ML20058D779

Draft Safety Evaluation, SEP Topic VI-10.A,Testing of RCS & Esf

ML20058D779
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	03/04/1982
From:	Vanderbeek R EG&G, INC.
To:	NRC
Shared Package
ML18086B747	List: ML18086B745 ML20058D779
References
TASK-06-10.A, TASK-6-10.A, TASK-RR 0043J, 43J, NUDOCS 8207270357
Download: ML20058D779 (23)

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SYSTEMATIC EVALUATION PROGRAM 2

TOPIC VI-10.A TESTING OF REACTOR PROTECTIVE SYSTEM AND ENGINEERED SAFETY FEATURES HADDAM NECK Docket No. 50-213 March 1982 R. VanderBeek EG&G Idaho, Inc.

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CONTENTS 1.0 I N TR O D U CT I O N....................................................

1 2.0 CR I T ER I A........................................................

I 3.0 R EA CTOR PR OTE CT I VE S YS TEM.......................................

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

4

3. 2 Evaluation................................................

5 4.0 ENGIN EER ED S AFETY FE ATUR ES S YS TEM...............................

11 4.1 De s cr i p t i o n...............................................

11 4.2 Evaluation.................................................

11 5.0

SUMMARY

11 6.0 R EF E R EN C E S......................................................

22 TABLES 1.

Comparison of Haddam Neck RPS instrument surveillance requirements with PWR Standard Technical Specification r eq u i r eme n t s....................................................

6 2.

Comparison of Haddam Neck Engineered Saftey Features (ESF) instrument surveillance requirements with PWR Standard Techn ical Speci fi cation Requir ements............................

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4 SYSTEMATIC EVALUATION PROGRAM TOPIC VI - 10. A TESTING OF REACTOR PROTECTIVE SYSTEM AND ENGINEERED 5AFETY FEATURE 5 HADDAM NECK

1.0 INTRODUCTION

The objective of this review is to determine if all reactor protective system (RPS) components, including punps 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:

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a.

There is no practicable system design that would permit operation of the 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 equipment during reactor operation, and c.

The actuated equipment can be routinely tested when the reactor is shut down.

IEEE Standard 338-1977, " Periodic Testing of Nuclear Power Generating Station Class 1E 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.

and, in part, in Section 6.3.4:

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 overall response times given in the Safety Analysis Report.

Sufficient overlap shall be provided to verify overall system response.

The response-time shall include as nuch 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 2

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

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 practical, the performance of the full operational sequence that brings 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.

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 Evalua-tion of Conformance to IEEE STO 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 pennit 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.5 3

4 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 Haddam 6

Neck Technical Specifications are:

Logic Required for No. of Full Power Parameter Channels Operation Manual Trip 2

1 out of 2 Nuclear Overpower Reactor Trip 4

2 out of 3 Pressurizer Variable Low Pressure 3

1 out of 2 Reactor Trip Pressurizer Fixed High Pressure 3

1 out of 2 Reactor Trip Low Coolant Flow Reactor Trip 4 Loop Operation 4

1 out of 4 3 Loop Operation 3

1 out of 3 Pressurizer low Pressure Signal 3

1 out of 2 (For Safety Injection Trip)

Steam-Feedwater Flow Mismatch 4

1 out of 4 Coincident with low Steam Generator Level-Reactor Trip High Steam Flow Isolation Valve Trip 4 loop operation 4

2 out of 4 3 loop operation 3

2 out of 3 4

4 3.2 Evaluation. Table 1 provides a comparision between the require-ments for surveillance as established by the PWR Standard Technical Speci-fications and those set forth by the Haddam Neck Technical Specifications.

The following items resulted from the review and evaluation of the testing and surveillance of the Haddam Neck RPS.

1.

The Haddam Neck RPS is not tested for automatic or manual actuation during reactor operation.

2.

Section 4.1 of the Haddam Neck Technical Specifications provides the surveillance requirements for the RPS. This section states that surveillance is determined by the operating organizations and responsible safety review personnel, and therefore is not specified in the technical specifications. This fact prevents a comparison or a means of determining adequate surveillance for the RPS.

3.

Of the Haddam Neck RPS, the Manual trip, the Pressurizer Low Pressure trip, and the High Steam Flow Isolation Valve trip circuitry are not included for testing, calibration, and channel checks. These parameters do not appear in Table 4.2-1 of Section 4.2 of the technical specifications.

4.

The testing, calibration, and channel check requirements set forth in Table 4.2-1, Section 4.2, of the technical specifica-tions include testing and channel checks for Intermediate Range Neutron Flux and Reactor Coolant Temperature. These-two para-meters are not identified as part of the Haddam Neck RPS but they are identified as RPS in the Standard Technical Specifications.

5.

The Haddam Neck RPS utilizes a Low Reactor Coolant Flow trip which is not a required STS parameter for the RPS.

6.

The technical specifications do not require channel functional testing for the Steam-Feedwater Flow Mismatch and Low Steam Generator Water Level.

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TABLE 1.

COMPARISON OF HADDAM NECK RPS INSTRUMENT SURVEILLANCE REQUIREMENTS WITH PWR STANDARD TECHNICAL SPECIFICATION REQUIREMENTS (STS)

CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CALIBRATION TEST IS REQUIRED HADDAM HADDAM HADDAM HADDAM STS NECK STS NECK-STS NECK STS NECK Manual Reactor Trip N.A.

N.A.

S/U(1) 1, 2, and

• Power Range, Neutron Flux S

S D(2),

D M

BW 1, 2 Q(6),M(3)

Power Range, Neutron Flux, N.A.

R(6)

M 1, 2 High Positive Rate Power Range, Neutron Flux, N.A.

R(6)

M 1, 2 High Negative Rate m

S/U(1)

S/U(1) 1, 2, and

• Intermediate Range, S

S R(6)

Neutron Flux M and S/U(1) 2,3,4 Source Range, Neutron Flux S(7)

S(7)

R(6)

S/U(1) 5, and

• Overtemperature AT S

S R

R M

1, 2 Overpower AT S

M 1, 2 R

1, 2 Pressurizer Pressure--Low S

R M

Pressurizer Pressure--High S

S R

R M

W 1, 2 Pressurizer Water Level--High S

S R

R M

W 1, 2 Low Reactor Coolant Flow S

R SS

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TABLE 1.

(continued)

CHANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CHECK CAllBRATION TEST IS REQUIRED HADDAM HADDAM liADDAM HADDAM STS NECK STS NECK STS NECK STS NECK Reactor Trip System Interlocks S/U(8) 2, and

• R(9)

A.

Intermediate Range N.A.

Neutron Flux, P-6 S/U(8) 1 R(9)

B.

Low Power Reactor N.A.

Trips Block, P-7 S/U(8) 1 R(9)

C.

Power Range Neutron N.A.

Flux, P-8 y

S/U(8) 1, 2 R(9)

D.

Power Range Neutron N.A.

Flux, P-10 S/U(8) 1 E.

Turbine Impulse Chamber N.A.

R(9)

Pressure, P-13 M(5) 1, 2, and

• i Reactor Trip Breaker N.A.

N.A.

and S/U(1) a M(5) 1, 2, and

• N.A.

Automatic Trip Logic N.A.

TABLE 1.

(continued)

CilANNEL MODES FOR WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE CilECK CALIBRATION TEST IS REQUIRED llADDAM HADDAM llADDAM HADDAM STS NECK STS NECK STS NECK STS NECK 1

A.

Loss of Flow - Single Loop S

R M

1 R

N.A.

B.

Loss of Flow - Two Loop S

1, 2 Steam Generator Water Level--

S R

M Low-Low Steam /Feedwater Flow Mismatch S

S R

R M

1, 2 and Low Steam Generator Water Level 1

M Undervoltage - Reactor N.A.

R Coolant Pumps M

1 Underfrequency - Reactor N.A.

R Coolant Pumps Turbine Trip S/U(1) 1 A.

Low Fluid Oil Pressure N.A.

N.A.

S/U(1) 1 N.A.

B.

Turbine Stop Valve Closure N.A.

M(4) 1, 2 Safety injection input N.A.

N.A.

from ESF 1

R N.A.

Reactor Coolant Pump Breaker N.A.

Position Trip

TABLE 1.

(continued) i TABLE l--NOTATION Not performed or available function.

With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

(1) Each startup or when required with the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal, if not performed in previous 7 days.

(2) Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel if absolute difference greater than 2 percent.

(3) Compare incore to excore axial flux difference above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference greater than or equal to (2) percent.

(4) Manual ESF functional input check every 18 months.

u)

(5) Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(6) Neutron detectors may be excluded from CHANNEL CAllBRATION.

(7) Below P-6 (Block of Source Range Reactor Trip) setpoint.

(8) Logic only, each startup or when required with the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal if not performed in previous 92 days.

(9) The total interlock function shall be demonstrated OPERABLE during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

TABLE 1 (continued)

TABLE l--NOTATION (continued)

S At least once per R

At least once per refueling 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> outage (18 months) 0 At least once per N.A. -

Not applicable 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SA At least once per 184 days BW At least once per 14 days S/U -

Prior to start up M

At least once per AM Alternate channels tested 31 days on a stagged basis at least once per 62 days.

Q At least once per W

Every 6 weeks.

3 months

_,o SS Each scheduled shutdown if not tested or cali-brated in preceding 6 months.

7.

The frequency for the channel functional test for the Power Range Neutron Flux, High Pressurizer Pressure, an'd High Pressurizer Water level does not correspond with that established in the STS.

8.

The Haddam Neck Technical Specifications do not establish Time Response Testing for the RPS.

The present surveillance and testing requirements for the Haddam Neck RPS as established in the Haddam Neck Technical Specifications do not meet the reactor licensing criteria set forth in Section 2 of this document.

4.0 ENGINEERED SAFETY FEATURES SYSTEM 4.1 Description. The Engineered Safety Features System consists of the Containment Isolation System, the Containment Spray System, and the High Pressure Safety Injection System.

4.2 Evaluation. Table 2 provides a comparison between the require-ments for testing and surveillance as established by the PWR Standard Tech-nical Specifications and those set forth by the Haddam Neck Technical Specifications. The Haddam Neck Technical Specifications do not establish any testing and surveillance requirements for the ESF. The present Haddam Neck testing and surveillance requirements do not meet the present Licens-ing Criteria of Section 2 of this document.

5.0

SUMMARY

The review of the reference material has determined that the present testing and surveillance requirements established by the Haddam Neck Tech-nical Specifications for the RPS and ESF do not meet the criteria of Sec-tion 2.0 of this Technical Evaluation.

An apparent basis for not meeting the present criteria of Section 2.0 is set forth in the basis of Section 4.2 of the technical specifications which states:

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TABLE 2.

COMPARISON OF HADDAM NECK ENGINEERED SAFETY FEATURES (ESF) INSTRUMENT SURVEILLANCE REQUIREMENTS WITH PWR STANDARD TECHNICAL SPECIFICATIONS (STS) REQUIREMENTS.

CHANNEL CHANNEL FUNCTION MODES FOR WHICH CHANNEL CHECK CALIBRATION TEST SURVEILLANCE IS REQUIRED HADDAM HADDAM HADDAM HADDAM STS NECK STS NECK STS NECK STS NECK SAFETY INJECTION, TURBINE TRIP AND FEEDWATER ISOLATION a.

Manual Initiation N.A.

N.A.

M(1) 1, 2. 3, 4 b.

Automatic Actuation Logic N.A.

N.A.

M(2) 1, 2, 3, 4 c.

Containment Pressure--High 5

R M(3) 1, 2, 3 E

d.

Pressurizer Pressure--Low S

R M

1, 2, 3 e.

Differential Pressure Between 5

R M

1, 2, 3 Steam Lines--High f.

Steam Flow in Two Steam 5

R M

1, 2, 3 Lines--High Coincident with T

-Low-Low or Steam Lin$ Pressure--Low CONTAINNENT SPRAY a.

Manual Initiation M.A.

N.A.

M(1) 1, 2, 3, 4 b.

Automatic Actuation Logic M.A.

N.A.

M(2) 1, 2, 3, 4 c.

Containment Pressure--

S R

M(3) 1, 2, 3 High-High

TABLE 2.

(continued)

CHANNEL CHANNEL FUNCTION HODES FOR WilCH CHANNEL CHECK CAllBRATION TEST SURVEILLANCE IS REQUIRED HADDAM HADOAM HADDAM HADDAM STS NECK STS NECK STS NECK STS NECK CONTAINMENT ISOLATION a.

Phase "A" Isolation I) Manual N.A.

M(1)

N.A.

1, 2, 3, 4

2) From Safety Injection M.A.

N.A.

M(2) 1, 2, 3, 4 Automatic Actuation Logic b.

Phase "B" Isolation w

1) Manual N.A.

N.A.

M(1) 1, 2, 3, 4

2) Automatic Actuation Logic N.A.

M(2)

N.A.

1, 2, 3, 4

3) Containment Pressure--

S R

M(3) 1, 2, 3 High--High c.

Purge and Exhaust isolation

1) Manuel N.A.

N.A.

M(1) 1, 2, 3, 4

2) Automatic Actuation Logic N.A.

N.A.

M(2) 1, 2, 3, 4

3) Containment Radio-S R

M 1, 2, 3, 4 Activity--High i

TABLE 2.

(continued)

CHANNEL CHANNEL FUNCTION MDDES FOR WHICH CHANNEL CHECK CAllBRATIDN TEST SURVEltLANCE l$ REQUIRED HADDAM HADDAM HADDAM HADDAM STS NECK STS NECK STS NECK STS NECK STEAM LINE ISOLAil0N a.

Manual N.A.

N.A.

M(1) 1, 2, 3 M(2) 1, 2, 3 b.

Automatic Actuation Logic N.A.

N.A.

c.

Containment Pressure-S R

M(3) 1, 2,-3 High--liigh g

d.

Steam Flow in Two Steam S

R M

1,2,3 Lines--High Coincident with T

-Low-Low or Steam LlNPressure--Low TUR8INE TRIP AND FEEDWATER ISDLATIDN 1, 2, 3 a.

Steam Generator Water 5

R M

Level--High--High AUXILIARY FEEDWATER N.A.

M(1) a.

Manual N.A.

1, 2, 3 b.

Automatic Actuation Logic M.A.

N.A.

M(2) 1, 2, 3 c.

Steam Generator Water 5

R M

1, 2, 3 Level--Low-Low

g TABLE 2.

(continued)

CHANNEL CHANNEL FUNCTION MODES FOR WHICH CHANNEL CHECK CAL I BR AT IDN TEST SURVE ll L ANCE 15 REQUIRED HADDAM HADDAM HADOAM HADOAM SIS NECK SIS NECK STS NECK STS NECK d.

Undervoltage - RCP S

R M

1, 2 Safety injection (see above) e.

f.

Station Blackout N.A.

R N.A.

1, 2, 3 g.

Trip of Main Feedwater Pumps N.A.

N.A.

R 1, 2 AUTOMATIC SWITCHOVER TO m

CONTAINMENT SUMP La R

M 1, 2, 3, 4 a.

RSWT Level - low 5

Colncident with 1, 2, 3, 4 Containment Sump Level - High 5

R M

and Safety injection (see above)

LOSS OF POWER M

1, 2, 3, 4 a.

4.16 kV Emergency Bus 5

R Undervoltage (Loss of Voltage)

M 1, 2, 3, 4 b.

4.16 kV Emergency Bus 5

R Undervoltage (Degraded Voltage)

ENGINEERED SAFETY FEATURE N.A.

R(5)

M(4) 1, 2, 3 ACTUATION SYSTEM INTERLOCKS s

TABLE 2.

(continued)

TA8tE l--NOTATION (1) Manual actuation switches shall be tested at least once per 18 months during shutdown. All other circuitry associated with manual safeguards actuation shall receive a CHANNEL FUNCTIONAL TEST at least once per 31 days.

(2) Each train or logic channel shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(3) The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter.

(4) Logic for the interlocks shall be demonstrated OPERABLE during the automatic actustion logic test of each channel af fected by interlock operation.

(5) The total interlock function shall be demonstrated OPERABLE during CHANgEL CALIBRATION testing of each channel affected by interlock operation.

5 At least once per R

At least once per refueling 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> outage (18 months)

D At least once per N.A. -

Not Applicable 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SA -

At least once per 184 days BW -

At least once per 14 days S/U -

Prior to start up M

At least once per AM -

Alternate channels tested 31 days on a stagged basis at least once per 62 days 0

At least once per 3 months

" Periodic testing of instrumentation channels when no deficiency is apparent is not considered desirable. The risk of inadvertently trip-ping the plant is increased whenever a channel is being tested. The risk that a channel will be left inoperable after a test, while small, is nevertheless also increased. Furthermore, while the channel is on test, it is not available to perform its design function, be it con-trol or protection. Experience with process instrumentation in industry, thermal plants and at the Yankee-Rowe power reactor supports these statements."

It is left to the NRC staff to determine whether enough plant operat-ing information is available to support non-conformance to the criteria of Section 2.0.

The evaluation of the RPS also resulted in conflict between available documented information on the Haddam Neck RPS. Comparison of items A through F readily indicate the differences as to what parameters constitute the Haddam Neck RPS.

A--Dwg. EOSK 312790-F Rev. 7 (1) High Flux Trip 2 out of 4 (2) Safety Injection Trip (3) High Pressure Trip 2 of 3 (4) Low Pressure Trip 2 of 3 (5) Steam Isolation Valve Trip (6) Turbine Trips (7) Loss of Reactor Coolant Flow Trip (8) High Startup Rate Trip (9) High Pressurizer Water Level Trip 2 of 3 (10) Steam /Feedwater Flow Mismatch Trip (11) Manual Trips (12) High Steam Flow Trip 2 of 4 i

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B--Dwg. EDSK 313551 Shts 33-38 Rev. 11 6/27/73 (1) Pressurizer High Pressure (sheet 34)

(2) Steam Line Break (3) High Startup Rate (4) High Power Level (5) Safety Injection (6) Turbine Trip (7) Steam Line Isolation Valve Position (8) Pressurizer Low Pressure (9) Pressurizer High Pressure (sheet 36)

(10) Steam Line Break (sheet 36)

(11) High Startup Rate (sheet 36)

(12) High Power Level (sheet 36)

(13) Safety Injection (sheet 36)

(14) Steam-Feedwater Flow Mismatch (15) Turbine Trip (sheet 37)

(16) Steam Line Isolation Valve Position (sheet 37)

(17) Pressurizer Low Pressure (sheet 37)

(18) Loss of Flow C--Docket 50213-26, " Supporting Information for Full Term Operating License Application (Conn Yankee Atomic Power Co. Haddam)," Dec.1969 (1) High Pressurizer Pressure Reactor Trip 2 of 3 (2) High Pressurizer Level Reactor Trip 2 of 3 (3) Variable Low Pressure Trip 2 of 3 (4) High Startup Rate Reactor Trip 1 of 2 (5) High Neutron Flux Reactor Trip 2 of 4 18

(6) Low Reactor Coolant Flow Trip 2 of 4 (7) Reactor Trip from Turbine Trip 2 of 3 (8) Steam Flow-Feedwater flow Mismatch (9) Steam Line Excess Flow Reactor Trip 2 of 4 D- "Haddam Neck Plant Facility Description and Safety Analysis (FSAR),"

Volume 1 & 2, Section 7.2-3, October 1970 (1) High Startup Rate Reactor Trip (2) High Flux or Overpower Reactor Trip (3) Variable Low Pressure Reactor Trip (4) High Pressurizer Pressure Reactor Trip (5) High Pressurizer Level Reactor Trip (6) Low Reactor Coolant Flow Protection (7) Reactor Trip from Turbine Trip (8) Steam-Feedwater Flow Mismatch (9) Steam Line Excessive Flow Reactor Trip i

(10) Reactor Trip from Isolation Valves E- " Technical Report: Reliability Study of the Connecticut-Yankee Reactor Protection System," by W. G. Jordan and F. T. Eggleston under Shop Order HRP-45773 (1) Pressurizer High Pressure 2 of 3 (2) Steam Line Break i of 2 (3) Nuclear Trips 2 of 4 and 1 of 2 (4) Safety Injection 1 of 2 (5) Steam / Feed Mismatch & Low Gen. Level 1 of 4 (6) Pressurizer High level 2 of 3 (7) RCP Bus Undervoltage 1 of 2 i

19

4 (8) Pressurizer Low Pressure 2 of 3 (9) Steam line Isolation Position 1 of 4 (10) Turbine Stop Valve Position 1 of 1 (11) Turbine Auto Stop Oil Pressure 2 of 3 (12) RCS Low Flow 2 of 4 (13) RCP Breaker Position 2 of 4 (14) Manual Reactor Trip 1 of 2 F--Appen' dix A to Facility Operating License DPR-61, " Technical Specifica-tions for the Connecticut Yankee Atomic Power Company, Haddam Neck Plant, Haddam, Connecticut," February 19, 1982, Section 2.4, Maximum Safety Set-tings Protective Instrumentation.

(1) Pressurizer Pre:sure High (2) Pressurizer Level High (3)

Variable Low Pressure (4)

Nuclear Overpower (5) Low Coolant Flow (6) Reactor Coolant Loop Valve--Temperature Interlock (7) High Steam Flow Table 3.9-1--Minimum Instrumentation Operating Conditions of Appendix A, At Power February 19, 1982 (1) Nuclear Overpower Reactor Trip 2 of 3 (2) Pressurizer Variable Low Pressure Reactor Trip 1 of 2 (3) Pressurizer Fixed High Pressure Reactor Trip 1 of 2 (4) Pressurizer High Water Level Reactor Trip 1 of 2 (5) Low Coolant Flow Reactor Trip--4 loop operation 1 of 4

--3 loop operation 1 of 3 20

A (6)

Pressurizer Low Pressure Signal (For Safety Injection Trip) o (7) Pressurizer Low Water Level Signal (For Safety Injection Trip) 1 of 2 (8) Manual Trip 1 of 1 (9)

Steam-Feedwater Flow Mismatch Coincident with Low Steam Generator Level-Reactor Trip 1 of 4 (10) High Steam Flow Isolation Valve Trip--4 loop operation 2 of 4

--3 loop operation 2 of 3 It is left to the NRC Staff to determine whether the conflict in information--specifically plant drawings--needs further evaluation.

6.0 REFER ENCES 1.

General Design Criterion 21, " Protection System Reliablility and Test-ability," of Appendix A, " General Design Criteria for Nuclear Power Plants," 10 CFR part 50, " Domestic Licensing of Production and 'Utili-zation Facilities."

2.

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

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 Facility Operating License DPR-61, " Technical Specifica-tions for the Connecticut Yankee Atomic Power Company Haddam Neck Plant, Haddam, Connecticut." Anmended as of 2-19-1982.

7.

Standard Technical Specifications for Westinghouse Pressurized Water Reactors, NUREG-0452, Revision 3, September,1980.

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_