ML20054B678

SEP Topic VI-10.A,Testing of Reactor Protective Sys & ESF, Technical Evaluation Rept

ML20054B678
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	02/01/1982
From:	Vanderbeek R EG&G, INC.
To:	NRC
Shared Package
ML20054B676	List: ML20054B675 ML20054B678
References
TASK-06-10.A, TASK-6-10.A, TASK-RR 0630J, 630J, NUDOCS 8204190053
Download: ML20054B678 (30)

Text

'

i 1

0630J SYSTEMATIC EVALUATION PROGRAM TOPIC VI-10.A TESTING OF REACTOR PROTECTIVE SYSTEM AND ENGINEERED SAFETY FEATURES LA CROSSE BOILING WATER REACTOR Docket No. 50-409 February 1982 R. VanderBeek EG&G Idaho, Inc.

2/1/82 8204190053 e20414 PDR ADOCK 05000409 P

PDR

A s

/

1 CONTENTS

1.0 INTRODUCTION

1 2.0 CRITERIA........................................................

I 3.0 REACTOR PROTE CTI VE SY STEM.......................................

4 3.1 Description...............................................

4 3.2 Evaluation................................................

6 4.0 ENGINEERED SAFETY FEATURES SYSTEM...............................

13 4.1 Description...............................................

13 4.2 Evaluation.................................................

14 5.0

SUMMARY

27

6.0 REFERENCES

27 TABLES 1.

Comparison of La Crosse Boiling Water Reactor RPS instrument surveillance requirements with BWR Standard Technical Specification requirements......................................

7 2.

Comparison of La Crosse Boiling Water Reactor Engineered Safety Features (ESF) instrument surveillance requirements with BWR Standard Technical Specification Requirements..........

15 I

ii L

4

s i

a SYSTEMATIC EVALUATION PROGRAM TOPIC VI - 10.A TESTING OF REACTOR PROTECTIVE SYSTEM AND ENGINEERED SAFETY FEATURES LA CROSSE BOILING WATER REACTOR

1.0 INTRODUCTION

The objective of this review is to determine if all reactor protective system (RPS) components, including pumps and valves, are included in com-ponent and system tests, if the scope and frequency of periodic testing is adequate, and if the test program meets current licensing criteria. The review will also address these same matters with respect to the engineered safety features (ESF) systems.

2.0 CRITERIA General Design Criterion 21 (GDC 21), " Protection System Reliability and Testability," states, in part, that:

The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a cap-ability to test channels independently to determine failure and losses of redundancy that may have occurred.I Regulatory Guide 1.22, " Periodic Testing of the Protection System Actuation Functions," states, in Section D.l.a, that:

The periodic tests should duplicate, as closely as practicable, the performance that is required of the actuation devices in the event of an accident; and further, in Section D.4, it states that:

When actuated equipment is not tested during reactor operation, it should be shown that:

1

a.

There is no practicable system design that would permit operation of tne actuated equipment without adversely affecting the safety or operability of the plant, b.

The probability that the protective system will fail to initiate the operation of the actuated equipment is, and can be maintained, acceptably low without testing the actuated equipn:ent during reactor operation, and c.

The actuated equipment can be routinely tested when the reactor is shut down.2 IEEE Standard 338-1977, " Periodic Testing of Nuclear Power Generating Station Class lE Power and Protection Systems," states, in part, in Sec-tion 3:

Overlap testing consists of channel, train, or load-group verification by performing individual tests on the various components and subsys-tems of the channel, train or load group. The individual component and subsystem tests shall check parts of adjacent subsystems, such that the entire channel, train or load group will be verified by test-ing of individual components or subsystems.3 and, in part, in Section 6.3.4:

i Response time testing shall be required only on safety systems or sub-systems to verify that the response times are within the limits of the j

overall response times given in the Safety Analysis Report.

Sufficient overlap shall be provided to verify overall system response, l

The response-time shall include as much of each safety system, from sensor input to actuated equipment, as is practicable in a single test.

Where the entire set of equipment from sensor to actuated equipment cannot be tested at once, verification of system response time shall be accomplished by measuring the response times of discrete portions l

l l

/

e of the system and showing that the sum of the response times of all is within the limits of the overall system requirement.

In addition, the following criteria are applicable to the ESF: Gen-eral Design Criterion 40 (GDC 40), " Testing of Containment Hcat Removal System," states that:

a The containment heat removal system shall be designed to permit appro-priate periodic pressure and functional testing to assure:

a.

The structural and leaktight integrity of its components.

b.

The operability and performance of the active components of the system.

c.

The operability of the system as a whole and under conditions as close to the design as

• al, the performance of the full operational sequence t

_,ings the system into operation, including operation of applicable portions of the protection systems, the transfer between normal and emergency power sources, and the operation of the associated cooling water system.4 GDC 37, " Testing of Emergency Core Cooling System," GDC 43, " Testing of Containment Atmosphere Cleanup Systems," and GDC 46, " Testing of Cooling Water System," are similar.

Standard Review Plan, Section 7.1, Appendix B, " Guidance for Evaluation of Conformance to IEEE STD 279," states, in Section 11, that:

Periodic testing should duplicate, as closely as practical, the over-all performance required of the protection system. The test should confirm operability of both the automatic and manual circuitry. The capability should be provided to permit testing during power operation.

When this capability can only be achieved by overlapping tests, the test scheme must be such that the tests do, in fact, overlap from one test segment to another.

3 1

t i

i 3.0 REACTOR PROTECTIVE SYSTEM (RPS) 3.1 Description. The Reactor Protection System (RPS) includes the sensors, amplifiers, logic and other equipment essential to the monitoring of selected nuclear power plant conditions.

It must reliably effect a. rapid shutdown of the reactor if any one or a combination of parameters deviates beyond preselected values to mitigate the consequences of a postulated design basis event.

The RPS parameters and their logic channels, as identified in the j

La Crosse Technical Specification,6 are as follows:

PARAMETER:

1.

Manual Reactor Trip 2.

Source Range, heutron Flux (Nuclear Channels 1 and 2) 3.

Source Range, Reactor Pe;iod-Short (Nuclear Channels 1 and 2) i 4

Intermediate Range, Neutron Flux (Nuclear Channels 3 and 4) 5.

Intermediate Range, Reactor Period-Short (Nuclear Channels 3 and 4) 6.

Power Range, Neutron Flux a.

Reactor Power-High < 15 2 2% indicated power on Nuclear Chan-nel 7 and 8 (Nuclear Channels 5 and 6) b.

Reactor Power-High > 15 2 2% indicated power on Nuclear Chan-nel 7 and 8 (Nuclear Channels 5, 6, 7 and 8 with Automatic Gain Control) i 7.

Reactor Pressure-High (Pressure Safety Channels I and 2) t 4

~-

8.

Reactor Power-to-Forced Circulation Flow Abnormal, (Power-Flow Safety Channels 1 and 2) 9.

Reactor Coolant Flow Rate-Low (Power-Flow Safety Channels I and 2)

10. Reactor Water Level-High (Water Level Safety Channels 1 and 2) and Low Water Level Safety Channels 1, 2 and 3.

11. Main Condenser Vacuum-Low (Main Condenser Vacuum Switches 1 and 2)

12. Main Steam Isolation Valves a.

Containment Bldg. MSIV Nat Fully Open (Valve Closure Relays 1 and 2) b.

Turbine Bldg. MSIV Not Fully Open (Valve Closure Relays 1 and 2) c.

Turbine Stop Valve Not Fully Open (Valve Closure Limit Switch)

13. Control Rod Drive Accumulators a.

Oil Level-Low (Limit Switch) b.

Gas Pressure-Low (Pressure Switch) 14 Bus Voltages a.

2400 v Bus lA-Low Voltage (Undervoltage Relays Phases A and C) i b.

2400 v Bus 18-Low Voltage (Undervoltage Relays Phases A and C) 4 c.

2400 v Bus lA-Low Voltage (Undervoltage Relay Phase A) and 2400 v Bus IB-Low Voltage (Undervoltage Relay Phase A) d.

2400 v Bus lA-Low Voltage (Undervoltage Relay Phase C) and 2400 v Bus 1B-Low Voltage (Undervoltage Relay Phase C) 5 t-

i e.

Containment Bldg. MCC-1A-Low Voltage (Undervoltage Relays Phases A and C) f.

Turbine Bldg. MCC-1A Low Voltage (Undervoltage Relays Phases A and C)

15. Reactor Scram Relays

16. Automatic Scram Logic 3.2 Evaluation. Table 1 provides a comparision between the require-ments for surveillance as established by the BWR Standard Technical Speci-fications (STS) and those set forth by the La Crosse Boiling Water Reactor Technical Specifications.

Evaluation of the La Crosse Technical Specifications indicate that:

1.

Six of the La Crosse protective systems comprising the RPS cor-respond to the BWR Standard Technical Specification RPS; however the Primary Containment Pressure for La Crosse does not actuate a scram function.

2.

The STS requires channel calibration at least once per refueling outage (18 months) for the Intermediate Range Neutron Flux High, the Main Steam Line Isolation Valve-Closure, and the Turbine Stop Valve-Closure. The La Crosse Technical Specifications do not require channel calibration for these three parameters.

3.

The La Crosse Technical Specifications include in the RPS the Main Condenser Vacuum Low, the Control Rod Drive Accumulators, the Bus Voltages, the Reactor Scram Relays, and the Automatic Scram Logic parameters. There is no channel check performed on these parameters. The STS do not require these parameters for the RPS and therefore there are no requirements set forth for a 6

l l

TABLE 1.

COMPARISON OF LA CROSSE BOILING WATER REACTOR RPS INSTRUMENT SURVEILLANCE REQUIREMENTS WITH BWR STANDARD TECHNICAL SPECIFICATION REQUIREMENTS (STS)

CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST IS REQUIRED a

STS LA CROSSE STS LA CROSSE STS LA CROSSE STS LA CROSSE 1.

Intermediate Range Moni-tors:

a.

Neutron Flux--High D

0(K)

R S/U(b)(c),

S/U(k) 2,3,4,5 Level W

2,3,4,5 b.

Inoperative NA NA W

2.

Average Power Range Moni-tor:

a.

Neutron Flux--High, 15% S b),

b)(c),

2 b.

Fixed Biased Neutron S

W(e)(f),

S/U(b),y j

Flux--High SA c.

Fixed Neutron Flux--

S W(d),SA 5/U(b),y j

liigh, 120%

W 1,2,5 d.

Inoperative NA NA W

1 e.

Downscale NA NA f.

LPRM D

(g)

NA 1, 2, 5 3.

Reactor Vessel Steam Dome NA Q

1, 2 M

Pressure--liigh

TABLE 1.

(continued) 1 CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE I

CHECK CALIBRATION TEST IS REQUIRED a

STS LA CROSSE STS LA CROSSE STS LA CROSSE STS LA CROSSE 4.

Reactor Vessel Water D

D Q

R M

M 1, 2 Level--Low, Level 3 5.

Main Steam Line Isolation NA R(h)

M S/U, M 1

Valve--Closure 6.

Main Steam Line Radiation D

R(3)

W(I) 1, 2

--High Q

R M

1, 2 7.

Primary Containment NA i

Pressure--High co 8.

Scram Discharge Volume NA R(h) 1, 2, 5 Q

Water Level--liigh 9.

Turbine Stop Valve--Closure NA R(h)

M S/U, M 1

1

10. Turbine Control Valve NA R

M Fast Closure, Trip Oil Pressure--Low t

11. Reactor Mode Switch in NA NA 1,2,3,4 R

Shutdown Position 5

12. Manual Scram NA NA NA NA Q

M 1, 2, 3, 4, 5

13. Reactor Power-to-Forced D

R M

Circulation Flow Abnormal

TABLE 1.

(continued)

CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CAllBRATION TEST IS REQUIRED a

STS LA CROSSE STS LA CROSSE STS LA CROSSE STS LA CROSSE

14. Reactor Pressure--High D

R S/U, M

15. Wide range and power range (channels 5, 6, 7, and 8 (k))

a.

Nuclear Instrumenta-M M

tion and Automatic Gain Control Sub-System b.

Nuclear Instrumenta-M, S tion and Automatic Gain Control Sub-

System, c.

Automatic Gain Con-M R

tol Sub-System

16. Reactor Water Level High D

R

'M

17. Reactor Coolant Flowrate D

R M

Low

18. Main Condenser Vacuum Low S/U, M R

19. Control Rod Drive W. M l

Accumulators

20. Bus Voltages R

l

\\

TABLE 1.

(continued) s i

CilANNEL MODES FOR WHICH CilANNEL CilANNEL FUNCTIONAL SURVEILLANCE CilECK CALIBRATION TEST IS REQUIRED STS LA CROSSE STS LA CROSSE STS LA CROSSE STS LA CROSSE a

M

21. Reactor Scram Relays M

22. Automatic Scram Logic TABLE 1.

NOTATION i

S

- At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> D

- At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> E5 W

- At least once per 7 days M

- At least once per 31 days Q

- At least once per 92 days R

- At least once per refueling outage (18 months)

Not applicable N.A.

SA

- At least once per 6 months S/U

- Prior to start up

- Not performed or available function 1.

POWER OPERATION 2.

STARTUP

TABLE 1.

(continued)

TABLE 1.

NOTATION (continued) 3.

HOT SHUTDOWN p

4.

COLD SHUTDOWN 5.

REFUELING (a) Neutron detectors may be excluded from CHANNEL CALISRATION.

(b) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.

(c) The IRM, APRM and SRM channels shall be compared for overlap during each startup, if not performed within the previous 7 days.

(d) When changing from CONDITION 1 to CONDITION 2, perform the required surveillance within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter entering

_. CONDITION 2.

(e) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL POWER. Adjust channel if the absolute dif ference > 2%.

('f) This calibration shall cansist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.

(g) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH) using the TIP system.

(h) Physical inspection and actuation of position switches.

(i)

Instrument alignment using a standard current source.

(j) Calibration using a standard radiation source.

(k) Testing of the Nuclear Instrumentation and Automatic Gain Control Sub-System shall be done concurrently.

channel check.

It is left to the NRC Staff to determine whether there is enough plant operating experience to justify no periodic channel check of the above parameters.

4 The La Crosse Technical Specifications do not require a channel calibration for the following parameters.

Wide Range and Power Range (channels 5, 6, 7, and 8).

Control Rod Drive Accumulators Bus Voltages Reactor Scram Relays Automatic Scram Logic.

The STS does not address these parameters; therefore there are no established requirements for periodic calibration of these parameters. It is noted that these parameters are subjected to periodic channel functional testing.

It is left to the NRC Staff to determine whether there is enough plant operation experience to justify no periodic channel calibration and that channel functional testing adequately supports the justificatiori for no periodic channel calibration.

5.

There are no requirements specified in the La Crosse Technical Specifications establishing the modes for which surveillance is required.

6.

There are no specific testing requirements established in the La Crosse Technical Specifications for the manual or automatic scram.

It is assumed that a monthly test of the full scram cir-cuits includes both manual and automatic. However, a test for hot short by means of built-in test switch does not comply with current licensing criteria.

k 7.

There are no specific requirements established in the La Crosse t

Technical Specifications for response times for those systems I

comprising the RPS.

12

4.0 ENGINEERED SAFETY FEATURES SYSTEM 4.1 Description. The Engineered Safety Features System consists of the High Pressure Core Spray, the Alternate Core Spray (Low Pressure Cool-ant Injection), the Containment Isolation, and the Shutdown Condenser Sys-tem.

The High Pressure Core Spray (Safety Injection) System is designed to automatically actuate the injection of water from the overhead storage tank into the core spray header which supplies the lines leading to a spray noz-zie just above each fuel assembly.

Operation of the High Pressure Core Spray (Safety Injection) System is initated automatically by an actuation signal generated upon low reactor water level or high containment building internal pressure. The high pressure core spray pumps can be started and tripped with individual control switches in the control room. However, an interlock prevents manual starting of the pumps unless the reactor is scrammed.

The Low Pressure Coolant Injection System is designed to provide addi-tional cooling water to the core under the conditions of low reactor pres-sure, low reactor water level, and high reactor building pressure which would exist following the maximum credible accident (MCA). This system also provides the means of flooding the reactor building following an MCA.

Approximately 900 gpm of cooling water can be supplied through a nozzle in 1

the top head of the reactor vessel and impinged on perforated deflector plates located above the reactor core. The cooling water is supplied from the river by either of two diesel-driven service water pumps located in the crib house. Two parallel control valves are provided in the line to the reactor and both open on signal, providing redundant paths for coolant injection.

The Containment Isciation System is designed to establish containment integrity within a short time after a major system rupture, preventing escape of fission products from damaged fuel elements to the outside atmosphere. Operation of the Containment Isolation System is initiated automatically by an actuation signal generated upon high containment 1

13 1

l

radiation monitor, high containment building pressure, low water level channels 1 and 2, low steam line pressure and low vacuum, and high reactor pressure. The containment isolation signals can be reset in the control room. All control valves are designed so that initiating the reset signal does not cause automatic reopening of any valve.

The Shutdown Condenser System is designed to condense reactor steam when the reactor is isolated from the main condenser upon closure of the reactor building steam isolation valve or the turbine building steam isola-tion valve. The system is automatically actuated when (1) the reactor building steam isolation valve is not fully open, (2) the turbine building steam isolation valve is not fully open, or (3) the reactor pressure is above 1325 psig. When the system is initiated automatically, the steam inlet valve and the off-gas vent valve to the waste gas system open immediately.

4.2 Evaluation. Table 2 provides a comparison between the BWR Stan-dard Technical Specification requirements and those of La Crosse Boiling Water Reactor Technical Specifications for the surveillance of the Engineered Safety Features (ESF) System.

Evaluation of the La Crosse Technical Specification indicate that:

1.

During each reactor shutdown for major refueling or an interval l

no greater than one year, the valves for the Containment Isola-tion System are tested to demonstrate their operability. The Containment Isolation System utilizes the Low Reactor Vessel

(

Water Level parameter which corresponds to that required by the

(

STS and utilizes the High Containment Building Internal Pressure, the High Containment Radiation Monitor, the Low Steam Line Pres-sure and Low Vacuum, and the High Rear. tor Pressure parameters which are not required by the STS. The only periodic test of the Low Reactor Vessel Water Level is a channel functional test whereas the STS requires a periodic channel check and channel calibration in addition to the channel functional test. The test 14

TABLE 2.

COMPARIS0N OF LA CROSSE BOILING WATER REACTOR ENGINEERED SAFETY FEATURES (ESF) INSTRUMENT SURVEILLANCE REQUIREMENTS WITil BWR STANDARD TECHNICAL SPECIFICATION (STS) REQUIREMENTS.

CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST IS REQUIRED STS LA CROSSE STS LA CROSSE STS LA CROSSE STS LA CROSSE 1.

PRIMARY CONTAINMENT ISOLATION a.

Reactor Vessel Water Level Q

R 1, 2, 3 (1) Low, Level 3 0

M 1,2,3 Q

(2) Low Low, Level 2 0

M 1,2,3 Q

M b.

Drywell Pressure--High NA c.

Main Steam Line (1) Radiation--High D

W(a)

R 1, 2, 3 1

Q (2) Pressure--Low NA M

1,2,3 M

Q (3) Flow--High D

1,2,3 R

d.

Main Steam Line Tunnel NA M

Temperature--liigh 1, 2#, 3#

Q M

e.

Condenser Vacuum--Low NA 1, 2, 3 Q

M f.

Main Steam Line Tunnel NA A Temperature--High g.

Containment Radiation Monitor--High

TABLE 2.

(continued)

CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST IS REQUIRED STS LA CROSSE STS LA CROSSE STS LA CROSSE STS LA CROSSE h.

Low Steam Line Pressure and Low Vacuum 1.

Reactor Pressure High 2.

SECONDARY CONTAINMENT ISOLATION a.

Plant Exhaust Plenum D

M(a)

R 1, 2, 3, 5

(

Radiation--High and

• c.

b.

Drywell Pressure--High NA M

Q 1,2,3 c.

Reactor Vessel Water D

M Q

1, 2, 3 Level--Low, Level 3 d.

Refueling Floor D

M(a)

Q 1,2,3,5 Exhaust Radiation--

and

• High 3.

REACTOR WATER CLEANUP WSTEM ISOLATION a.

a Flow--High D

M R

1,2,3 b.

Area Temperature--High NA M

R 1, 2, 3 c.

SLCS Initiation NA R

NA 1, 2, 3

TABLE 2.

(continued)

CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CAllBRATION TEST IS REQUIRED STS LA CROSSE STS LA CROSSE STS LA CROSSE STS LA CROSSE 1, 2, 3 d.

Area Ventilation a NA M

R Temperature--High 1, 2, 3 e.

Reactor Vessel Water D

M Q

Level--Low Low, Level 2 4.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION Q

1,2,3 O

a.

RCIC Steam Line NA M

Flow--High 1,2,3 b.

RCIC Steam Supply NA M

Q Pressure--Low 1,2,3 c.

RCIC Turbine Exhaust NA M

Q Diaphragm Pressure--

High R

1, 2, 3 d.

RCIC Equipment Room NA M

Temperature--High e.

RCIC Steam Line Tunnel NA M

1, 2, 3 R

Temperature--High f.

RCIC Steam Line Tunnel NA 1, 2, 3 R

M a Temperature--High

n TABLE 2 (continued)

CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST IS REQUIRED STS LA CROSSE STS LA CROSSE STS LA CROSSE STS LA CROSSE 5.

RHR STEAM CONDENSING SYSTEM ISOLATION a.

RHR Equipment Area a NA M

1, 2, 3 R

Temperature--High b.

RHR Area Cooler NA M

1,2,3 R

Temperature--High c.

RHR Return Line Flow--

NA M

R 1,2,3 liigh en 6.

SHUTDOWN COOLING SYSTEM ISOLAT10N a.

Reactor Vessel Water D

M Q

3, 4, 5 Level--Low Low, Level 2 h.

RHR Pump Suction NA M

Q 1,2,3 Pressure--High c.

Drywell Pressure--High NA M

Q 1,2,3 d.

RHR Pump Suction NA M

Q 1, 2, 3 Flow--High

TABLE 2.

(continued)

CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST IS REQUIRED STS LA CROSSE STS LA CROSSE STS LA CROSSE STS LA CROSSE 7.

LOW PRESSURE CORE SPRAY SYSTEM a.

Reactor Vessel Water D

M Q

SA 1,2,3,4, Level--Low Low Low, 5*

Level I b.

Drywell Pressure--High NA M

1,2,3 Q

c.

Injection Valve D

1, 2, 3, 4 M

Q g

Differential Pressure 5*

--High d.

Pump Discharge Flow-Low NA M

Q 1,2,3,4, (Bypass) 5*

e.

Reactor Vessel NA Q

SA 1, 2, 3, 4 M

Pressure--Low 5*

f.

Reactor Building SA Pressure--High 8.

LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM a.

Reactor Vessel Water D

M Q

1,2,3,4 Level--Low Low Low, 5*

Level 1

TABLE 2.

(continued)

CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CAllBRATION TEST IS REQUIRED STS LA CROSSE STS LA CROSSE STS LA CROSSE STS LA CROSSE 1, 2, 3 b.

Drywell Pressure NA M

Q

--High 1, 2, 3, 4, c.

Injection Valve 0

M Q

Differential Pressure 5*

--Low 1, 2, 3, 4 d.

RHR Pump Start Time NA SA R

Delay Relay 5*

Q 1, 2, 3, 4, m

e.

Pump Discharge Flow--

NA M

Low (Bypass) 5*

9.

HIGH PRESSURE CORE SPRAY SYSTEM Q

SA 1, 2, 3 a.

Reactor Vessel Water D

M Level--Low Low, Level 2 1, 2, 3 b.

Drywell Pressure--High NA Q

M 1, 2, 3 c.

Condensate Storage NA M

Q Tank Level--Low d.

Suppression Chamber D

Q 1,2,3 M

Water Level--High

TABLE 2.

(continued)

CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CAllBRATION TEST IS REQUIRED STS LA CROSSE STS LA CROSSE STS LA CROSSE STS LA CROSSE Q

1,2,3 M

e.

Reactor Vessel Water NA Level--High Q

1, 2, 3 f.

Pump Discharge NA M

Pressure--High 1,2,3 g.

Pump Suction Pressure NA M

Q

--Low h.

HPCS System Flow NA M

Q 1, 2, 3 S

Rate--High SA 1.

Containment Building Internal Pressure--

High

10. AUTOMATIC DEPRESSURIZA-TION SYSTEM Q

1, 2, 3 a.

Reactor Vessel Water D

M Level--Low Low Low, Level I Q

1, 2, 3 b.

Drywell Pressure--

NA M

High 1,2,3 R

NA c.

ADS Timer NA

=-

TABLE 2.

(continued)

CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST IS REQUIRED STS LA CROSSE STS LA CROSSE STS LA CROSSE STS LA CROSSE d.

Low Pressure Core NA M

1,2,3 Q

Spray Pump Discharge Pressure--High e.

RHR LPCI MODE Pump NA M

Q 1,2,3 Discharge Pressure--

High f.

Reactor Vessel Water D

1, 2, 3 M

Q Level--Low Low, level 2 mm

11. SHUTDOWN CONDENSER SYSTEM a.

Reactor Building A

Isolation Valve Not Fully Open b.

Turbine Building A

Isolation Valve Not Fully Open c.

Reactor Pressure A

Above 1325 psig.

TABLE 1.

NOTATION S

- At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

n-TABLE 2.

(continued)

TABLE 1.

NOTATION (continued)

D

- At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> W

- At least once per 7 days At least once per 31 days M

Q

- At least once per 92 days At least once per refueling outage (18 months)

R Not applicable N.A.

At least once per 6 months SA g;

S/U

- Prior to start up A

- Annually Not performed or available function 1.

POWER OPERATION 2.

STARTUP 3.

HOT SHUTDOWN 4.

COLD SHUTDOWN 5.

REFUELING

1. When reactor steam pressure > (

) psig and/or any turbine stop valve is open.

i

TABLE 2.

(continued)

TABLE 1.

NOTATION (continued)

When handling irradiated f uel in the secondary containment.

a.

Instrument alignment using a standard current source.

b.

Not applicable when the system is not required to be OPERABLE.

I frequency for the channel functional test also deviates from that k

required by the STS.

The High Containment Building Internal Pressure is periodically subjected to a channel functional test but not subjected to a channel calibration or channel check. The High Containment Radiation Monitor, the Low Steam Line Pressure and Low Vacuta and the High Reactor Pressure parameters are not subjected to

)

periodic channel checks, channel calibrations, and channel functional testing.

It is left to the NRC Staff to determine whether operating information can justify testing adequacy of the Low Reactor Vessel Water Level and High Containment Building Internal Pressure parameters utilizing channel functional testing only and whether there is enough plant operating experience to justify no testing for the High Containment Radiation Monitor, the Low Steam Line Pressure and Low Vacuum, and the High Reactor Pressure parameters.

2.

The Low Pressure Coolant Injection System controls and remotely operated valves are tested seni-annually to demonstrate their operability and an integrated system test is performed annually.

The Low Pressure Coolant Injection System utilizes the Low Reactor Vessel Water Level and the Low Reactor Vessel Pressure parameters which correspond to that requirec by the STS and utilizes the High Reactor Building Pressure parameter which is not required by the STS. The only periodic test of the Low Reactor Vessel Water Level l

and the Low Reactor Vessel Pressure parameters is a channel func-tional test, whereas the STS require a periodic channel check and channel calibration in addition to the channel functional test.

The test frequency for the channel functional test also deviates from that required by the STS.

k r

25 I,

D

\\

I J

s g

The High Reactor Building Pressure parameter is periodically

}

subjected to a channel functional test but is not subjected to a channel check or channel calibration.

It is left to the NRC Staff to determine whether there is enough plant operating experience to determine testing adequacy for the Low Pressure Coolant Injection System using only periodic channel functional testing.

7 3.

The High Pressure Core Spray System controls and remotely-operated valves are tested semi-annually to demonstrate their operability and an integrated system test is performed annually. The High Pressure Core Spray System utilizes the Low Reactor Vessel Water Level parameter which corresponds to that required by the STS and utilizes the High Containment Building Internal Pressure para-meter which is not required by the STS. The only periodic test-ing of the Low Reactor Vessel Water Level is a channel functional test whereas the STS require a periodic channel check and channel calibration in addition to the channel functional test. The test frequency for the channel functional test also deviates from that required by the STS.

The High Containment Building Internal Pressure parameter is periodically subjected to a channel functional test but is not subjected to a channel check or channel calibration.

It is left to the NRC Staff to determine whether there is enough plant operating experience to determine testing adequacy for the High Pressure Core Spray System using only periodic channel i

l functional testing.

i 4.

The Shutdown Condenser System control valves are tested quarterly to demonstrate their operability and an integrated system test is

(

performed annually. The Shutdown Condenser System is not l}

required by the STS. The system is not subjected to periodic channel checks or channel calibration.

)

26

l It is left to the NRC Staff to determine whether there is enough plant operating experience to determine testing adequacy for the

(

Shutdown Condenser System using only annual integrated system testing.

5.

There are no requirements established in the la Crosse Technical Specifications determining the modes for which surveillance is required.

6.

There are no specific requirements establ'ished in the La Crosse j

Technical Specifications for response times for those systems comprising the ESF.

s 5.0

SUMMARY

The review of the reference material has determined the present testing and testability of the La Crosse RPS and ESF do not meet the criteria of Section 2.0 of this Technical Evaluation.

The Technical Specifications do not establish that testing per the specified criteria of Section 2.0 would adversely affect the safety or the operability of the unit, nor has the licensee established that the probability of system f ailure is acceptably low without regular testing during reactor operation. The licensee has also not established why the Technical Specifications do not require chan-nel calibration or system time response testing for the RPS and ESF.

There l

is no effort made to determine whether the La Crosse RPS and ESF is adequate for reactor operating purposes. Reference to the Standard Technical Speci-l fications for General Electric Boiling Water Reactors was made to give a general comparison.

l

6.0 REFERENCES

1.

General Design Criterion 21, " Protection System Reliability and Test-ability," of Appendix A, " General Design Criteria for Nuclear Power l

Plants," 10 CFR part 50, " Domestic Licensing of Production and Utili-zation Facilities."

a 2.

Regulatory Guide 1.22, " Periodic Testing of the Protection System Actuation Functions."

i 27 3

3.

IEEE Standard 338-1975, " Periodic Testing of Nuclear Power Generating Station Class lE Power and Protection Systems."

4.

General Design Criterion 40, " Testing of Containment Heat Removal Systems," of Appendix A, " General Design Criteria for Nuclear Power Plants," 10 CFR Part 50, " Domestic Licensing of Production and Utili-zation Facilities."

5.

Nuclear Regulatory Commission Standard Review Plan, Section 7.1, Appen-dix B, " Guidance for Evaluation of Conformance to IEEE STD 279."

6.

Appendix A to Provisional Operating Authorization No. DPRA-6, " Tech-nical Specifications for the La Crosse Boiling Water Reactor (LAC 8WR),"

Amendment 26, 1981.

l 7.

f0 REG-0123, Rev. I " Standard Technical Specifications for General Electric Boiling Water Reactors."

I i

28 m