulatory Guide 1.41 for Unit No. 1.

Provide a description of the lineup for both plants during similar preoperational testing on Unit No. 2 subsequent to initial criticality of Unit No. 1. The descriptions should address both normal and emergency A.C. and D.C. power distribution systems. Provide assurance that crossties will not exist which could cause loss of emergency bus power to one unit due to testing of the other unit.

RESPONSE

Unit 1 and Unit 2 13.8 KV systems will be jointly tested before Unit 1 initial criticality (Acceptance Test A3.1).

Unit 1 and Unit 2 4.16 KV systems will be jointly tested before Unit 1 initial criticality (Preoperational Test P4.l).

Lineups for these tests will be for Unit 1 testing. There will be no testing of the Unit 2 systems after initial criticality of Unit 1.

This integrated testing and system design will satisfy the requirements of Regulatory Guide 1.41 and will assure that no crossties exist which might cause loss of emergency bus power to one unit due to testing the other unit.

The design of the 480 volt systems allows for the isolation of each unit from the other. This allows testing on one unit without affecting the other.

The Unit 1 and Unit 2 DC systems are entirely separated. There are no crossties.

Rev. 46, 06/93 423.17-1

SSES-FSAR QUESTION 423.18 Provide a commitment to include in your test program any design features to prevent or mitigate anticipated transients without scram (ATWS) that may be incorporated in your plant design.

RESPONSE

FSAR Table 15.0-la lists ATWS analysis as, "still under discussion." Testing will be done to the extent practicable to ensure compliance to any ATWS design when that design is finalized for Susquehanna SES.

Rev. 46, 06/93 423.18-1

SSES-FSAR QUESTION 423.19 Provide preoperational test descriptions (or modify existing descriptions) to assure that each engineered safety feature pump operates in accordance with the manufacturer's head-flow curve. Include in the description the bases for the acceptance criteria. (The bases provided should consider both flow requirements for ESF functions and pump NPSH requirements.)

RESPONSE

Testing to verify that ESF pumps operate within their design pump-head curves and with adequate NPSH will be done. This testing is committed to in the General Test Statement as part of the answer to Q423.12.

Steam conditions from the two station auxiliary boilers will permit turbine testing of RCIC and HPCI systems. However it will prohibit any pump testing. This will be verified under the power test program using nuclear steam.

Rev. 46, 06/93 423.19-1

SSES-FSAR QUESTION 423,20 Our review of the power test abstracts provided in your FSAR disclosed that they are not sufficiently descriptive to conclude that comprehensive testing is planned or that satisfactory test acceptance criteria have been established.

The individual test abstracts should be modified as indicated below.

(1) Modify your acceptance criteria for test PT-1, Chemical and Radiochemical, to provide a level 2 acceptance criteria of design basis for your condensate demineralizers and RWCU.

(2) Your acceptance criteria for test PT-2, Radiation Measurements, is not consistent with the design objectives of ALARA. Therefore, revise your acceptance criteria to be consistent with your plant design objectives.

(3) Modify your test abstract for Full Core Shutdown Margin to specify the value R. In addition, specify a quantitative value for your level 2 criterion and that value be considered a level 1 criterion.

(4) Revise all power test acceptance criteria where you use the term "specified value" to provide a specific numerical value for those acceptance criteria.

( s) Test PT-4, Ful 1 Core Shutdown Margin - Provide the temperature of the core for the shutdown margin test.

(6) Test PT-5, Control Rod Drive system - The level 1 acceptance criteria for control rod withdrawal speeds are inconsistent and nonconservative in respect to the times assumed in your accident analysis. Resolve this inconsistency. Also, this test abstract should be expanded to provide assurance that dash-pot performance will be in accordance with design requirements and acceptance criteria should be provided for control rod scram times.

(7) Test PT-9, Water Level Measurement - Revise Figure

14. 2-5 to include water level measurement tests at Test Conditions 1, 3, 4, & 5 in addition to those already specified.

( 8) Test PT-10, IRM Performance - Revise the acceptance criteria to include a check for the IRM scram trip point.

Rev. 46, 06/93 423.20-1

SSES-FSAR

{9) Test PT-14, RCIC System State your plans to demonstrate the capability of the system to start from the "cold" condition. Also, clarify or justify the Level 1 acceptance criteria provided in Paragraph 4.

Based on operating experience to date, the apparent reliability of the reactor core isolation cooling system (RCIC) in BWR plants has been poor. Because it appears that many of the causes for the failure should have been detected and corrected during initial testing of the RCIC system, this system should be given a very thorough checkout during the initial testing program.

Your current test proposal does not appear adequate to establish confidence in the reliability of the system for your facility. Your application should be modified to show that several consecutive successful cold starts of the RCIC system will be demonstrated during your power ascension phase.

(10) Test PT-15, HPCI System - Based on operating experience to date, the apparent reliability of the high pressure core injection system (HPCI) in BWR plants has been poor. Because it appears that many of the causes for the failure should have b~en detected and corrected during initial testing of the HPCI system, this system should be given a very thorough checkout during the initial testing program. Your current test proposal does not appear adequate to establish confidence in the reliability of the system for your facility. Your application should be modified to show that several consecutive successful cold starts of the HPCI system will be demonstrated during your power ascension phase.

(11) Test PT-16, Selected Process Temperatures - Modify your Level 1 acceptance criteria to include the pump in an idle loop; and your Level 2 acceptance criteria to relate to loop temperature.

(12) Test PT-18, Core Power Distribution - Revise your test method to specify how many sets of TIP data will be taken to determine the overall TIP uncertainty.

(13) Test PT-22, Pressure Regulator - Specify the mode of control (auto or manual) of each of the other principal control systems at each test condition.

(14) Test PT-23, Feedwater System - Modify your test objectives to include the loss of a feedwater heater.

Specify the mode of control (auto or manual) of each of the other principal control systems at each test condition for the feedwater control setpoint changes.

Rev. 46, 06/93 423.20-2

SSES-FSAR Also, the test description should be modified for the feedwater heater trip to specifically identify: (a) the type of trip to be initiated; (b) the feedwater heater(s} involved; and (c) a discussion of how the planned trip relates to the worst case limiting event for your design that could result from a single equipment failure or operator error. Modify your test method to include the loss of all feedwater flow.

Provide justification for performing the feedwater pump trip in Master Manual Flow Control Mode rather than Automatic Flow Control Mode for feed water pump trip and for feedwater heater loss.

(15} Test PT-25, Main Steam Isolation Valves - Provide clear acceptance criteria for relief valve and RCIC performance during this transient.

(16) Test PT-26, Relief Valves - Describe your test method and acceptance criteria for bypass

valve flow calibration and capacity. Modify your acceptance criteria to include opening times (to full capacity) of relief valves.

(17) Test PT-27, Turbine Trip and Generator Load Rejection -

Modify your test abstract to: (a) identify the method of tripping the main generator breaker; (b) identify the conditions for each trip planned; (c) identify the variables or parameters to be monitored for each trip; (d} provide assurance that test results will be compared with predicted results for the actual tests to be run (for each trip); (e) provide quantitative acceptance criteria and their bases for the required degree of convergence of actual test results with predicted results for the monitored variables and parameters for each trip; and (f) provide acceptance criteria for grid stability, voltage and frequency following generator load rejection trips.

( 18) Test PT-28, Shutdown From Outside the Main Control Room

- State whether the plant's electrical system will be aligned for normal full power operation and provide acceptance criteria for the performance of plant equipment and the variables or parameters to* be monitored during the test.

(19) Test PT-29, Recirculation Flow Control - Specify the mode of control (auto or manual} of each of the other principal control systems at each test condition.

(20) Test PT-30, Recirculation System - Modify the test Rev. 46, 06/93 423.20-3

SSES-FSAR abstract to define the types of trips to be conducted at each test condition and the manner by which the pumps will be tripped. Also, modify the test description and provide quantitative acceptance criteria for flow coastdown and trip of both the recirculation pumps. Also, provide stability criteria for plant performance following the trips.

(21) Test PT-31, Loss of Turbine-Generator and Offsite Power

- Modify the test abstract to: (a} describe the initial plant conditions for the test, including the lineup of the plant's electrical system; (b) describe the type of trip to be conducted; (c) identify the variables, parameters, and plant equipment to be monitored; (d) provide assurance that test results will be compared with predicted results for the actual test case; (e) provide quantitative acceptance criteria and their bases for the required degree of convergence of actual test results with predicted results for the monitored variables and parameters; and (f) provide functional acceptance criteria for plant equipment that should function during or following the test. Also, correct the Level 1 acceptance criteria to be consistent with your facility design.

(22) Test PT-32, Containment Atmosphere Circulation System -

Modify your acceptance criteria to include Level 1 criteria based on concrete temperatures.

(23) Test PT-35, Recirculation System Flow Calibration -

Modify the test method to add calibrations at test conditions 2 and 5.

RESPONSE

1) The design basis for the condensate demineralizers and RWCU are contained in the Water Quality Specifications which is part of the Level 1 Acceptance Criteria. See revised abstract for ST-1.

2) Acceptance Criteria based on ALARA objectives are included. See revised abstract for ST-2.

3) The value of R, an exposure dependent correction factor, and the predicted critical is contained in the Cycle Management Report, which is not yet available.

The test abstract will be updated when the report becomes available.

Rev. 46, 06/93 423.20 ... 4

SSES-FSAR Level 1 Criterion normally relate to the value of a process variable assigned in the design of the plant.

Since the predicted critical is an expected value relating to the performance of the plant and not a design variable, it should remain a Level 2 Criterion.

4) The term "specified value" has been replaced by specific numerical values for those tests for which such values are available.

5) The formula used in ST-4 to determine the shutdown margin includes a moderator Temperature Coefficient to relate the shutdown margin at actual moderator temperature to the shutdown margin at a moderator temperature of 68°F upon which the Acceptance Criterion is based. See revised abstract for ST-4.

6) The withdraw speeds given in ST-5 are in agreement with figure 15-0-2 curve c. The withdraw speeds are considerably slower than the SCRAM rod speeds. SCRAM speed acceptance criteria are in accordance with the Technical Specifications. Control rod buffer performance is tested in PSS.1.

7) Figure 14-2-5 has been revised to include testing at Test Condition 1, 3, 4 & 5.

8) The IRM SCRAM Trip point is verified and tested during the preop test program. ST-10 does not plan to perform a functional test of the IRM trip point. The IRM's are further tested during normal plant Surveillance Testing.

9) &

10) See revised abstracts for HPCI and RCIC testing.

11) The Level 1 Acceptance Criteria have been modified to include a pump in an idle loop. The Level 2 Acceptance Criterion has been transferred to ST-7 and now does relate to loop temperature.

12) The number of sete of data to be taken is described as a note to the acceptance criteria for ST-18.

13) ST-22 Pressure Regulator - See Figure 14.2-5 Sheet 1 for a description of the control mode of the recirculation system for this test. The feedwater control system will be in the mode suitable to operating plant conditions (typically 3-element Master Auto).

Rev. 46, 06/93 423.20-5

SSES-FSAR

14) ST-23-Feedwater System-See revised abstract for ST-23.

Testing for loss of feedwater heating in the manual flow control mode is a more severe transient than testing in the Auto mode as described in Sect ion 15.1.1. Testing for loss of feedwater flow will cause a Recirculation System Runback. The ef feet is the same regardless of Master Manual or Master Automatic Flow Control. The loss of power to the extraction steam bleeder trip valves for one feedwater heater train results from failure of the electrical feed to the valves and is the event which is tested in ST-23.

We have reviewed possibilities for loss of all feedwater flow including pump or valve. failures, feedwater controller failures, operator errors, and reactor system variables. Based on our evaluation, total loss of feedwater flow testing will not be included in ST-23.

Of the above mentioned failures, no single failure will cause loss of all feedwater. Pump or valve failures may reduce system capacity but will not result in loss of all feedwater. Feedwater controller failure will energize an annunciator circuit if the control signal to the RFP is lost. The alarm furnishes contacts which are utilized by the F.P. turbine speed control circuit to maintain turbine speed at the level existing at the time of signal loss. The feedwater control will transfer from 3 element control (level, steam flow, feedwat~r flow) to single element control on level.

Reactor variables such as water level will cause scram at (L-4) low water level, and alarm on (L-7) high water level, but will not cause loss of all feedwater before scram or operator initiated shutdown.

Loss of feedwater is considered in Section 15.2 which addresses increase in reactor pressure. Increase in reactor pressure start-up tests of similar intent but greater impact are performed as part of the start-up test program (e.g. ST-25 and ST-27 Turbine Trip and Generator Load Rejection) MSIV closure. The sequence of equipment response and operator action for ST-25 and ST-27 are identical to the loss of feedwater test.

Reactor pressure and level instrumentation functional tests are being added to the pre-operation test program described in Section 14. 2 to verify proper operation of this instrumentation.

15) See revised abstract for ST-25 MSIV's.

Rev. 46, 06/93 423.20-6

SSES-FSAR

16) Bypass valve flow calibration, capacity and opening times are not tested during the Startup Test Program.

Testing is done which verifies proper operation and reseating of each relief valve and verifies that no major blockages in the relief valve discharge piping exist. Opening times and capacity for the relief valves are tested at the factory and are not repeated.

See revised abstract for ST-26.

17) See revised abstract for ST-27 for items (a) thru (e).

The Initial Test Program is designed to demonstrate the performance of structures, systems, components, and design features that will be used during normal operations of the facility and also demonstrate the performance of standby systems and features that must function to maintain the plant in a safe condition in the event of malfunctions or accidents. The Susquehanna SES Initial Test Program does not include Acceptance Criteria for non-Susquehanna SES designed systems, such as the electrical grid system, whose performance is not under the control of the plant.

18) Plant's electrical system alignment is included in revised abstract for ST-28. The main objective of this test is to demonstrate that the reactor can be shutdown from outside the main control room. The performance of plant systems in response to transients and abnormal conditions is demonstrated in individual system's tests.

19) ST Recirc. Flow Control - The mode of control of the recirc. system is specified in Figure 14.2-5 Sht. 2. The feedwater control system will be in the mode of control which is specified by plant operating procedures for the various power levels.

20) Quantitative acceptance criteria for the flow coastdown after a two pump RPT trip wil 1 be included in the Transient Safety Analysis Design Report which is not yet available. The acceptance criteria will be revised when the information becomes available. Revised abstract for ST-30 describes the pump trips in more detail and provides acceptance criteria.

21) See revised abstract for ST-31 for items a, b, c and f.

This test is performed at 30% to demonstrate the proper performance of the electrical distribution system and safety systems during a loss of the turbine-generator and offsite power. Predictions are made for the worst case transients rather than low power transients. The Rev. 46, 06/93 423.20-7

SSES-FSAR proper performance of the plant to a turbine trip at 100\ power is demonstrated in ST-2?.

22) See revised abstract for ST-32. Acceptance criteria is based upon containment air temperatures not concrete temperatures.

23) Recirculation flow calibrations are done at Test Conditions 3 & 6 where flow is sufficient to provide meaningful flow data. Additional data at TC 2&5 would not provide any additional meaningful data.

Rev. 46, 06/93 423.20-8

SSES-FSAR QUESTION 423,21 You state in Subsection 14.2.4.6 that the completion of Phase II on safety-related systems is a prerequisite for commencement of the Power Test Program. Describe any preoperational tests shown in Tables 14.2-1 and 14.2-2 that you consider need not be completed prior to the commencement of the Power Test Program.

RESPONSE

The tests listed in Table 14.2-1 are a pre-requisite to commencement of the start-up test program. The test results and exceptions to the tests will be evaluated, reviewed and approved per Subsection 14.2.5. The tests listed in Table 14.2-2 may be conducted on non-safety-related equipment. Table 14.2-2 is not a pre-requisite to commencement of the start-up test program.

Rev. 46, 06/93 423.21-1

SSES-FSAR QUESTION 423.22 Describe any preoperational and startup tests that you will conduct on Unit No. 1 that you may not conduct on Unit No. 2.

RESPONSE

See the response to Question 423.34.

Rev. 46, 06/93 423.22-1

SSES-FSAR QUESTION 423.23 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.l. 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.

Rev. 46, 06/93 423.23-1

SSES-FSAR QUESTION 423.24 Your answer to parts (a) and (b) of Question 423.l regarding the qualification requirements for persons performing the functions of preoperational test directors and st~rtup test directors are not satisfactory.* We consider that the minimum qualifications for persons that direct or supervise the conduct of preoperational tests include a bachelor's degree in engineering or the physical sciences or the equivalent and one year of applicable power plant experience. Included in the one year of experience should be at least 3 months of indoctrination/training in nuclear power plant systems and component operation in a nuclear power plant that is substantially similar in design to the type at which the individual will perform the function. We consider that the minimum qualifications for persons that direct or supervise the conduct of individual startup tests should include a bachelor's degree in engineering or the physical sciences or the equivalent and two years of applicable power plant experience, at least one year of which should be applicable nuclear power plant experience. Revise your FSAR to indicate conformance to the staff position.

RESPONSE

see revised response to Question 423.1.

Rev. 46, 06/93 423.24-1

SSES-FSAR QUESTION 423.25 The response to item 423.8 stated that FSAR Subsection 14.2.7 would be revised to show a 60 day period for NRC review of test procedures. The revision was made in 14. 2. 3, not 14. 2. 7.

Correct the item 423.8 response.

RESPONSE

See revised response to Question 423.8.

Rev. 46, 06/93 423.25-1

SSES-FSAR QUESTION 423.26 The response to item 423.10 is incomplete. Provide abstracts for the following tests: A84 .1; ASS. 2; A87 .1; i\99. 2; and A99.6.

RESPONSE

See Subsection 14. 2 .12. 3 for test abstracts A84 .1, ASS. 2, A99. 2 and A99.6. Test A87.1 has been incorporated into test A98.l.

Rev. 46, 06/93 423.26-1

SSES-FSAR QUESTION 423.27 Your response to item 423.11 states that current readings of containment recirculation fans will be higher during ILRT than at accident conditions. Provide technical justification for this statement. Address such issues as air density, temperature, humidity, fan speed and blade angle.

RESPONSE

See revised response to Question 423.11.

Rev. 46, 06/93 423.27-1

SSES-FSAR QUESTION 423.28 The response to item 423.14 indicates that testing described in Regulatory Guide 1.80 sections C.? through C.10 will not be done since the testing will have already been done during "various system preoperational tests". Either provide test descriptions that show testing equivalent to that specified in regulatory positions C.8, C.9, and C.10 will be performed, or modify your preoperational test program to include an integrated loss of air test and provide an abstract of that test.

RESPONSE

See revised response to Question 423.12.

Table 423.28-1 lists all operator valves/HVAC dampers which a re tested for loss of air Preoperational tests within which the loss of air testing is accomplished is also provided in Table 423.28-1.

Further testing is performed for the ADS/SRV valves as fol lows:

1. Verify minimum capacity of accumulator in acceptance criteria.

2. Verify 'ADS/SRV' s are operated from their respective accumulator/supply with other supplies depressurized.

3. Record pressure at which an open valve begins to close for safety/relief valves and verify valve fails to closed on loss of air.

4. Verify an open ADS valve is maintained open at accumulator pressure of 75 + O - 2 PSIG and fails closed on loss of air.

Rev. 46, 06/93 423.28-1

SSES-FSAR TABLE 423.28*1 SYSTEM VALVE NUMBER PREOPERATIONAl NO. INSTRUMl:NT AIR OR PRIMARY CONTAINMSfT INSTRUMENT OAS RHR 1-E11-F050A,B P49.1 rnstrument Gas 1-E11-F122A,B ,nstrument Gas 1-E11-F051 A,B tnstrument Air 1-E11-F052A,B 1-E11-F053A,B 1-E11-F111A,B 1-E11-F129A,B 1-E11-F132A,B 1-E11-F136, F137, F140 RCIC HV-E51-1F008 P50.1 Instrument Gas HV-E51-1F025, 1F026 Instrument Air HV-E51-1F004, 1F005 HV-E51-1 F054 Core Spray HV-E21-1 F006A,B P51.1 Instrument Gas HV-E21 -1 F037A,B Instrument Gas HPCI HV-E41-1F028, 1F029 P52.1 Instrument Air HV-E41-1F025, 1f026 LV-E41-1 F054, 1F100 1F100 Gas Others Instrument Air CAD C 12-F002A,B P55.1 Instrument Air XV-1F010, 1F011 RECIRC HV-831-1 F019, 1 F020 Both* F019-lnstrument Gas F1 ()().Instrument Air Fire Protection XV-12248,49 P13 Instrument Air XV-02248 XV-02215 RBCCW HV-11315 P14 Instrument Air Rev. 46, 06/93 Page 1 of 3

SSES-FSAR TABLE 423.28 .. 1 (Continued)

SYSTEM VALVE NUMBER PREOPERATIONAl NO. INSTRUMENT AIR OR PRIMARY CONTAINMENT INSTRUMENT OAS RB HVAC HD17534A,B,C,O,E,F,H Alf* P34.1 Instrument Air HD17502A,B; HD17514A,B AU*

HD1 7564A,B; H017524A,B AH HD17576A,8; H017586A,B All*

H017508A,B Both*

HD17651 RWCU HV-14506A,B; 14507A,B P61.1 Instrument Air HV-14508A,B; 14510A,B HV-14511A,B; 14512A,B HV-14513A,B; 14514A,B HV-14566A,B; 14522 HV-14523, 14528, 14516 HV-14518, 14519, 14520 HV-14521, G33-1F033 Liquid Radwaste HV-1 61 OBA 1 , HV-1 6116A 1 P69.1 Instrument Air HV-16108A2, HV-16116A2 Both*

Containment HV-15721, 23, 24, 22, 25 An* P73.1 Instrument Air Recirculation HV-15704, 05, 14 All*

HV-15703, 13 HV-15711 R.B. HVAC POD17501 A,B; H017511 A,B P34.1 Instrument Air HD17521A,B; HD17513A,B HD17518A,B; H017516 H017523A,B; HD17528A,B PDD175 78A,B; HD17526 H017566A,B; H017588A,B "HD17538A,8 Rev. 46, 06/93 Page 2 of 3

SSES-FSAR TABLE 423.28-1 (Continued)

SYSTEM VALVE NUMBER PREOP£RATIONAL NO. INSTRUMENT AIR OR PRIMARY CONTAINMENT INSTRUMENT OAS RB Chilled Water TV-18726A 1,A2,B1 ,B2 P34.2 Instrument Air TV-18741 A,B,C,D TV-18743A,B TV-18751 A,B,C,D TV-18753A,B TV-18764A,B FV-18771A,B,C,O HV-18781 A 1,A2,B1 ,B2 AU* Instrument Gas HV-18782A 1,A2,B1 ,82 AU*

HV-18791 A 1,A2,B1 ,82 AU*

HV-18792A 1,A2,B1 ,82 All*

Control Structure HOM-07802A,B Both* P30.1 Instrument Air HVAC HOM-07833A,B; HOM-07824A2,B2; HOM-07824A 1,B1 L.lnM.n'7A?A.AA. RA* MnM.n'7AR1 A D I -""191 °" I ...,&,,.,---r-~--,,,,....,..--,,,, * ,...,., ... , * ..., P ...,'-' t -'""',..,.

HOM-07872A,B; HOM-07873A,B All*

TV-07813A,B TV-08602A,B P30.2 Feedweter FV-10604A,B,C; HV-10640; LV-10641 P45.1 Instrument Air HV-14107A,B: HV-10650 HV-10606A,B,C TV-10663A 1,A2,B1 ,B2,C1 ,C2 LV-10664A,B,C Rev. 46, 06/93 Page 3 of 3

SSES-FSAR QUESTION 423.29 The response to item 423.15 is not complete. It is the staff's position that you (1) provide quantitative acceptance criteria for drywell floor bypass leakage and (2) modify your Reactor Protection System test to account for the delay time of interfacing hardware (e.g., sensing lines) on channel response time.

RESPONSE

{l) The quantitative acceptance criteria for drywell floor bypass leakage is 176.4 scfrn at a differential pressure of 4.3 psid.

(2) Per the NRC Standard Technical Specification (NUREG 0123), the Reactor Protection Response Time shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energized of the scram pilot valve solenoids. Therefore, no modification for the RPS is required. No account is made in the RPS test for delay times in sensing lines for pressure as delay time contribution to channel response in negligible.

Rev. 46, 06/93 423.29-1

SSES-FSAR QUESTION 423.30 Modify PSl.l to make it consistent with the response to item 423.16.

RESPONSE

See revised response to Question 423.16, Preoperational Test PSl.l and new Preoperational Test P51.1A.

Rev. 46, 06/93 423.30-1

SSES-FSAR QUESTION 423.31 The response to item 423 .17 states that Unit 1 and Unit 2 preoperational testing on the 4.16 kV system (P4.l) will be accomplished jointly in one month commencing 14 months prior to fuel load on Unit 1 (Figure 14.2 - 4a) and 11 months prior to fuel load on Unit 2 (Figure 14.2 - 4b). Section 14.2.11 states that because "the initial fuel loading of Unit 2 is scheduled to occur 18 months after Unit 1, the test programs will not overlap." Modify Chapter 14 and the response to item 423 .17 as necessary to correct this discrepancy. In addition it will be necessary for you to provide the information requested in item 423.17 (i.e., electrical lineups) in enough detail for us to determine the following:

(1) That during the Regulatory Guide 1.41 testing on each unit, there will be no crossties from the other unit's electrical system that could compromise the validity of the test results.

(2) That if Unit 1 is licensed at the time the Unit 2 test is performed, there will be no crossties that could cause a loss of power to Unit l emergency bus.

RESPONSE

As stated in the response to Question 423 .17, the preoperational test encompassing the 4.16 kV systems for both units (P4.l) will be completed prior to fuel load of Unit 1.

4.16 kV preoperational testing to be performed prior to Unit 1 fuel load will include ES Transformers 101 and 201 and ES busses lC, 1D, lE, lF, 2C, 2D, 2E and 2F including all feeder breakers. Since the 4.16 kV system is common to both units up to the feeder breakers for the unit ES busses, discussion of crossties affecting the validity of testing is not pertinent.

The isolation points between Unit 1 and common and untested Unit 2 equipment are the feeder breakers for the Unit 2 ES busses. During testing of Unit 2 equipment with Unit 1 in operation, these feeder breakers prot~ct against Unit 1 EA bus power failures originating in Unit 2.

13.8 kV preoperational testing to be performed prior to Unit l fuel load will include Startup Transformers 10 and 20, Startup busses 10 and 20 and Auxiliary busses lA, 1B, 2A and 2B including all feeder breakers except those from the Unit 2 Auxiliary Transformer. Since 13.8 kV system is common to both units, discussion of crossties affecting the validity of testing is not pertinent. The isolation points between Unit 1 and common and untested Unit 2 equipment are the feeder Rev. 46, 06/93 423.31-1

SSES-FSAR breakers at auxiliary busses 2A and 2B and Startup bus 20.

During testing of Unit 2 equipment with Unit l in operation, these feeder breakers protect against Unit l power failures originating in Unit 2.

Figures 14.2-4a and 14.2-4b show typical preoperational test schedules and are not intended to be updated.

Rev. 46, 06/93 423.31-2

SSES-FSAR QUESTION 423.32 The response to item 423.19 states that: "Testing to verify that ESF pumps operate within their design pump head curves and with adequate NPSH will be done. This testing is committed to in the General Test Statement as part of the answer-to Question 423.12." The general test statement says this will be done, but only "where possible". Modify the response to indicate that all ESP pumps will be completely tested.

RESPONSE

The response to Question 423.12 has been revised to specify the scope of pump testing to be performed during the preoperational test program. The testing to be performed on the HPCI pump is described in the ST-15 abstract. Insufficient auxiliary steam capacity precludes preoperational HPCI testing. This also applies to the RCIC system.

Rev. 46, 06/93 423.32-l

SSES-FSAR QUESTION 423.33 The response to item 423.20 indicates that certain changes will be made to the initial test program. Some of these changes have not yet been reflected in Chapter 14.

1. Modify Figure 14.2-5 as stated in sub-item 7.

2. Revise the abstracts for HPCI and RCIC tests to include the demonstration of several successful cold starts as stated in sub-items 9 and 10.

3. Modify the PT-26 abstract to state that a review of factory test results (flow and opening times) is conducted as part of the overall test review program as described in the response to sub-item 16.

RESPONSE

See revised Subsection 14.2.12 and Question 423.20.

Rev. 46, 06/93 423.33-1

SSES-FSAR QUESTION 423.34 Modify Figure 14.2-4b to make it consistent with the response to item 423.22. (Add or correct Tests P70.l, P30.2, P88.l, and P28 .1.

RESPONSE

As stated in the response to Question 423.31, Figure 14.2-4b is a typical preoperational test sequence. The following information reflects the testing to be performed:

P30.2 and P28.1 will be performed on Unit 1 only.

P88.l will be performed on Unit 1 and Unit 2

P70.1 will consist only of a negative leak rate test on Unit 2.

Rev. 46, 06/93 423.34-1

SSES-FSAR QUESTION 423.35 Expand or explain the following terms:

"Interlocks the RFPT alternate II (P45.1)

"high-high temperature . . . " (A30.3)

RESPONSE

(1) See revised acceptance criteria for P45.1 (2) For charcoal filters, high temperature actuates an alarm and high-high temperature actuates an alarm and the fire protection system.

Rev. 46, 06/93 423.35-1

SSES-FSAR QUESTION 423. 36 Include testing of the communications system in the preoperational tests or provide assurance that the test procedure and results will be reviewed in a manner similar to the preoperational tests.

RESPONSE

This acceptance test (A99.2) and its results will be reviewed in a manner similar to the preoperational tests.

Rev. 46, 06/93 423.36-1

SSES-FSAR QUESTION 423.37 The exception to Regulatory Guide 1.108, "Periodic Testing of Diesel Generator Units as Onsite Electric Power Systems at Nuclear Power Plants" (Revision 1), concerning the number of necessary consecutive val id tests per diesel is not acceptable.

It is the staff's position that you perform the~ starts in accordance with Regulatory Position 2.a(9). Modify Subsection 14.2.7 to state that your test will be conducted in accordance with this position or provide a description of tests that you will perform to demonstrate the required reliability.

RESPONSE

See revised Subsection 14.2.7.

Rev . 4 6 , O6 / 9 3 423.37-1

SSES-FSAR QUESTION 423.38 Include Regulatory Guide Ll40, "Design, Testing and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Absorption Units of Light~Water-Cooled Nuclear Power Plants" (Revision 1), in Subsection 14. 2. 7. Provide justification for any exceptions to Regulatory Positions c.s and C.6.

RESPONSE

See revised Subsection 14.2.7.

Rev. 46, 06/93 423.38-1

SSES-FSAR QUESTION 423.39 Revise Table of Contents listing of Tables and Figures to reflect the current status of FSAR Section 14.

RESPONSE

See revised Table of Contents.

Rev. 46, 06/93 423.39-1

SSES-FSAR QUESTION 423.40 Our review of recent licensee event reports disclosed that a significant number of reported events concerned the operabi 1 i ty of hydraulic and mechanical snubbers. Provide a description of the inspections or tests that will be performed following system operation to assure that the snubbers are operable.

These inspections or tests should be performed preoperational ly if system operation can be accomplished prior to generation of nuclear heat.

RESPONSE

Existing QA records on the construction installation and inspection of safety-related snubbers will be assembled into a package for review by the Superintendent of Plant. This package will provide assurance that the preoperational condition of the snubbers is acceptable and that they are installed in accordance with design.

After system preoperational testing and prior to fuel load, snubbers will be visually examined* and manually tested for freedom of movement over the range of stroke in both compression and tension. This meets the requirement of IE Bulletin 81-01 Rev. 1. No hydraulic snubbers are utilized in safety applications at Susquehanna SES.

Rev. 46, 06/93 423.40-1

SSES-FSAR QUESTION 423.41 Revise acceptance test A39 .1 (Condensate demineral izer Abstracts) to correct the following inconsistencies:

1) State whether the system will process water at 120%

above rated capacity (Test Method) or at 120\ of rated flow (Acceptance Criteria).

2) Ensure that monitored conditions are at least held at design specifications (Test Method}.

RESPONSE

See revised acceptance test A39.1.

Rev. 46, 06/93 423.41 ... 1

SSES-FSAR QUESTION 423.42 Clarify the first acceptance criterion in P45 .1 (Feedwater System Preoperational Test).

RESPONSE

See revised preoperational test P45.1.

Rev. 46, 06/93 423.42-1

SSES-FSAR QUESTION 423,43 Modify the test method of P99.l (Reactor Building Crane Preoperational Test) so that operation is completely checked in both directions vice "either direction" as stated.

RESPONSE

See revised preoperational test P99.1.

Rev. 46, 06/93 423.43-1

SSES-FSAR QUESTION 423.44 Modify the Figure 14. 2-3 references to refer to the proper figures.

RESPONSE

See revised Figure 14.2-3 references.

Rev. 46, 06/93 423.44-1

SSES-FSAR QUESTION 423.45 (1) The response to item 423.12 is not completely acceptable. Several acceptance test abstracts (A3.l, A13.l-A13.4, A15.1, A41.1, A45.1, A45.2, and A681,) are labeled as preoperational tests. Correct these inconsistencies (2) The response to several sub-items (i.e., l.i.2, 5.+,

S.u) does not address valve closure times. Modify the response to address them or provide technical justification for the deletions.

(3) If the factory testing of a component substitutes for inplant testing, then: 1) the method of testing, 2) the results of the testing, and 3) how these results are extrapolated to actual plant conditions should be reviewed and retained. Modify your response to provide commitment.

RESPONSE

(1) Tests 13.1 through 13.4, 45.1 and 45.2 are preoperational tests. Tests 3.1, 15.1, 41.1 and 68.1 are acceptance tests. See also the revised response to Question 423.12.

(2) See revised response to Question 423.12 and preoperational test P59.1.

(3) See revised response to item 5.t of Question 423.12.

Rev. 46, 06/93 423.45-1

SSES-FSAR QUESTION 423.46 Revise the test method of A93.1 (Turbine Lube Oil System) so that it indicates the actual test method.

RESPONSE

See revised acceptance test A93.1.

Rev. 46, 06/93 423.46-1

SSES-FSAR QUESTION 423.47 Explain the status of ST-84 (RPV Internals Vibration). It has been deleted as a startup test in section 14.2.12.2, is included as a startup test in figure 14.2-5 sheet 3, is also included in Section 14.2.12.1 with preoperational test abstracts, and yet is not included in Table 14.2-2 or Figure 14.2-4. Revise the applicable sec~ions to address the internals vibration tests.

RESPONSE

Reactor internals are tested in accordance with provisions of Regulatory Guide 1.20, Revision 2, for Non-prototype category I plants, as described in FSAR section 3.9.2.4. This testing is performed prior to fuel load in TP 2.16, "Reactor Internals Vibration and Inspection," an abstract of which is included in Section 14.2.12.1 as requested in an earlier item. The Startup Test "RPV Internals Vibration" was deleted since the testing is not repeated after fuel load. The number 34 was subsequently reassigned to the "Rod Sequence Exchange" Startup Test.

Rev. 46, 06/93 423.47-1

SSES-FSAR QUESTION 423.48 Your response to item 423.22 states that P30.2 and P28.1 are only test that will be conducted on Unit 1, and not on Unit 2.

Modify test descriptions for P13.1 and ST-31 to indicate that testing will be accomplished on both units, or modify your response to item 423.22 to justify not conducting Pl3.l 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 for Pl3.l has been 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).

Rev. 46, 06/93 423.48-1

SSES-FSAR QUESTION 423.49 Your response to several subitems of 423.12 are not acceptable.

Provide the requested information:

(1) 64.79 - Modify preoperational test descriptions P81.l 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% 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 125\

test documentation. Testing at 100% of rated capacity is accomplished during the preoperational test program by TP2.23. An abstract of TP2.23 follows P99.1.

The refueling bridge main hoist (1200 pound capacity) will be tested to 125% of capacity utilizing a Technical Procedure. Preoperational Test P81.l 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 approximately 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 insulation. 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 Atmosphere and Main Stearn Tunnel Cooling," demonstrates that the temperature of the atmosphere inside the drywell is maintained within design limits.

With the reactor at rated temp. during the drywell inspection (described in ST-17) a check will be made to estimate the concrete temp. surrounding one of the main steaml'ine penetrations by measuring the temperature at several accessible points on the containment liner plate or containment concrete surface.

Rev. 46, 06/93 423.49-1

SSES-FSAR QUESTION 423.50 Provide testing to verify that containment spray nozzles and headers, 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.l (RHR System Preoperational Test) provides for testing of the containment drywell 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.

The containment wetwell spray nozzle test consists of directing RHR system water through the containment wetwell spray header and verifying that each nozzle is not plugged and is spraying water. 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 similar drywell spray nozzle has been accomplished in the factory as described in FSAR Subsection 6.2.2.2.

Rev. 46, 06/93 423.50-1

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 their design temperature range with- the equipment operating in a manner that will produce the maximum heat load in the 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 that it is not apparent that post-accident design heat loads will be 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, Heat exchangers 1E230B, 1E230D, 1E217B, 1E217D, 1E218B, 1E218D, 1E257B, 1E231B, 1E231D, 1E229A and 1E229B were performance tested by the vendor to demonstrate conformance to design specifications. During the preoperational test program, the ESF equipment room coolers air flow and cooling water flows will be measured as part of hydronic balancing and air balancing procedures. These balancing procedures provides a comparison of design values and actual values for the heat load encountered. On the basis of meeting the design specification for heat removal the procedures will validate the vendor performance tests for design maximum heat removal.

Rev. 46, 06/93 423.51-1

SSES-FSAR QUESTION 423.52 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 1.68 (revision 1, 1/7?), 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 priqr to high power turbine trips and generator load rejects.

Rev. 46, 06/93 423.52-1

SSES-FSAR QUESTION 423.53 Modify ST-31 to provide assurance that 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 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.

Rev. 46, 06/93 423.53-1

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 P59.1.

Rev. 46, 06/93 423.54-1

SSES-FSAR QUESTION 423.55 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.

Rev. 46, 06/93 423.55-1

SSES-FSAR QUESTION 423.56 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

Question 423.12 has been revised to incorporate the responses to Questions 423.32 and 423.45.

Rev. 46, 06/93 423.56-1

SSES-FSAR Question Rev. 4 7 QUESTfON 423. 57 Modrfy ST-25 to address the following:

( 1} The present method for determining MSIV closure times is inaccurate. Modify the test method to measure the full travel of the valves or provide technical justification for extrapolating the full dosure 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 time.

(2) Provide a description of a test which demonstrates that the MSIV-LCS 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. The valves must close fast enough to limit the release of reactor coolant, and they must close slow enough so that simultaneous closure of all steamlines will not induce transients that exceed the nuclear steam design limits. MSIV closure time is calculated using limit switches which actuate when valve stem travel indicates 10% and 90% valve closure. Extrapolations using this data assumes linear valve closure.

Two equations are necessary to accurately calculate elapsed trmes. The slow criteria equation must include the delay time from solenoid deenergization to valve stem movement, whereas the fast criteria equation, which is concerned onty wrth valve movement, does not include this delay. The two equations are:

(1) for fast criteria Tc= 1.25 (Tso - T,o)

(2) for slow criteria Tcwct = Tgo + 0.125 {T90-T10)

FSAR Rev. 59 423.57-1

SSES-FSAR Question Rev. 47 where:

valve closure time, excluding delays valve closure time, with deJays Tgo = elapsed time .from solenoid deenergization to valve 90% closed T,o = elapsed time from solenoid deenergization to valve 10% closed.

(2) The MSIV-LCS is designed to control and minimize the release of fission products which could leak through closed MISV's following a LOCA. The MSIV-LCS is initially verified operable in P83.1 using air and subsequently verified operable on a periodic basis in accordance with Technical Specifications. Complete system testing and isolation valve leak testing is performed only during reactor shutdown to preclude inadvertent steam discharge. Interlocks are provided to preclude system operation at excessive MSIV leak rates. No further Startup Testing is deemed necessary.

FSAR Rev. 59 423.57-2

SSES-FSAR QUESTION 423.58 Update Table 14.2-3 (Startup Test Procedures) and Figure 14.2-5 (Individual Startup Test Sequence) to reflect the current status of Subsection 14.2.12.2.

RESPONSE

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

Rev. 46, 06/93 423.58-1

ML23292A215

Text

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

RESPONSE

Navigation menu

Search