ML20003E750

Responds to NRC 810126 Ltr Providing Addl Guidance Re NUREG-0737,Item I.G.1.Addl Safety Analyses & Waivers of Certain Tech Specs Would Be Required Re Training During Low Power Testing.Encl Program Completes Response

ML20003E750
Person / Time
Site:	Grand Gulf
Issue date:	04/07/1981
From:	Dale L MISSISSIPPI POWER & LIGHT CO.
To:	Tedesco R Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737 AECM-81-84, TAC-61196, NUDOCS 8104100412
Download: ML20003E750 (20)

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O MISSISSIPPI POWER & LIGHT COMPANY Helping Build Afississippi P. O. B O X 16 4 0. J A C K S O N. MIS SIS SIP PI 3 9 2 0 5 April 7, 1981 _.

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huCLEAR PRooUCT1oM oEPARTMENT Mr. Robert L. Tedesco, - $'  :

Assistant Director for Licensing ..

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j Division of Licensing f U.S. Nuclear Regulatory Commission -. .,'

Washington, D.C. 20555 *l

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Dear Mr. Tedesco:

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SUBJECT:

Grand Gulf Nuclear Station Units 1 and 2 Jocket Nos. 50-416 and 50-417 -

File 0260/0277/0755/L-860.0 TMI Action Item I.G.1, Training During Low Power Testing AECM-81/84 This letter is in reply to your letter, received January 26, 1981, which provided additional guidance in regard to NUREG-0737 Item I.G.1 as applied to BWR OL applications. That letter asked that we conduct during the Grand Gulf startup testing a simulated loss of AC power test.

We were also asked to commit to augmented operator training by their participation in the preoperational and startup test programs.

In regard to the proposed simulated loss of AC power test, the only significant experience to be gained over testing already planned and described to you in FSAR Chapter 14.2 is in the area of RCIC operation.

Therefore, for reasons provided below, extensive testing of RCIC under loss of AC power conditions is proposed as an alternative to your simulated loss of AC power test. .

Your letter suggests performing the loss of AC power test with a "real or simulated source of decay heat." In order to simulate decay heat, the reactor must remain critical at approximately 5 percent rated power. This would require that certain systems remain energized to assure the simulated decay heat can be generated. All other systems would be " blacked out" from loss of AC power (loss of offsite power and loss of emergency diesel generators).

The first point which should be made is that testing or operation in such a degraded operational state is beyond the scope of existing safety analyses and are not allowed by current Technical Specifications.

Thus, additional safety analyses and waivers of certain Technical bog)/

Specifications would be required.

.5 Second, despite the results of such safety analyses, we feel it would be difficult if not impossible to maintain the 5 percent power condition and simulate other relevant reactor and plant conditions under

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loss of AC power simultaneously. For example, on a loss of AC 100Qlg Member Middle soutn utilities system

MISSisslPPI POWER O LICHT COMPANY AECM-81/84 Page 2 condition, the Reactor Protection System (RPS) HG sets would trip on underv71tage resulting in a loss of power to the RPS trip circuitry which would cause a reactor scram. Power to the RPS would have to be maintained to prevent a reactor scram in order to maintain the simulated decay heat. Another normal and relevant action that would occur upon loss of RPS power would be the actuation of the Nuclear Steam Supply Shutoff System (NS4) and the closure of the MSIV's. With the reactor mode switch in the Startup/ Hot Standby position a direct scram from MSIV closure would be avoided. However, upon closing the MSIV's there is no exit for the steam flow, and it is extremely likely that the resulting pressurization transient would cause a reactor scram on high reactor pressure arj/or high neutron flux resulting from void collapse (IRL' hi-hi). This scram would also terminate the " simulated decay heat" and make any further testing futile. Since the results of the test depend upon the continued generation of steam at 5 percent power, it appears that any attempt to conduct a test with " simulated decay heat" [s impractical and would not achieve the desired results.

The remaining proposed alternative is performing the loss of AC test with actual decay heat within the first 1500 MWD /T core exposure immediately following 7 days of operation at 80% rated power or above.

Initiation of a loss of AC from these conditions would result in a coast down of the RPS MG sets within a few seconds which would result in a reactor scram and loss of power to the NS4 which again results in MSIV closure. The decay heat causes rapid pressurization and subsequent actuation of safety / relief valves blowing steam to the suppression pool.

l From this point on, the transient is similar to a MEIV closure with no i loss of AC power. Steam blowdown to the suppression pool lowers reactor water level to level 2 initiating RCIC. The design basis for RCIC is that it be capable of recovering water level to normal from precisely I this condition. Therefore, RCIC would return reactor water level to level 8 and then trip on high level. This cycle (starting and tripping RCIC) would continue for several hours with adequate RCIC suction supply from the condensate storage tank backed by the suppression pool. The addition of heat to the suppression pool through the SRV's would be less than that assumed in the LOCA analyses since the initial tett power level would be less than or equal to rated thermal power. Thus, up to the point at which suppression pool cooling is established in an accident, the loss of AC test is essentially identical (from the reactor response standpoint) to the MSIV closure transient.

A review of plant instruments has established that reactor vessel l pressure and level instrumentation does exist which is powered from DC or uninterruptible AC power sources. This would allow adequate monitoring of reactor vessel conditions during the test. While a loss of AC power test represents an unanalyzed condition of operation, a preliminary assessment shows that expected plant response is within the design basis of key systems, e.g., RCIC and important reactor instrumentation. The only operation which is essential 13 responding to this transient is the injection of water to the reactor by RCIC with no AC power present. The operational characteristics and capabilities of l

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MisslSSIPPI POWER 8: LIGHT COMPANY AECM-81/84 Page 3 RCIC under loss of AC present the only operationally significant portion of this transient which is different from the MSIV full closure test

already planned as part of the power ascension test program (see FSAR section 14.2.12.3.22.1). Therefore, the extended operational tests of RCIC included in Appendix C of the enclosure to this letter are proposed as an alternative to the loss of AC pover tes't.

These tests will be conducted in conjunction with the Integrated i ECCS testing and thus will be combined with the actual partial loss of AC power conditions which are effectively the sasa as a complete loss of

. AC power for RCIC.

Two additional tests, also described in Appendix C of the enclo'sure, are planned. The first is an integrated reactor pressure s

vessel level instrument functional test which will test the proper operation of all RPV level instruments during the preoperational test l phase. The second is an integrated drywell and containment pressure instrumentation test to be done in conjunction with the containment integrated leak rate test (ILRT). This test will prove the proper

- operation of all drywell and containment pressure instruments.

These new tests including the expanded RCIC tests will be l incorporated into existing detailed, written preoperational test procedures.

l The enclosure also describes MP&L's program for augmented operator training during the preoperational and startup test phases. This l

program is based on guidelines which were developed by the BWR Owners' Group in which MP&L actively participated. The Owners' Group guidelines have been specifically adapted for the GGNS startup and training l

programs.

As previously committed in FSAR Appendix 3A, the completed test procedures, data forms, graphs, photographs, etc. constitute the official historical record of the preoperational test program during which the additional tests reqaired by ites I.G.1 are conducted. Copies of this preoperational test documentation will be made available to the NRC for their review. A Startup Summary Report describing startup test phase activities will be submitted in accordance with Regulatory Guide 1.16.

We believe the program of additional testing and augmented operator i training which we have described constitutes a complete and satisfactory l response to ites I.G.1 of NUREG-0737.

l i Yours truly, jfY W

L. F. Dale f

Nuclear Project Manager MRW/CLT/JDR:Im I Enclosure cc: (See Next Page)

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MISSISSIPPI POWER & LIGHT COMPANY AECM-81/84 Page 4 cc: Mr. N. L. Staapley Mr. G. B. Taylor Mr. R. B. McGehee Mr. T. B. Conner Mr. Victor Stello, Jr. , Director Division of Inspec ion & Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555 1

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GRAND GULF NUCIE.AR STATION PROGRAM FOR TRAINING DURING LOW P0kT.R TESTING NUREG-0737, ITEM I.G.1 I

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GRAND GUIE NUCLEAR STATION PROGRAM FOR TRAINING DURING LOW POWER TESTING NUREG-0737, ITEM I.G.I INDEX SECTION TITLE PAGE INTRODUCTION 1 I PREOPERATIONAL TESTING 2 II COLD FUNCTIONAL TESTING 2 III HOT FUNCTIONAL TESTING 3 IV STARTUP TESTING 3-4 V ADDITIONAL TRAINING AND TESTING 4 CONCLUSIONS 4 APPENDIX A COLD FUNCTIONAL TESTING 5-6 APPENDIX B HOT FUNCTIONAL TESTING 7-9 APPENDIX C ADDITIONAL TESTING AND TRAINING 10 - 14

I INTRODUCTION The NRC has identified new requirements for GE BWR plant testfag and training. These requirements in NUREG-0737, ites I.G.1, are applicable to near-term operating license (NTOL) facilities. The following quotes are from the earlier NUREG documents addressing these requirements:

NUREG-0660 May, 1980 TASK I.G PREOPERATIONAL AND LOW-POWER TESTING

< A. OBJECTIVE: Increase the capability of the shift crews to operate facilities in a safe and competent manner by assuring that training for plant evolutions and off-normal events is conducted. Near-ters operating license facilities will be required to develop and implement intensified. training exercises during low-power testing programs. This may involve the repetition of startup tests on different shifts for training purposes. Based on experiences frce the near-term (perating license facilities, requirements may be applied to other new facilities or incorporated into the plant drill requirement (Item I.A.2.5). Review comprehensiveness of test programs.

NUREG-0694 June, 1980 I.G.1 TRAINING DURING LOW-POWER TESTING r

> Define and commit to a special low-power testing program approved by NRC to be conducted at power levels no greater than 5 percent for the purposes of providing meaningful technical information beyond that obtained in the normal startup test program and to provide supplemental training.

NUREG-0737 reiterated the requirements cet forth in NUREG-0694.

Review of the present BWR Preoperational and Startup Test Programs against the above listed requirements has identified a number of areas where increased emphasis on operator training can be beneficial.

Additionally, several new tests have been identified that are responsive

. to the NRC requirements. As a result of this review, a program Las been developed for the Grand Gulf Nuclear Station and is described herein.

The test program has been divided into five sections for purposes of this report. They are:

I - Preoperational Testing II - Cold Functional Testing III - Hot Functional Testing IV - Startup Testing V - Additional Training and Testing .

The first four sections briafly discuss the present test program and changes made to improve the training benefit. The last section contains new testing proposed te provide meaningful cechnical information and enhance training.

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I PREOPERATIONAL TESTING ,

The preoperational test program serves a two-fold purpose.

Primarily, it controls and documents the preoperational test effort. A secondary benefit of the program is that during the test phase, a detailed knowlege of the systems and their performance characteristics will be obtained by the plant operating personnel. ,

Plant operating personnel will obtain hands on experience by testing of these systems thereby helping to satisfy the training concerns of NUREG-0737. The preoperational test program described in FSAR Section 14.2.12.1 will be conducted by plant operations

personnel where operation of plant equipment is involved and thus i

contributes significantly to operations personnel training on these I systems. The Integrated ECCS with Loss of Offsite Power test.is one of the more significant tests performed during the l

preoperational test phase which significantly supports operator training. FSAR Section 14.2.12.1.44 describes this test.

i To enhance the training benefit of this test, Integrated ECCS

testing will be scheduled so that each shift will participate in at least one of these tests to obtain training. ' Operators will obtain

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an appreciation and feel for control room and plant conditions / limitations and will be required to resolve operational problems associated with the loss of emergency battery and diesel generators during a time when emergency equipment is required to operate.

II COLD FUNCTIONAL TESTING l

l Cold Functional Tests are performed on a plant for several rearons.

Some of the more important reasons are as follows:

A. Assure that plant systems are available to support fuel loading.

B. Assure that shift personnel have operating experience with plant equipment.

C. Assure that certain plant operating procedures and surveillance procedures have been tried and are usable.

D. Assure that each shift has functioned together to operate the l_ plant systems on sa integrated basis.

E. Assure that specified plant equipme. t has been tested and the plant and personnel are ready for f'.el loading.

The Cold Functional Tests are performes using plant procedures and are controlled and documented by use of checklists. The checklist provides a signoff sheet to assure that each shift has received training and cxperience on some of the specified systems. The Shift Supervisor will be responsible to ensure, by signing the checklist, that his shift has performed the operation specified.

Systems to be included are found in Appendix A.

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III HOT FUNCTIONAL TESTING Hot Functional Tests are performed to assure that as much as possible the system, procedures, and personnel are ready for operativa at various power levels. This verification is done by operating systems in an integrated fashion at operating temperatures and pressures at the earliest opportunity for meaningful checks.

The Hot Functional Tests cover those areas of the plant systems which are not tested by the Startup Test Procedures, but where it is felt that additional data over and above the Cold Functional Tests is beneficial.

Typically, the Hot Functional Tests will begin after fuel is loaded when nuclear heat is available. Startup testing provides two phases which of fer Hot Functional Test opportunities. These phases are listed below:

A. During heatup from ambient temperature and 0 psig reactor pressure to rated reactor temperature and pressure.

B. After increase from rated reactor temperature and pressure to 30 percent reactor power.

The F;; Functional Tests are not intended to replace any of the startup test procedures, although there are portions which will be conducted simultaneously.

Those systems whose environment does not change during ascension to rated temperature and pressure will nct receive' additional testing.

Hot Functional Tests to be performed on systems are listed in Appendis B.

During the performance of this testing the Operations Superintendent shall cause a review to be performed of d2e applicable system operating procedures and ensure that necessary changes are made to these procedures as specified by admini:trative procedures. The Training Superintendent will additionally verify that operations personnel on each shift have been familiarized with the changes to procedures through the use of information acknowledgements.

IV STARTUP TESTING A typical startup test program is composed of phases characterized by differences in plant and test conditions. Startup tests are comorised of four phases which include fuel loading and subsequent tests:

A. Open Vessel Testing B. Initial Heatup C. Power Tests D. Warranty Tests Tests to be performed during open vessel, reactor heatup and power ascension are described in FSAR Section 14.2.12.3.

All manipulcticn of centrole fer startup tseting purp:Ocs will ba cerductsd by qunlifisd cparetiens per:cnn21. This tacting will provide many beneficial opportunities for operator training. In particular, this training shall verify that each operating shift observes or performs the following:

1. At least one reactor scram transient.

2. At least one pressure controller transient.

3. At least one turbine trip or load rejection transient.

4. Operation of the RCIC system.

5. At least one water level setpoint transient.

Since testing will, in general, be conducted on a round the clock basis, the other testing will be balanced as much as practicable to ensure even exposure to testing for all operating shifts.

V ADDITIONAL TRAINING AND TESTING Because of our efforts to provide as comprehensive a test program as possible several new tests will be added to the test program.

These tests will provide additional technical information to aid in system and plant operational readiness evaluations. ihe tests will also provide some operator training by the operator participation i in the conduct of the testing.

Appendix C contains test descriptions defining the scope of the new tests to be added to the preoperational test program. Detailed test procedures will be written from the scope of those

descriptions.

CONCLUSIONS As explainea is this report, each phase of the testing program provides a building block for the next phase and provides the necessary oveelap and depth to ensure that the operating staff will obtain maximum meae.ingful inplant training to assure that crews will operate their facilities in a safe and competent manner and that all safety related systems are thoroughly tested. The increased emphasis on operator training, described in this l

submittal, and the addition of new testing, when coupled with the

! present testing and training programs, more than adequately l

satisfies the requirements of I.G.1 of NUREG-0737.

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APPENDIX A Systems to be included as part of this program are:

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Reactor Vessel & Auxiliary Systems Recirculation System Reactor Water Cleanup System Control Rod Drive System -

Reactor Vessel Level Instrumenta, tion Standby Liquid Control Remote Shutdown System ECCS System LPCS RHR (including LPCI, Shutdown Cooling, and Suppression Pool Cooling)

HPCS Emergency Electrical System i

Standby Diesel Generators and Emergency Buses Emergency Batteries Plant Support Systems Stcudby Service Water Component Cooling Water Turbine Building Cooling Water Make'ap Water Treatment System Fuel Pool Cooling and Cleanup System Condensate / Refueling Water Transfer System Instrument and Service Air Plant Service Water Circulating Water Auxiliary Steam System

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APPENDIX A 1

other Systems Standby Gas Treatment Systen l Auxiliary Building HVAC 1

Fuel Handling HVAC i Control Room HVAC h

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APPENDIX B DURING HEATUP FROM AMBIENT TEMPERATURE AND 0 PSIG REACTOR PRESSURE TO RATED REACTOR TEMPERATURE AND PRESSURE SYSTEM MODE OF OPERATION AND HOT FUNCTIONAL TESTS CRD S,vstem In continuous normal operation, check each fully withdrawn CRD for coupling as it is

- withdrawn. Observe that temperatures are in limits. Observe proper position indication. Record rod . patterns.

Containment and Drywell Both should be in continuous operation Cooling per procedure.

Process Radiation Monitors In continuous operation. Check ier response to increasing power levels.

Ventilation Systems In continuous operation. Check that oteam tunnel temperature is within temperature limits at rated temperature and pressure.

Verify proper operation of leakage detection systems.

Turbine EHC Pressure Start heatup with pressure controller set Controls at 150 psig and by-pass opening jack at O.

When reactor pressure is greater than 150 pais check that controller responds to setpoint changes.

i Rod Pattern Controller In continuous operation. Verify proper operation as rods are withdrawn.

Main Steam Relief Valves Record the discharge thermocouple readings from recorder and determine that the valves do not have seat leakage.

Condensate Cleanup System Verify performance of system to control l adequately water quality by observing that water quality stays within limits l

specified by plant chemist.

TIP System Make traces when flux level permits.

Verify air / nitrogen purge.

l Reactor Water Cleanup In continuous operation ;L approximately

! System 50 percent to 100 pa; cent flow. Place cleanup recirculation pumps in operation at pressure and operate in all modes.

Check that valves operate properly.

Reject reactor water back to condenser and radwaste to check reject valve for proper operation.

APPENDIX B Reactor Recirculation In continuous operation per operating System procedure. Check that seal cavity, bearing and winding temperatures are within limits. Check that cavity pressures follows heatup pressure.

Check that recire. loop temperature recorder indicates the proper temperature increase. ,

Condensate and Feedwater In continuous operation to maintain reactor level. Start standby feed pump turbine per procedure, place in service and remove replaced turbine from service.

SRM and IRM In continuous operation. Check for proper retract operation as they are withdrawn.

Insert and check for proper operation / indication.

Turbine Seal Place in continuous operation per operating prasedure. Check that seal steam regulator controls seal pressure.

Vacuus Pump Place in service per operating procedure.

Steam Jet Air Ejectors Place in service per operating procedure.

Place backup air ejectors in service.

Reactor Vessel Temps and Should be in continuous service. Head Head Leak Detection seal leak detector should be valved per operating procedure.

Circulating Water Continuous operation to maintain adequate condenser vacuum.

1 APPENDIX B AFTER INCREASE FROM RATED TEMPERATURE AND PRESSURE TO 30% P0hT.R A few significant system environmental changes occur between arrival at rated temperature and pressure and completion of 30 percent testing which requires the following additional hot functional checks. )

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NODE OF OPERATION AND HOT FUNCTIONAL j SYSTEM CHECKS l Turbine Generator During this period the turbine generator will be placed in operation on nuclear steam for the first time and the following checks should be performed which are not  !

part of the formal test program. Verify l procedure for turbine warmup and roll to )

1,800 rpm. Perform the turbine generator i

no-load tests. Check turbine vibration at '

critical speed and 1,800 rpm. Verify 1 proper operation of primary cooling system l (rotor and stator) and generator seal oil l systems. Verify operator familiarization j with turbine generator instrumentation and ,

controls both local and remote. Verify )

oil flow indication at each bearing i inspection spout. Verify that expansion ,

4 (stretchout) is satisfactory. Perform  !

overspeed checks. ]

Feedwater Heater Controls Put feedwater heaters in service, and establish level control. Feedwater i

temperature will rise. Inspect feedwater ,

line and feedwater pump casings to assure thermal expansion has not opened flanges i or affected mechanical seal operation. i CCW System Check temperatures of cooled components.

Readjust as necessary to maintain proper temperature in components as specified in the operating procedures.

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t APPENDIX C TEST: Startup of the RCIC system after loss of AC power to the system.

PURPOSE: Verify the design basis ability of the system to start without j the aid of AC power with the exception of the RCIC DC/AC inverters.

INITIAL CONDITIONS:

-o Preoperational test has been performed on RCIC system.

I o If test is performed prior to the availability of nuclear steam, sufficient auxiliary boiler capacity and piping to run the RCIC turbine / pump must be available, o System valves in normal standby lineup (suction from CST)

NOTE: 1) If the auxiliary boiler is used as the turbine steam supply, tag closed the steam supply isolation valves E51-F064 and E12-F052A & B.

2) Flow can either be directed to the reactor "

pressure vessel or back to the condensate storage tank.

, o Power to all RCIC components fed by site AC power shall be secured.

o Station batteries shall be fully charged.

o Instrument air shall be available for operation and control of applicable valves. L o Instruments shall be calibrated and setpoints, where i spplicable, shall be verified.

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o Perform a manual initiation of the RCIC system utilizing the manual i.sitiation switch and verify the proper operation of all components required for the RCIC startup tracalent to rated flow.

NOTE: Manual manipulation of some valves will be required if flow is returned to the CST or auxiliary boiler steam is used.

ACCEPTANCE CRITERIA:

o Proper operation of all components for the RCIC startup transient until rated flow is obtained. '

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APPENDIX C TEST: Operation of the RCIC system with a sustained loss of AC power to the system.

PURPOSE: To verify the operation of RCIC to evaluate the ILuits of system operation with extended loss of AC power to it and support systems with the exception of the RCIC DC/AC inverters.

INITIAL CONDITIONS:

o Preoperational test has been performed on RCIC system.

o If test is performed prior to the availability of nuclear steam, sufficient auxiliary boiler capacity and piping to run the RCIC turbine / pump must be available.

o System valves in normal standby linerp (suction from CST).

NOTE: 1) If the auxiliary boiler is used as the turbine steam supply, tag closed the steam supply isolation valves, E51-F064 and E12-F052 A & B.

o Power to all RCIC components fed by site AC power shall be secured, including RCIC area coolers and battery chargers supplying the station battery from which RCIC DC loads are powered.

o RCIC batteries shall be fully charged.

o Instrument air shall be available for operation and control of applicable valves.

o Instruments shall be calibrated and setpoints, where applicable, shall be verified.

TEST DESCRIPTION:

o Start and operate the RCIC system with return to the CST and run for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or until aar system limiting parameter is approached (e.g. , ' sigh RCIC area temperature, low battery voltar,e, or high suppression pool temperature). During thf.s period of RCIC system operation, shutdown the system and restart two times to verify systes restart capability.

NOTE: Testing will be conducted in a manner such that equipment is not placed in Jeopardy of damage and such that equipment qualification is not degraded.

ACCEPTANCE CRITERIA:

o None

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APPENDIX C TEST: RCIC operation to prove DC separation.

PURPOSE: Verify proper operation of the RCIC DC components when non-RCIC station batteries are disconnected.

INITIAL CONDITIONS:

o Preoperational test has been performed on RCIC system, o Test to be per2)rmed prior to fuel load, o This test is performed prior to the availability of nuclear steam, sufficient auxiliary boiler capacity and piping to run the RCIC turbine / pump must bc available.

o System valves in normal standby lineup (suction from CST).

o Steam supply isolation valves E51-F064 and E12-F052A & B-tagged shut.

o Station batteries shall be fully charged, o Instrument air shall be available for operation and control of applicable valves.

o Instruments shall be calibrated and setpoints, where applicable, shall be verified.

TEST DESCRIPTION:

o Start and operate the RCIC system with return to the CST.

During system operation disconnect, each non-RCIC station battery from its bus and verify proper operation of each RCIC DC component.

ACCEPTANCE CRITERIA:

o Proper operation of RCIC DC components with non-RCIC station batteries disconnected.

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APPENDIX C TEST: Integrated reactor pressure vessel level functional test.

PL1tPOSE: Verify that instruments connected to the RPV are tubed properly, that the tubing is not blocked and that instrument tracking is proper. .

INITIAL CONDITIONS:

o All instruments connected to the RPV have been calibrated and are operable.

j o RPV has been flushed and is clean.

o All RPV instrument tubing has been filled, all instruments are vented, and proper valve lineup verified.

o A source of desineralized water is available to fill the RPV.

o Fuel has not been loaded into the RPV.

o RPV head removed or adequately vented to prevent

. pressurization.

4 TEST DESCRIPTION:

o Raise and lower (or lower and raise, whichever is most convenient) the RPV water level through the range of RPV levels necessary to verify the proper operation and tracking of each RPV connected instrument.

NOTE: The temperature and pressure conditions at which this test will be performed are not the conditions for which the various instruments are calibrated. There will not be a one-to-one correspondence between actual reactor vessel level change and indicated level change.

ACCEPTANCE CRITERIA:

o Each affected RPV instruments operacion and tracking is satisfactory.

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APPENDIX C TEST: Integrated drywell and containment pressure instrumentation test (test to be performed in conjunction with containment integrated leak rate testing).

PURPOSE: Verify proper connection and tracking of drywell and containment pressure instruments and that the tubing supplying these instruments is not blocked.

INITIAL CONDITIONS:

o All initial conditions for containment integrated leak rate testing have been established.

. o Drywell and containment press'are instruments have been calibrated and are valved into service.

TEST DESCRIPTION:

o As containment pressure is increased, during the containment integrated leak rate test, verify proper operation of specified drywell and containment pressure instruments.

ACCEPTANCE CRITERIA:

o Specified drywell and containment instruments perform their design function. ,

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