ML20041B741

SEP Topic VI-10.A,Testing of Reactor Protective Sys & Esfs, Yankee Nuclear Power Station

ML20041B741
Person / Time
Site:	Yankee Rowe
Issue date:	12/31/1981
From:	Vanderbeek R EG&G, INC.
To:	NRC
Shared Package
ML20041B740	List: ML20041B739 ML20041B741
References
TASK-06-10.A, TASK-6-10.A, TASK-RR 0548J, 548J, NUDOCS 8202250148
Download: ML20041B741 (27)

Text

..

~

e-0548J SYSTEMATIC EVALUATION PROGRAM TOPIC VI-10.A TESTING OF REACTOR PROTECTIVE SYSTEM AND ENGINEERED SAFETY FEATURES YANKEE NUCLEAR POWER STATION Occket No. 50-29 December-1981 R. VanderBeek EG M Idaho, Inc.

i 12/30/81 8202250148 820201 PDR ADOCK 05000029 P

PDR

j a

\\

o CONTENTS 1

1.0 INTRODUCTION

1 i

i 2.0 CR I TER I A........................................................

1 3.0 R EA CTOR PROTE CTIVE SYSTEM.......................................

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

4 3.2 E v al u a ti o n................................................

S 4.0 ENGIN EERED SAFETY FEATUR ES SYSTEM................................

12 4.1 Description...............................................

12 4.2 Evaluation.................................................

14 5.0

SUMMARY

14 6.0 R EFER EN C ES......................................................

22 TABLES 1.

Comparison of Yankee Nuclear Power Station RPS instrument surveillance requirements with PWR Standard Technical Specification requirements......................................

6 2.

Comparison of Yankee Nuclear Power Station Engineered Saftey Features (ESF) instrument surveillance requirements with PWR Standard Technical Specification Requirements..........

15 ii w -

.~ym.

y.,s

SYSTEMATIC EVALUATION PROGRAM TOPIC VI - 10.A TESTING OF REACTOR PROTECTIVE SYSTEM AND y M NEERED SAFETY FEATURE 5 YANKEE NUCLEAR POWER STATION

/

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 frequercy 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 three 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 0.1.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

--m

.-,v-3.-

a.

There is no practicable system design that would permit operation of the actuated equi;...;ent without acversely affecting the safety or operability of the plant, b.

The probability that the protective system will fail to initiate the operation of the actuated equipmt:nt 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 lE Power and Protectich 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 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 2

of the system and showing that the sum of the response times of all is within the limits of the overa'l system requirement.

In addition, the following 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 38, " 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 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.5 3

m---

m

,smw,.,._

_,,_._-m._

m.__-. - - _

o

___ _=

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

It must reliably effect a rapid shutdown of the rea.: tor 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 Yankee-0 Rowe Technical Specifications are as follows:

No. of PARAMETER CHANNELS TRIP LOGIC Manual Reactor Trip 3

1 out of 3 Power Range and Intermediate Power Range Nuclear Instrumentation 6

2 out of 4-Intermediate High Start-up Rate 2

1 out of 2a Nuclear Instrumentation Source Range Nuclear Instrumentation

~

a.

Startup 2

NA b.

Shutdown 2

NA b

low Main Coolant Flow (Steam 4

2 out of 3 GeneratoraP)

Main Coolant Pump Overcurrent/

Undercurrent D

System A 4

2 out of 3 b System B 4

2 out of 3 High Main Coolant System Pressure 3

2 out of 3 Low Main Coolant Pressure 3

2 out of 3 High Pressurizer Water Level 1

1 out of 1 Low Steam Generator Water Level 4

2 out of 3 Turbine Trip 1

1 out of 1 4

~ - ~ +

v ye9'

"+ff+44++*a L-*

4

--.sv.eyeetg g af +g

-yv---@

w+)-M-g 1g-w g+mp.M N*my*

f MMTM fWM -' e W S De TIEi+ M a.

N

-e a

m*hN**

a..

No. of P AP.AMETER CHANNELS TRIP LOGIC Generator Trip 1

1 out of 1 Reactor Trip Breaker 2

1 out of 2 Automatic Trip Logic (SCRAM 2

1 out of 2 Amplifiers)

Main Steam Isolation Trip Logic 2

1 out of 2 a.

Below 15 mWe.

b.

Above 15 mWe.

3.2 Evaluation. Table 1 provides a comparision between the require-ments for surveillance as established by the PWR Standard Tecnnical Speci-7 fications and those set forth by the Yankee Nuclear Power Station Technical Specifications.

The evaluation of the technical specifications indicates that:

1.

The Yankee Rowe RPS utilizes eight Main Coolant Pump Current channels, four Low Main Coolant Flow Channels, a Generator Trio Channel and two Main Steamline Isolation Trip Logic Channels which are not specified by the standard technical specifications.

2.

The following Reactor Protection System Parameters specified in the standard technical specification are not used in the Yankee Rowe RPS.

a.

Power Range Neutron Flux, High Positive Rate, b.

Power Range Neutron Flux, High Negative Rate, c.

Overtemperature AT, 5

J

.n..,

,..-.~,r.n..

i

[

IADtE 1.

COMPARISON OF TANKEE NUCLE AR POWER STATION RPS INSTRtMENT SURVEILLANCE REQUIREMENIS Willi PWR SIANDARD TECimICAL SPECIFICATION REQUlillMINIS (SIS)

}

CilANNEL MODES FOR WillCH CilANNEL CllANNEL IUNCTIONAL SURVE ILL ANCC CllECK CALIBRATION lEST 15 REGilRED i

YANKEE VANKEE YAfEEE YANKEE SIS R0WE STS R0WE' STS R0WE SIS R0WE Manual Reactor Trip N.A.

N.A.

N.A.

ft. A.

S/U(l)

S/U(l)

I, 2, and a 1, 2, and

• 0(2)l

• {l2land M

M 1, 2 1 2 D(2 q(6),M( 3)

Power Range, Neutron flux 5

S Q(6

),

Intermediate Power Range, S

D(2),

M 1, 2, and Neutron flux Q(6)

• (12)

Power Range, Neutron flux, N. A. ' --

R(6)

N

), 2 liigh Positive Rate

), 2 Power Range, Neutron flux, M.A.

R(6)

M liigh Negative Rate intermediate Range, 5

5 R(6)

R(6)

S/U(l)

M

), 2, and

• l(13),i, j

and

• Neutron Flux SourceRange,Neutronflux 5(7)

S R(6)

R(6)

M and S/U())

2, 3, 1, 2(,3,1, S/U(l)

S, and *

5. and 1

f 6

L

• (16) f Overtevnperature AT S

R M

1, 2 i

l 11 p

Overpower AT S

R M

1, 2 l

1 t

e l

l P

L 4

i-il' TABLE 1 (continued)

CHA1NEL MODE 5 [0R WHICH i

I.

CitANNEL CHANNEL fuMCil0NAL SLEVE ILL ANCE CllECK CAllBRATION TEST 15 REQUIRED YANKEE YANKEE YA*EE YANKEE SIS R0WE STS R0WE STS ROWE SIS R0WE l-F Pressurizer Pressure--Low 5

S R

R(10)

H H

1, 2 1,2(15) j Pressurizer Pressure--illgh 5

S R

lt(10)

M M

1, 2 1,2(IS)

Pressurlier Water Level--Hfgh 5

S R

R(10)

M H(ll)

I, 2 1,2(15)

Loss of Flow - Single Loop S

R M

]

4-Loss of Flow - Two Loop 5

lt N.A.

1 l

Low Main Coolant flow S

R(10)

M(ll) 1(14)

(Steam Generator Diff Pressure)

Low Main Coolant flow S

it H

1(14)

Systers A and B (MC Pump Overcurrent/

y Undercurrent)

Steam Generator Water Level--

S S

R R(10)

Pl M

l. 2 l(14) i Low-Low i

f Steam /Feedwater flow Hismatch 5

R M

), 2 and Low Steam Generator Water level l

i I

o 1

4 r

i 4

TABLE I.

(continued)

CHANNEL N0 DES fbR WHICH g.

EHANNEL CHANNEL

[UNCTIONAL SURVE llL ANCE CHECK CALIBRAll0N TEST IS REQUIRED 4

YANKEE YANKEE YANKEE JANKEE 515 R0WE STS R0WE SIS R0WE SIS RDWE i:

l'

+

I undervoltage - Reactor M.A.

R N

I Coolant Pumps Underfrequency - Reactor N.A.

R M

1 Coolant Pumps Turbine Trip t-A.

Low Fluid 0l1 Pressure N.A.

N.A.

5/U())

]

B.

Turbine Stop and Control N.A.

M.A.

N.A.

N.A.

S/U(l)

S/U(l) l l(14.17)

Valve Closure i

Safety injection [pput N.A.

N.A.

M(4) 1, 2 from ESF R

]

i Reactor Coolant Pump Breaker N.A.

N.A.

a, l

Position Trip 1.

l Reactor irlp System Interlocks l.

?

5/U(8) 2, and

• 8 A.

Intermediate Range M.A.

R(9)

Neutron flux, P-$

I

+

B.

Low Power Reactor N.A.

R(9) 5/U(8) 1 I

Trips Block, P-7 e

i i

i, j.

1 1

i EE t

I

t i

l

'l TABLE I.

(continued)

~

CHANNEL MODES f 0R WillCil j

CHANNEL EHANNEL

[UNCTIONAL SLRVEILLANCE C11ECK CALIBRAll0N 1EST IS REQUIRED YANKEE JANKEE YANKEE YANKEE

,i)

SIS R0WE STS R0WE STS R0WE STS ROWE C.

Power Range Neutron N.A.

R(9) 5/U(8) 1 flux, P-8 D.

Power Range Neutron N.A.

R(9) 5/U(8) 1, 2 flus, P-10

.)

s E.

Turbine Impulse Chamber M.A.

R(9) 5/U(8) l Pressure,P-13 i

Reactor irlp Breaker N.A.

N.A.

N.A.

N.A.

M(5)

S/U(l)

), 2, and *

), 2, and

• l and S/U(l)

Automatic Trip logic N.A.

N.A.

N.A.

N.A.

M(5)

S/U(l)

J, 2, and *

], 2, and

• Generator Trip N.A.

M.A.

5/U(l) l( 1,18)

Hafn Steam Isolation irlp N.A.

N.A.

Q 1,2(15) l e

l l

T 1:

e i

H L--l 4

.... [

\\

i 1

?

t IAlstE I.

(continued) f

?

i TABLE l--NOTAil0N

--- Not performed or avajjable function.

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

i liigh voltage to detector is automatically de-energized above 5 x l0-9 Amperes on the Intermediate Range.

(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) lleat balance only, above 15% of RATEP THERMAL POWER. Adjust channe) If absolute difference greater than 2 percent.

(3) Compare incore to excore antal flux dif ference above 15% of SAKO ilfRM*1 POWER. Recalibrate (( the absolute difference greater than or equal to (2) percent.

(4) Manual ESF functiona) loput check every la months.

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

i (6) Neutron detectors may be excluJed from CllANNEL CAllpRATION.

i g

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

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

(9) The total interlock function shall 1;e denunstrated OPERABLE durlp9 HANNEL cal!8 RATION testing of each channel C

affected by interlock operation.

r (10) known pressure gpplied to sensor.

(11) When shutdown longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, if not performed in the previous 34 days.

A e

s e

1

.f 1

TABLE I.

(continued)

TABLE l--NOTATION (continued)

(12) Power Range, Neutron flux, tow Setpolot Trip may be manually bypassed at a 15Ne. Bypass sha)) be manually removed at s 15 MWe.

I (13) Intermediate Range, Neutron flux, liigh Startup Aate Trip is automatically bypassed a 15 NWe, Bypass is i

l automatically removed at s 15 MJe.

(14) Irlp may be manually bypassed s 15 MWe. Bypass is automatically removed at e 15 MWe, (15) Trip may be manually bypassed wiven the reactor is not critical.

(16) Startup rate alarm setpoint s l.) decade / minute.

(11) Turbine shall be protected by at least the following protective trips: rotor excessive axial movement, low bearing oil pressure, icw condenser vacuum, and overspeed.

(18) Generator shall be protected by at least the following protectlye trips: overturrent, differential, ar.) loss of field.

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)

I i

D

- At least once per N.A. - Not applicable g

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 5A

- At least once per 184 days DW

- At least once per 14 days S/U -

Prior to start up M

- At least once per M - Alternate channels tested I

31 days on a stagged basis at least once per 62 days.

Q

- At least once per 3 sonths t

i s

b l

4 w--

x-..

=.

d.

Overpower AT, e.

Loss of flow-single and two loop, f.

Steam /feedwater flow mismatch, g.

Undervoltage-Reactor Coolant pumps, h.

Underfrequency-Reactor Coolant pumps, i.

Safety Injection Input from ESF, j.

Reactor Coolant Pump Breaker Position Trip and k.

Reactor Trip System Interlocks; 3.

The channels of the Yankee Rowe RPS are checked, tested or cali-brated; and, 4.

The test frequency and/or surveillance made for the Yankee Rowe, Power Range--Neutron Flux, Intermediate Range--Neutron Flux, Source Range--Neutron Flux, Reactor Trip Breaker, and Automatic Trip Logic do not correspond to those required by the standard technical specifications. Refer to Table 1.

Although not specified in the Yankee Rowe technical specifications, monthly operational tests are performed on the nuclear instrumentation sensors and other RPS systems by use of operational procedures.

4.0 ENGINEERED SAFETY FEATURES SYSTEM 4.1 Descriotion.

The Engineered Safety Features System consists of the Safety Injection Actuation System (SIS), the Containment Isolation Actuation System (CIS), and the Main Steam Isolation System (MSI).

12

- = -

. = _ =. -..

Th: Engineered Safety Features (ESF) Systen parameters and logic 6

channels as identified in the Yankee Rowe Technical Specification are as follows.

Total No.

Channels Of Channels And Sensors Functional Unit And Sensors To Trio 1.

Safety Injection a.

Actuation Channel #1 1

1

1) RPS Low Main Coolant Pressure Channel

2) High Containment Pressure Sensor 1

1

3) Manual Initiation 1

1 b.

Actuation Channel #2 1

1

1) Low Main Coolant Pressure Sensor

2) High-Containment Pressure Sensor 1

1

3) Manual Initiation i

1 2.

Containment Isolation a.

Manual Initiation 2

1 b.

Actuation Channel A 1

1 i

l

1) High Containment Pressure Sensor 1

1

2) Safety Infection (All Safety Injection (non-essential valves)

Initiating Functions and Requirements) c.

Actuation Channel B 1

1

1) High Containment Pressure Sensor 1

1 l

2) Safety Injection (All safety injection (non-essential valves) initiating functions and requirements) 13

.=

..a - - -. -. -

Total No.

Channels Of Channels And Sensors Functional Unit And Sensors To Trio 3.

MAIN STEAM ISOLATION c.

Low Steam Line Pressure 3/ Steam Line 2/Stea.n Line b.

Automatic. Trip Logic 2

1 c.

Manual Initiation 2

1 d.

High Containment Pressure Trip Containment Isolation 2

1 4.2 Ev aluation. Table 2 provides a comparison between the PWR Stan-dard Technical Specifications requirements and those of the Yankee Nuclear Power Station Technical Specifications for the rurveillance of the Engi-neered Safety Features (ESF) System.

Evaluation of Table 2 shows that a channel functional test of the

' Safety Injection System, Containment Isolation System, and Main Steam Isola-tion System is not performed at the recommended testing frequency of once per 31 days, but at an 18 month interval. However, although not specified in the Yankee Rowe technical specification, a monthly operational test is performed on sensors and valves of the safety injection isolation actuation system and the containment isolation acutation system by use of operational procedures.9 5.0

SUMMARY

The Technical Specifications for Yankee Nuclear Power Station were compared with the PWR Standard Technical Specifications used for current reactor licensing.

It was found that (a) five of the Reactor Protection System and the ESF system instrumentation channels are not tested at the same testing interval required by the Standard Technical Specifications (STS) (See Section 3.2 and 4.2), (b) several instrumentation channels specified by the STS for the Reactor Protection System (RPS) and ESF System 14

)

W l

TABLE 2.

COMPARISON OF YANKEE NUCLE AR POWER SIAll0N ENGINEERED SAf fly [EAIURES (ESF) INSIRUMENT SURVEitt ANCE REQUIREMENIS WIIH PWR STANDARD IECllNICAL SPEClflCATIONS (STS) REQUIREMENIS.

CilANNEL CilANNEL IUNCil0g MODES FOR WillCil CitANNEL CllECK CAL IBRATION TEST SURVE IL L ANCE IS REQUIRED Vank ee Yankee Yankee Yankee SIS Rowe SIS Rowe SIS Rowe SIS Rowe i

SAFEIY INJECTION, IURBINE TRIP AND FEEDWATER ISDL ATION 5, als), 4.

I, 2 3 4

a.

Manual Inttjation N.A.

N.A.

N.A.

N./.

M(i)

R

(, 2. 3, 1 i

b.

Automatic Actuation Logic N.A.

5 N.A.

N.A.

M(2)

M(6) 1, 2, 3, 4 1,2,3,f(10,II) t c.

Containment Pressure -liigli S

S R

R(8)

M(3)

M(3) 1, 2, 3 1,2,3,f(10,ll) j d.

Pressurf rer Pressure--t og S

R H

1, 2, 3 e.

Differential Pressure Between 5

R M

1, 2, 3 l

Stemn Lines--liigh i

f.

Steam Flow in Two Steam S

R M

1, 2, 3 Lines--High Colncident wjth i

f i

-Lcw-Low or Steam l

Lin$ Pressure--Low f

g.

RPS Low Main Coolant S

R(8)

M(1)

),2,3,f(lo,II)

Pressure Channe' h.

Low Main Coolant Pressure S

R(8)

If(P) 1,2,3,f(10)

Sensor

}

4

r IABLE 2.

(continued) l CilANNEL CilANNEL [UNCil0N H0 DES FOR WilC!I CilANNEL EllEEK CALIBRATION TEST SURVE ILL ANCE IS REQUIRED Yankee Yankee Yankee Yankee l

SIS Rowe SIS Rowe 515 Rowe SIS Rowe i

t I

l CONTAINMIqi SPRAY I

a.

Manual Initiation N.A.

M.A.

M(1)

), 2, 3, 4 t

l b.

Automatic Actuation Logic N.A.

N.A.

N(2)

), 2, 3, 4

'{

c.

Containment Pressure--

S 8

N(3)

), 2, 3 High-High CONTAINHENT ISOLAi!0N t

i a.

Phase "A" laolat jon j

j a

I) Manual N.A.

M.A.

M(1) 1, 2. 3, 1

2) from Safety injection M.A.

N.A.

M(2)

), 2, 3,1 i

g Automat {c Actuation [ogic I

b.

Phase "B" Isolation

1) Hanual N.A.

M.A.

M(1)

},.2,3,4

), 2, 3, 4 M(2)

2) Automatic Actuatjon Logic N.A.

q.A.

3) Containment Pressure--

S R

N(3) 9, 2, 3 1

High--liigh l

t I

e

IABLE 2.

(continued) j (llANNEL CilAMNEL fUNCJ10N MODES FOR WillCil filANNEL CitECK CAllBRAll0N TEST SLRVE llt.ANEE IS REQUIRED l

Yankee Yankee Yar*. e Yankee i

SIS Rowe STS Rowe SIS Rowe SIS Rwe c.

Purge an4 Exhaust Isolation l} nanual N.A.

M.A.

N())

), 2, 3,1

, ~

2) Automatic Actuation tejtc q.A.

M.A.

N(2)

), 2, 3, 4

3) Containment Radio-S R

M 1, 2, 3,1 Act ly lty--Illgli i

i d.

Manual Initiation M.A.

N.A.

R 1, 2

,4, i

5.

• 9 e.

Actuation Channel A S

N.A.

M(li) 1, 2, 3, 4, i

5,*(9) t O

1) liigh Containment S

lt(8)

N(8)

), 2, 2, 4.

Pressure Sensor 5,*(9)

2) Safety Injection (15)

(15)

( l'.)

(15) f.

Actuation Chanael 8 S

4.A.

N(li) 1, 2, 3, 4 5,*(9)

1) liigh Containment S

R(8)

M(8) 1, 2, 3, 4 Pressure Sensor 5,*(9)

2) Safety injection (15)

(15)

(15)

(15) i i

4 I

e

i t.

TABLE 2.

(continued)

~

CilANNEL CilANNEL FUNC{l0N 000ES FOR WillCli CilANNEL CalECK CAllBRAil0N TEST Stsytlit#NCE 15 flEQUIRED Vank ee Yankee Yankee Yankee 515 Rowe SIS Rowe

,_ SIS Rowe 515 Rowe t

SIEAM LINE ISOLAI)0N a.

Manual N.A.

M.A.

M.A.

N.A.

M(l)

R 1, 2, 3 1, 2 b.

Automatic Actuattor) fogic M.A.

N.A.

N.A.

N.A.

N(7) y 1, 2, 3 1, 2(12) c.

Containment Pressure-S N.A.

R N.A.

M(1)

F.

l. 2, 3 1, 2 lii gh--liigh l

d.

Steam flow in Two Steam S

S R

R(8)

N N(3)

), 2, 3 1, 2 1ines--liigh Colncident wlth T

-Low-Low or Steam Lfn$ Pressure--Low IURBINE TRIP AND FEEDWA?tR ISOLAil04 g

i a.

Steam Generator Water 5

R M

1, 2, 3 Level--High--liigh 4

AUXILIARY TEEDWATER a.

Manual N.A.

N.A.

N(1) 1, 2, 3 b.

Automatic Actuation Logic M.A.

N.A.

M(2) 1, 2, 3 c.

Steam Generator Water 5

R K

12,.3 L evel--L ow-low l

i l

f L

a i

i 2

9

?

TABLE 2.

(continued) i CilANNEL CilAN.1EL FUNCTION MODES FOR WHICll CilANNEL CliECK CAllBRATION IEST SURVE l: LtaCE IS REQUIRED Vankee Yankee Yankee Yankee f.

STS Rowe SIS Rowe SIS Rowe

$15 Rowe d.

Undervoltage - RCP S

8 M

1, 2 e.

Safety tr.jection (see above) f.

Station Blackout N.A.

R N.A.

), 2, 3 9 Trip of Main feedwater funps M.A.

N.A.

R

), 2 AUIDMATIC SWITCH 0VER 10 CONTAINMENT SUMP l

R M

], 2, 3, 4 ll a.

RSWI level - Low 5

ColNCIDENT Wilif R

M

). 2, 3, 4 Containment Sump Level - iligh 5

AND Safety injection (see above)

LOSS OF POWER a.

4.16 kV Emergency Bus (13) 5 R

M M

I, 2, 3. 1 I, 2, 3, 4 l'

undervoltage (Loss of Voltage) b.

4.16 kV Emergency Bus (13) 5 R

N N

1, ?, 3,1 I, 2, 3, 4 Undervoltage (Degraded Voltage) l l.

I-1.

J L.

TABit 2.

(continued) i CilANNEL CHANNEL [UNCil0N NODES FOR WillCH CllANNEL CilECK CALIBRATION 1EST SLAVE llt ANCE IS REQUIRED Yenkee Yankee Yankee Yankee SIS Rowe SIS Rowe 515 Rowe SIS Rowe l

ENGINEERED SWETY FEATURE N.A.

R(5)

M(4) 1, 2. 3 1

ACTUATION SYSTEM INTERLOCKS TABLE l--NOTATION (1) Manual actuation switches shall be tested at least once per 18 months during shutdown. Ali other circuitry associated with r.anual safeguards actuation shall recelye a CHANNEL FUNCil0NAL TEST at least once per 31 days.

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

1 (3) The CHANNEL TUNCTIONAL TEST shall include exercising the $ransmitter by applying either a yacuum or pressure to the appropriate. side of $he $ransmitter.

i (4) Logic for the interlocks shall be demonstrated OPERABLE durlng the automatic actuation logic test of each (bannel affected by interlock operation.

j 1

(5) Ibe total Interlock function shall be demonstrated OPERABLE during CilANNEL CaLIBRAlloit testing of each channel af fected by 8

Ir.terlock operatlon.

I (6) Wt.en shutdown with mala coolant pressure < 1000 psig, if not performed wl thin the previous 31 days.

( 7) When shutdown longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, jf not performe4 in the previous 31 days, i

(8) lhe test shall include exercising $he sensor by applying elther a yacuum or pressure to the appropriate side of the sensor.

(9) Trip function may be bypassed in this NODE with main coolant pressure < 300 psig.

(10) Trip function may be bypassed in this NODE with malp coolant pressure < 1800 psig.

i i

Il o

s.

i gJ i

~

l 1

TABLE 2.

(continued)

I J

t 1ABLE l--N0iAll0N (continued)

I (11) Automatic Initiation of Actuation Channel #1 may be bypassed ln thls MODE during functional test of the Main Coolant Systes pressure channel.

(12) Irlp may be manually bypassed when the reactor is not critical.

i (13) Varaee Rowe does not have 4.16 ky toss of power trip but does have 480 V Emergency Bus irlps.

(14) When in C0tD SINIDOWN with main coolant pressure < 300 psig, if not performed within the preylous 31 days.

(15) A}l Safety injection Surveillance Requirements

-- Not performed function or available.

Not required in this NODE with main coolant pressure < 300 PSIG.

l Not required in this MUDE with main coolant pressure < ]100 PSIG.

1 S

At least'once per it 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-(18sonths) ca D

6-At least once per lj.A. -

Nut Applicable j

74 hours8.564815e-4 days <br />0.0206 hours <br />1.223545e-4 weeks <br />2.8157e-5 months <br /> i

SA -

At least once per 181 days Prior lo start up l

IN -

At least once per S/U 14 days l

M At least once per AM -

Alternate channels tested 31 days on a stagged basis a$ least once per 62 days Q

At least once per 3 sos.ths i

N 1!'

t

1. a i

1

are not provided for in the Yankee Rowe RPS and ESF; however there are many insti ument channels provided in the Yankee Rowe RPS not specified in the STS, and (c) there are no time response tests required for the RPS or ESF instrumentation channels.

With the exception that no time reposnse testing is performed, the Yankee Rowe testing and surveillance practices for the-RPS and ESF meet the current reactor licensing criteria. However, as noted in the summary of the RPS and ESF, the testing frequencies are not identical to those of the Standard Technical Specifications.

It is left to the NRC Staff to deter-mine whether the noted testing frequency deviations are acceptable.

6.0 REFERENCES

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 1E 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.

Yankee Nuclear Power Station Technical Specifications, Appendix "A" to License No. OPR-3, April 2,1981.

I j

7.

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

t 8.

OP-4601--Nuclear Instrumentation Channel Functional Test Rey. 10.

OP-4659--Main Coolant System Pressure Channels Functional Test Rev.1.

22

.m

s 9.

OP-4615--SI Accumulator Level Switch Operational Check and Check of functions performed by level switches, Rev. 3.

OP-4617--Containment Isolation Actuation Channel and Sensors Functional Test, Rev. 6.

OP-4622--Functional Test of the Accumulator Nitrogen Pressure Regulating Valves (SI-PR-58, 59, 602), pev. 2.

OP-4634--3afety Injection Actuation High Containment Pressure Sensors (SI-PS-238 and SI-PS-239) Functional Test, Rev. 9.

OP-4638--Functional Test of PS-SI-14 Low Main Coolv t Pressure Safety Injection Actuation, Rev. 9.

OP-4656--Functional Test of the NRV Main Steam Line Pressure Channels, Rev. 1.

OP-6456--Safety Injection Actuation Safety Valve Setpoint Check, Rev. 6.

i l

l l

23

-,m

+ -.

.;,e wp

ENCLOSURE SYSTEMATIC EVALUATION PROGRAM TOPIC VI-10. A YANKEE R0WE TOPIC: VI-10. A. Testing of Reactor Trip System and Engineered Safety Features, Including Response Time Testing I.

INTRODUCTION The purpose of this Topic is to review the reactor trip system (RTS) and engineered safety features (ESF) test program for verification of RTS and ESF operability on a periodic basis and to verify RTS and ESF response time in order to assure the operability of the RTS and ESF.

Response times should not exceed those assumed in the plant accident analyses. Accordingly, the test program of the RTS and ESF was reviewed in accordance with the Standard Review Plan, including applicable Branch Technical Positions.

II.

REVIEW CRITERIA The review criteria are presented in Section 2 of EG&G Report 0548J,

" Testing of Reactor Trip System and Engineered Safety Features".

III.

RELATED SAFETY TOPICS AND INTERFACES Topic VI-7.A.3 discusses the question of testing protection systems under conditions as close to design condition as practical. There-are no topics that are dependent on the present topic information for their completion.

IV.

REVIEW GUIDELINES Review guidelines are presented in Sections 3 and 4 of Report 0548J.

V.

EVALUATION Yankee Rowe complies with the current licensing criteria, except for IEEE Std 338-1975, Section 6.3.4 (response time testing).

VI.

CONCLUSION It is the staff's position that the design of systems which are re-quired for safety shall include provisions for perfodic verification that the minimum performance of instruments and control is not less than that which was assumed in the safety analyses. The bases for

- - = - - - - - - - - - - - = - - - - - - - - - - -

this position are General Design Criterion 21, Section 3,9 of IEEE Std. 279-1971, and IEEE Std. 338-1 971. Therefore, the licensee should im-plement a program for response time testing of all reactor protection systems (including engineered safety features systems such as.contain-mentisolation). As a part of this program, the response time test requirements should be stated in the Technical Specifications in a manner similar to that of the Standard Technical Specifications.