ML20099C028

Rev 1 to Topical Rept GENE-770-06-1, Bases for Changes to Surveillance Test Intervals & Allowed Out-Of-Svc Times for Selected Instrumentation TS, Technical Evaluation Rept

ML20099C028
Person / Time
Site:	Hope Creek
Issue date:	07/21/1992
From:	Lasher D EG&G IDAHO, INC.
To:	NRC
Shared Package
ML20099C018	List: ML20099C028
References
EGG-RTS-10142, EGG-RTS-10142-R01, EGG-RTS-10142-R1, NUDOCS 9208030256
Download: ML20099C028 (109)

Text

.

6 EGG-RTS-10142 Revision 1 TECHNICAL EVALUATION REPORT REVIEW 0F TOPICAL REPORT GENE-770-06-1

" BASES FOR CHANGES TO SURVEILLANCE TEST INTERVALS AND ALLOWED OUT-OF-SERVICE TIMES FOR SELECTED INSTRUMENTATION TECHNICAL SPECIFICATIONS" Don R. Lasher Published April 1992 Idaho National Engineering Laboratory EG&G Idaho, Inc.

Idaho. Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Comission Washington, D.C.

20555 Under DOE Contract No. DE-AC07-761001570 FIN No. 06045 Ob54 PDR ADO K O P

PDR

.o

6 ABSTRACT This EG&G Idaho, Inc. report provides a review of the subject topical report and subsequent submittals from the Boiling Water Reactors Owner's Group rhich present reliability analyses as the bases for increasing the surveillance test intervals and allowed out-of-service times for testing and repair for certain instrumentation systems included in the BWR4 and BWR6 Technical Specifications.

FOREWORO This report is supplied as part of the review work. performed for the Technical Specification Improvement Program under which the BWROG and G.E. supplied a series of topical reports which provide reliability centered bases for making changes to the Surveillance Test Intervals and Allowed Out-0f-Service Times for certain instrumentation systems included in the BWR4 and BWR6 Technical Specifications. This work is being done for the U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Operational Events Assessment by EG&G Idaho, Inc., NRR&T - Washington Technical Office.

The U.S. Nuclear Regulatory Commission funded this work under the authorization B&R No. 220-19-15-04-0, FIN No. D6045.

ii i

4

CONTENTS ABSTRACT...........................................................

11 FOREWORD..........................................................

11 I.-

INTRODUCTION.................................................

1 II.

PROP) SED CHANGES TO TECHNICAL SPECIFICATIONS.................

3 III.

ACCEPTANCE CRITERIA..........................................

5 IV.

EVALUATION OF BASES FOR STI AND A0T CHANGES..................

'7 Section 3.I - BWR4 Plant Systems Actuation Instrumentation..............................................

7 Section 3.2 - BWR4 and BWR6 E0C-RPT Actuation Instrumentation..............................................

8 Section 3.3 - BWR4 and BWR6 ATWS-RPT Actuation Instrumentation..............................................

8 9ection 3.4 - BWR4 and BWR6 RCIC System Actuation Instrumentation..............................................

9 Section 3.5 - BWR4 Safety / Relief Valves (S/RV) and Safety / Relief Valves Low Low Set (LLS) Function Instrumentation...............................................

10 Section 3.6 - BWR6 Safety / Relief Valves (S/RV) and Safety / Relief Valves low Low Set (LLS) Function Instrumentation.............................................

11 Section 3.7 - BWR6 Plant Systems Actuation Instrumentation.............................................

12 Section 3.8 - BWR4 Main Control Room Environmental Control-System (MCRECS) Instrumentati on.....................

15 Section 3.9-- BWR6 Control Room Fresh Air (CRFA) i

-Actuation Instrumentation...................................

15 Section 3.10 - BWR4 and BWR6 Control Rod Block i

Instrumentation............................................

16 i

e l

V.

REFERENCES..................................................

19 l

L ENCLOSURE'I................................................

1-1 l

iii l-L i

.s...

y REVIEW OF TOPICAL REPORT GENE-770-06-1

" BASES FOR CHANGES TO SURVEILLANCE TEST INTERVALS AND ALLOWED OUT-0F SERVICE TIMES FOR SELECTED INSTRUMENTATION TECHNICAL SPECIFICATIONS" FEBRUARY 1991 i

I. INTRODUCTION This topical. report presents analyses intended to support changes in surveillance test intervals (STIs) and allowed out--of-service times (A0Ts) for actuation instrumentation for selected systems for BWR 3,4,5,6 (Relay) and BWR 6 (Solid State) Plants. These systems include:

For the BWR4:

Plant Systems Actuation Instrumentation, Main Control j

-Room Environmental Control System (MCRECS), Safety / Relief Valves, and Safety / Relief Valves Low Low Set (LLS) Function.

For the BWR6:

Plant Systems Actuation Instrumentation, Control Room Fresh Air Actuation Instrumentation (CRFA), Safety / Relief Valves, and Safety / Relief Valves-Low Low Set (LLS) Function.

For the BWR4 and BWR6:

E0C-RPT System Actuation Instrumentatior., ATWS-RPT System Actuation Instrumentation, RCIC System Actuation Instrumentation and Control Rod Block Instrumentation.

The changes to STIs and A0Ts. proposed in this report are for actuation instrumentation for systems not covered in previous analyses of RPS, ECCS, and Containment Systems by the BWROG and are intended to be consistent with similar previously reviewed and accepted changes to tie RPS, ECCS, and Containment Systern actuation instrumentation.

The procedure adopted by the BWPOG in this report differs from that used in previous Technical Specification Improvement reports in that the systems were not directly modelled, but instead attempts were made to show similarity in components and function with previously reviewed and accepted actuation instrumentation. 'This procedure was applied to all the identified actuation instrumentation, however, the NRC staff required further, more detailed analysis be performed for the RCIC actuation instrumentation, therefore the RCIC analysis in the current report will not be evaluated or reported on here.

This additional analysis is presented in a separate topical report (GENE-770-06-2) and its review reported on in a separate TER.

t L

I

8J.. +

4 a-54 A-4-

- A 4-4.,.

e 4

'e Y

I-l

( '

l.

l l

l t

i n

l'.

I i

I t

l I

I l

l l

3 1V

-=

+

REVIEW 0F TOPICAL REPORT GENE-770-0G-1

" BASES FOR CHANGES TO SURVEILLANCE TEST INTERVALS AND ALLOWED OUT-OF-SERVICE TIMES FOR SELECTED INSTRUMENTATION TECHNICAL SPECIFICATIONS" FEBRUARY 1991

1. INTRODUCTION This topical report presents analyses intended to support changes in surveillance test intervals (STIs) and allowed out-of-service times (A0Ts) for actuation instrumentation for selected systems for BWR 3,4,5,6 (Relay) and BWR 6 (Solid State) Plants. -These systems include:

For the BWR4:

Plant Systems Actuation Instrumentation, Main Control Room Environmental Control System (MCRECS), Safety / Relief Valves, and Safety / Relief Valves low Low Set (LLS) Function.

For the BWR6:

Plant Systems Actuation Instrumentation, Control Room Fresh Air Actuation Instrumentation (CRFA), Safety / Relief Valves, and Safety / Relief Valves low Low Set (LLS) Function.

For the BWR4 and BWR6: EOC-RPT System Actuation Instrumentation, ATWS-RPT System Actuation Instrumentation, RCIC System Actuation Instrumentation and Control Rod Block Instrumentation.

The changes to STIs and A0Ts proposed in this report are for actuation instrumentation for systems not covered in previous analyses of RPS, ECCS, and Containment Systems by the BWROG and are intended to be consistent with similar previously reviewed and accepted changes to the RPS, ECCS, and Containment System actuation instrumentation.

The procedure adopted by the BWROG in this report differs from that used in previous Technical Specification Improvement reports in that the systems were not directly modelled, but instead attempts were made to show similarity in components and function with previously reviewed and accepted actuation instrumentation. - This procedure was applied to all the identified actuation instrumentation, however, the NRC staff required further, more detailed analysis be perfacmed for the RCIC actuation instrumentation, therefore the RCIC ans'ysis in the current report will not be evaluated or reported on here.

This additional analysis is presented in a separate topical report (GENE-770-06-2) and its review reported on in a separate TER.

1

~

j e

i 4

I i

i i

i i

I J

I B

)

F t

5 k

i A

P 9

1 I

.g 1

i l -

'n 2

i l

l~

.d+-

-cw-W w

g--wap c-w j-T serr- - - - = -+9s y,y__q n-w-g

=+-w7 y--

Te gg9-

=*<ym a

e-qrt1="'7. w q y-v e ev-ewe-*-w-gT*u--a-a-*"r-WMe e-e n-e-rs>p-1

' - *-N-rMPW-4'Y7*'M*d'4"'

y g-y

_. - _ _ _. _ - - ~.

.c 1

i

-II.

PROPOSED CHANGES T0 TECHNICAL SPECIFICATIONS The changes to-the technical. specifications for the identified actuation.

instrumentation proposed in this topical report include:

(a)

Change the Surveillance Test Interval (STI) from once every 31 days to once every 92 days.

(b)

Change the Ait:en Out-of-Service Time (A0T) for repair from 1 or 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 12 N rs for equipment common to the RPS or 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

' for all other equipment.

e (c)

Change the Allowed Out-of-Service Time (A0T) for test from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

These proposed changes are the same as those previously approved for the RPS, ECCS, and Containmcat Isolation actuation instrumentation.

f d

1

A

,.,m,

,a;J

-4~.kam A e a w a m.

ea

...A.m.

4 k a wa. Am s.

au4.1-J.-

4.s 4-_-a

. u-a s i~. u se.24 s a A.s.

_a.'ss.

=a.

a.Aar A+

a =

+.a.c,4,m6 ua+,v-..m.*2._

m k.

1 b

-j 9

b i

i-e

- I t

I J.

M h

I r-4

• ry-.

--,,,,,+

,,.- -.,.e--r-y

.w.--,

y,.s

-4

,-~ -.-.-,.-. - -. -.. -

III. ACCFPTANCE CRITERIA The general criteria for acceptance of the proposed changes to the actuatioti instrumentation tech;iical specifications differ in this topical report from the criteria in the previous topical reports in this series.

Instead of specific percentage limits on increases in system unavailability or f ailurt frequency or specific absolute limits on those increases as before, in this topical report, the new criteria are that if these changes are made to instrumentation of similar type to that used in the RPS, ECCS, or containment isolation instrumentation which has been previously analyzed, then this previous analysis can be used to justify the proposed changes.

This can be a reasonable procedure provided that equivalence in components, configuration, redundancy, and action performed is demonstrated in each case.

5

f J

Z e.44 4,$s6,-M 4

y,-w A

==N-<>me b-en-

'4 4AS,-

q---A-r -4 A

QA w

b -4 4 J

av+4-ud # w k & -L <=n Cav'n d,4 b

u-,,4=

~ \\

-An At

+A

+&

,$ ' f

' \\

s Y

d h

n 1

I h

4 6

h 8

i 9

9

- 5 6

d.

,I ~

.=...

_.;.m._,, _ _, _. _ _. _,. _.,

.._ ~._.

IV.

EVALDATION OF BASES FOR STI AND A0T CHANGES Topical Report GENE-770-06-1 presented the bases for STI and A0T changes in separate subsections of Section 3 for the various actuation systems.

Therefore the evaluations of.these bases are presented in this report as subsections which are numbered to correspond to the subsections in the topical report. presents a tabulation of the review results.

Section 3.1 - BWR4 Plant Systems Actuation Instrumertmt at iert 1

Changes requested include:

Change the Channel Functional Test frequency from monthly to quarterly, and add a note establishing an allowed out-of-service time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for surveillance testing.

These proposed changes to the STI and the A0T for testing are consistent with similar changes to STis and A0Ts previously made-and approved for other similar actuation instrumentation.

Evaluation of Basit For Change a)

The basis for including the feedwater System / Main Turbine Trip in the technical specifications is that this trip is applicable only to plants which do not have a direct reactor trip on reactor vessel water level 0 signals.

The trip is provided mainly as equipment protection against excessive moisture carryover into the main steam system and trips the reactor incidentally on main turbine trip. The trip has no safety significance because three other functions trip the reactor for this initiating event, thus the effect of losing this trip should be much less than the effect of losing the direct reactor vessel water level 8 reactor trip which was shown to be insignificant by the analysis in Ref. 1

~,

(Page 7-6).

We therefore agree that it is acceptable to justify the increase in STI and test A0T on the basis of this previous analysis.

b)

This topical report shows actuation instrumentation for the Suppression Pool (and Drywell) Spray System consisting of Drywell Pressure--High; Containment Pool Pressure--High; Reactor Vessel Water Level--Low Low Low, Level 1; and Timers, Systents A & B.

However, the basis states that the BWR4 design for this system is manually initiated and controlled and that there are no automatic functions or initiation instrumentation for this system.

This appeared contradictory and we requested clarification from the BWROG.

Their response confirmed that these systems are manually initiated and controlled for most BWR4 plants.

However, should a plant have automatic initiation and control, it will be similar to that for the BWR6 design and the bases for STI and A0T changes will be as those for the BWR6 given in Section 3.7.

The manual initiation and control statement in the basis and the response

-agrees with the Hatch 2 BWR4 plant design as an example. We conclude that these systems for most BWR4s are manually initiated 7

and controlled.

However, for those plants which have these systems automatically initiated and controlled, the bases presented in Section 3.7 of the topical report as modified by the response to Question 7 may be used to justify STI and A0T changes to these systems.

Section 3.2 - B_WR4 & BWR6 E0C. RPT Actuation Instrumentation Changes requested include:

Change the Channel Functional Test frequency from monthly to quarterly, change the allowed out-of-service time for surveillance testing from 2 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and change the allowed out-of-service time for repair from I hour to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. These proposed changes in STI and A0Ts are consistent with similar changes previously made and approved to STis and A0Ts for other similar actuation instrument 4 tion.

Evaluation of Basis For Change The current topical report states that the End of Cycle-Recirculation Pump Trip (E0C-RPT) is initiated by the "bine Stop Valve (TSV) Closure and Turbine Control Valve (TCV) Low Hy

.lic Pressure signals which are common to the PDS and that the STI and A0T uanges for these signals were analyzed in Ref. 1.

It further states that although the E0C-RPT trip functions were not explicitly identified in that analysis, the changes to the STI and A0Ts can be considered bounded by that analysis.

We noted that the logic arrangement is significantly different for these initiating signals (two-out-of-two TSV

. closure and two-out-of-two for TCV low oil pressure).

These two sets of signals are then "anded" with two channels of turbine first stage pressure >

4 30% RTP which act as a permissive.

It was not evident that the analysis of i

Ref. I covered this logic configuration nor was it evident that the Ref. I analysis results bounded this case. Additional information was provided by the BWROG which d. scribed the instrumentation differences between the RPS and

. DC-RPT functions together with calculations which compared the unavailabilities of the RPS and E0C-RPT functions. These comparisons indicated that the unavailabilities for the two functions were nearly the same,-thus we now conclude that the proposed STI and A0T changes for the EOC-RPT are acceptable en this basis.

Section 3.3 - BWR4 & BWR6 ATWS-RPT Actuation Instrumentation Changes requested include:

Change the Channel Functional Test frequency and the trip unit calibration frequency from monthly to quarterly; change the allowed out-of-service time for surycillance testing from 2 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; change I

the allowed out-of-service time for repair from I hour to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Evaluation of Basis For Change

- We did not agree on the ATWS-RPT logic for the BWR4 Improved Standard Technical Specifications (ISTS) Lead Plant (Hatch 2) which is given in the L

8 l

topical report as one-out-of-two per trip system with the tripping of both trip systems needed to trip both recirculation pumps.

Our logic diagram as obtained from the Hatch 2 UFSAR showed a two-out-of-two logic per trip system on-either high reactor pressure or low reactor vessel water level and trip of either trip system trips both pumps. After discussion with G.E.,

it appears that the newer BWR4 ATWS-RPT logic arrangement and that which is or will be installed on most BWR4 plants is the logic arrangement from the Hatch 2 UFSAR.

The Hatch I plant was stated to have been modified to this newer design at the last outage and the utility indicated that Hatch 2 (the lead ISTS plant) will be so modified during the upcoming outage.

This newer information should be made part of the topical report.

The current topical report stated that the effect of changes to STI and A0Ts for ATWS-RPT instrumentation on the reactivity shutdown failure frequency is negligible because the RPS failure frequency is low (-5.4E-06/yr. from Ref.1, page 5-29) and the change in overall ATWS-RPT function unavailability due to the STI and A0T changes (<1E-02/ demand) is small.

This indicated that even though the ATWS-RPT failure frequency may be higher, when it is combined with RTS, the combined effect is small. Additional information was provided by the BWROG which showed the calculation of the changes in ATWS-RPT channel unava11 abilities and the small effect of these changes on the reactivity shutdown failure frequency when the test interval is extended from one to three months.

Our review of this inform-tion confirmed the small effect of the ATWS-RPT STI and A0T changes on the reactivity shutdown failure frequency.

Thus we conclude that the basis for these changes is acceptable.

The last sentence on page 6 of the topical report states that the same small change in ATWS-RPT unavailability due to STI and A0T changes can be expected for other logic designs.

It was not clear how this could result since it would appear that the logic arrangement should influence the fault trees and their evaluation which means that a change in the logic configuration could result in a change in unavailability and hence failure frequency. Additional information provided by the BWROG indicated that the referenced statement was not intended to indicate that the reliability of a system is nearly independent of its logical arrangement, but that 'ogic designs having similar degrees of redundancy could be expet at to have similarly small changes in ATWS-RPT unavailability-This is a onable postulate and we conclude the statement as modified is acceptable.

Section 3.4 - BWR4 & BWR6 RCIC System Actuation Instrumentation The analysis of the RCIC actuation instrumentation and the results of changing STIs and A0Ts is more thoroughly presented in Ref. 7, therefore its review and evaluation is presented in a separate technical evaluation report.

l 9

l l

~..

Section 3.5 - BWR4 Safety / Relief Valves and Safetv/ Relief Valves LLS Function Instrumentation Changes requested include: Change the Channel Functional Test frequency from monthly to quarterly and change the allowed out-of-service time for surveillance testing from 2 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Evaluation of Basis For Change Relief Function:

for the BWR4, it is stated that there is no automatic lnitiation instrumentation associated with the relief function, therefore, there is nothing to review or report on.

LLS Function: We agree with the statt ant in the topical report that failure of the LLS function to initiate is not likely to have an immediate effect on the plant safety and therefore, that any small change in actuation instrumentation unavailability due to changes in the STI and A0Ts may be neglected.

However, the changes in the SRV LLS function actuation instrumentation unavailability caused by changes in the STI and A0Ts had not been shown to be small enough to neglect. The argument was made that the reasoning which supports makinn the requested changes was bounded by the analyses performed for the ECC5 and isolation actuation instrumentation and further that the analyses should apoly because the components used in the systems were the same or similar and the number of failures needed to disable the system were comparable. We found this argument, as presented, was not acceptable because it failed to consider the manner in which those components are assembled into the LLS logic.

For example, the LLS logic is an armed logic which is not similar to any of the ECCS or isolation actuation logic.

Some of the subsystems may-be similar but the manner in which they are organized into the logic is different, therefore the fault trees and their evaluation may differ and the results may not be bounded by the previous analyses.

In response to our Question Four, the BWROG submitted additional information consisting of fault trees and evaluations for the BWR4 LLS logic. In our review of this information, we agree that.four of the eleven S/RVs can be actuated by the LLS logic and that only one of the four needs to open to fulfill the LLS function. We did, however, question whether the channel logic used for the fault trees followed the channel logic as presented, for example, in the Hatch, Unit 2, BWR4 plant.

Our review showed two LLS logic channels in each Division: Channels A&C for Division I and channels B&D for Division II

.with each channel actuating one of the four LLS valves as follows. A permissive -relay receives a signal on reactor vessel high pressure from.

PT1(x)[N120(x)]. These relays are assigned as follows:

Channel A-K340A, Channel C-K370A, Channel B-K3408 and Channel D-K3708. -The contacts of each such permissive relay are wired in series with the coil of the arming relay -

and the paralleled contacts of the tail pipe switches associated with that channel. The arming relays are:

Channel A-K313A, Channel C-K314A, Channel B-K313B and Channel S K3148. Actuation of an arming relay seals it in through a set of auxiliary ce,;cacts.

Each of the 11 S/RV tailpipes is equipped with two pressure switches which senses actuation of the S/RV. One of these switches 10

b from each tailpipe is assigned to one trip system and the other to the other trip system.

One channel in each trip system receives input from five of these pressure switches and the other receives input from the remaining six tailpipe pressure switches.

Arming of one channel in a trip system provides an arming signal to the other channel in the trip system through auxiliary contacts on its arming relay.

The reactor high pressure inputs are derived from two pressure transmitters per channel, PTI(x)[N120(x)] for the permissive ar.J arming function and PT2(x)(N122(x)) to control the S/RV opening and closing. The two-out-of-two logic per channel will open the S/RV assigned to that channel at its LLS setpoint once the channel is armed.

Fault trees based on this logic were constructed by us to compare with those submitted by the BWROG.

Initially the two sets of fault trees did not agrce, principally with respect to the interpretation of the Hatch 2 logic configuration.

Atter further contact and discussion, comparability between the two sets of trees

{

was established.

Evaluation of the fault trees showed the change in unavailability due to the change in surveillance intervals from one to three months was 3.92 E-05 per demand which is relatively small. Because the change in unavailability is small, the changes in surveillance interval and allowed out-of-service times for testing are comparable to those for other similar systems, and failure of the LLS function does not have an immediate direct impact on plant safety, we conclude these changes are acceptable for this system.

Section 3.6 - BWR6 Safety / Relief Valves and Safety / Relief (S/PV) low Low Set (LLS) Function Instrumentation Changes requested include:

Change the Chr.nnel Functional Test and the trip unit calibration frequencies from monthly to quarterly and add a note establish 519 tne allowed out-of-service time for surveillance testing as 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Evaluation of Easis For Change LLS Function:

The same arguments were made for the BWRd LLS actuation instrumentation.s were made for the BWR4 LLS actuation instrumentation, therefore th' same concerns that were identified in the comments for the BWR4 should also apply for the BWR6 case.

In addition, since the same initiating instrumentation may be used to actuate both the Safety / relief and LLS functions, single failures could possibly result in the loss of actuation inputs to both the LLS and relief functions.

The additional information supplied by the BWROG in response to our Question Five addressed our concerns as follows.

The information provided a more complete description of the LLS logic and supplied the results of the fault tree evaluation.

There are three groups of LLS valves.

The high group, consisting of four valves:

F047D&G and F051ASF; the mid group, consisting of F051B; and the low group consisting of F0510.

Each group is sealed-in or armed by a one-out-of-three-twice logic per division.

After arming, the high group of four S/RVs is opened on two-out-of-two reactor vessel high pressure 11

s a

i per divaion and the mid and lou groups of one S/RV each are opened on one-out-of-one reactor vessel high pressure per division. One-out-of-two divisions is required-for both-seal-in and S/RV actuation.

The changes in unavailability caused by the increased surveillance intervals and the A0T were calculated and shown to be 4.48 E-08 for one logic group-and negligibly small for the one-out-of-three logic groups.

Therefore, since the changes _in unavailability are small, the failure of the LLS mode do'.urveillance intervals s not have a direct impact on plant safety, and the proposed changes in the and A0Ts are comparable to those for similar safety sys' ms, we conclude they are acceptable.

Relief P nction:

Scre concerns with the statements in the last paragraph on i

page 14 of the topical report arose.

First, in the level of redundancy 1

argument, it was not clear just what was being taken as redundant.

The three-actuation logic sets did m t appear to be entirely redundant because in two out of three cases, two of the logic _ sets plus part of the third were needed l

to open 13 out of the 20 relief valves required to preclude reactor overpressurization.

Second. it was not clear how it is known that the relief actuation-function is a small contributor to the overall S/RV function i

unavailability.

Third, it was not clear what the term "overall S/RV function unavailability" referrad to,'nor how it was determined.

Fourth, it was not clear how it has been datermined that changes in STI and A0Ts for the S/RV actuation logic result in.small contributions to the S/RV actuation logic unavailability. A previously reviewed actuation instrumentation system or subsystems had not been identified which this actuation instrumentation system or its subsystems could be compared against. Also, an analysis or calculation of the actuation instrumentation system unavailability did not appear to have been performed.

Therefore, there originally appeared to be no basis for accepting the statement that the contributions to unavailability caused by increases _ to STIs and A0Ts are insignificant and therefore acceptable.

Additiorial information supplied by the BWROG in response to our Question Six addressed the definition of what-was redundant by showing that the ASME Code allows the relief (logic actuated) and safe.y (spring) modes of actuation to be taken as redundant.

This information also included fault trees and their evaluation to show the changes in unavailabilities which occurred when the surveillance intervals were changed from one to three months.

The changes in i

unavailability were found to be 2.37-E-03 for the relief mode and 7.14 E-06 for the relief and safety mode which are relatively small, and the changes in surveillance intervals and allowed out-of-service times for testing are comparable to those granted for other similar systems. We therefore conclude the change in surveillance. intervals from one to three months and the

-inclusion of a six hour A0T for testing are acceptable.

Section 3.7

_BWR6 Pl_ art Systens Actuation Instrumentation This section contains three sulp unate systems:

The RHR Containment Spray System instrumentation, the Feedwater and Main Turbine Trip System instrumentation, and the Suppression Pool Makeup System instrumentation.

Each of these systems will be addressed independently, r

12 p

T

-~~.

s s

a)

RHR Containment Spray Instrumentation Changes requested include:

Change the Channel functional Test and the trip unit calibration frequencies from monthly to quarterly, the allowed out of-service time for repair from I hour to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and the allowed cut of-service time for surveillance testing from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

These proposed changes are consistent with those proposed and accepted for other BWR6 actuation instrumentatior,.

Evaluation of Basis For Change The actuation instrumentation performs functions similar to those performed by the isolation actuation instrumentation, therefore since the configuration is similar to that for the isolation actuation instrumentation, we agree that the analyses performed for that instrumentation can be extended to this instrumentation also.

Our review of the referenced analyses showed only a statement that valve unavailability is the dominant contributor to system unavailability. We rere unable to confirm the dominan:e of the valve unavailability because information on this unavailability analysis and its results was not presented.

Additional information was provided by the BWROG in response to our Question Seven which included a more complete system description, a fault tree for the initiation system, and calculation of the increase in unavailability caused by the increases in surveillance interval and A0T.

The increase in unavailability for one-out-of two trip systems is a relatively small 1.45 E-04 for the most conser~ative case and the valve opening and closing is shown to be the largest contributor to the individual trip system unavailability. On the basis of this information and the actuation system's similarity to those previously reviewed and accepted for the Emergency Core Cooling and Containment Isolation Systems, we conclude the proposed STI and A0T changes are acceptable.

b)

Feedwater and Main Turbine isvel 8 Trip Instrumentation Changes requested ine'.de:

Change the Channei functional Test frequency f rom month., to quarterly and the allowed out-of-service time for serveillance testing from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. These proposed enanges are consistent with those proposed and accepted i

for other BWR6 actuation instrumentF2 ion.

1 Eva N tion of Bar.is for Change The actuation instrumentation for the Feedwater/ Main Turbine Level 81 rip uses output, from level transmitters that are similar to those for the RPS Level 8 trip units.

These c:mpose a two-out-of-three logic for a single trip system.

The single trip system does not meet the single failure criterion but because BWR6 plants have a direct reactor trip on Reactor Vessel Water Level--High, level 8 13

i i

that is part of the RPS, the feedwater and main turbine trip on high reactor vessel water level 8 does not serve a safety function, but is for equipment protection only.

It was not clear however, that the RPS Level 8 trip analysis bounds the feedwater and main turbine trip because the feedwater and main turbine trip instrumentation is a single trip system using a two-out-of-three logic where the RPS Level 8 trip is a one-out-of-two per trip system with both trip systems required for trip.

Additional information describing the system and calculating the increase in system unavailability caused by increasing the surveillance interval to three months and the A0T from two to six hours was provided by the BWROG. Thc most conservative calculations using design analysis failure rates show an increase in single trip system unavailability of 1.8E-03. Using more realistic current experience failure rates shows an increase of 1.2E-04. These values for single system unavailability are small enough to conclude the requested increases in surveillance interval and A0T are acceptable.

c)

Suppression Pool Makeup System Instrumentation Changes requested include: Change the Channel functional Test and the trip unit calibration frequencies from monthly to quarterly, the allowed out-of-service time for repair from I hour to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and the allowed out of-service time for surveillance testing from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. These proposed changes are consistent with those proposed and accepted for other BWR6 actuation instrumentation.

Evaluation of Basis for Change The actuation instrumentation for the Suppression Pool Hakeup System (SPMS) does perform functions similar to those performed by the isolatiori system actuation instrumentation. However, the initiation and actuation logic configuration for this system appears to be quite different from that in the isolation actuation logic referred to. Our review of the referenced analyses shows only a statement that valve unavailability is the dominant contributor to system unavailability.

Additional information was provided by the BWROG in response to our Question Seven which included a more detailed system description which showed similarity to the LPCS ECCS initiation logic and certain primary / secondary isolation valve logic and calculation of the increase in SPMS unavailability caused by the increases in surveillance interval and A0T. Using the more conservative design failure rates gives a change in system unavailability of 2.0E-05 in going from one month to three month surveillance intervals which is small enough to conclude the requested increases in surveillance interval and A0T are acceptable.

14

o i

Section 3.8 - SWR 4 Main Cqatrol Roon Environental Control Systeq (MCRECS) Actuation Instrumentatian Changes requested include:

Change the Channel Functional Test frequency from monthly to quarterly, change the allowed out-of-service time for repair from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and change the allowed out of-service time for surveillance testing from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Those changes are consistent with those proposed and accepted for other BWR4 actuation instrumentation.

Evaluation of Basis for Change The reference given for the reactor vessel water level 1, high drywell pressure, and main steam line high flow inputs to the MCRECS was not acceptable bc:ause it did not appear to provide analyses against which to compare the reactor vessel water level 1 and main steam line high flow inputs, rarther, in the basis, such a comparison was not made and initiation of MCRECS was not addressed.

The current basis should also provide additional information on how the high control room and the refueling floor area radiation functions are similar to the reactor building exhaust-high radiation function to i

enable judgement to be made as to whether the fault trees and their evaluation are applicable to the high control room and the refueling floor area radiation functions. Addit 6nal information provided by the BWR00 in response to our Question Eigh6 addressed our concerns by providing, for each of six initiating events, the applicable logic configuratior per subsystem, references tu previously reviewed actuation instrumentation, and logic diagrams.

Each subsystem was shown to be similar to a previously analyzed system, thus the change in unavailabilities previously calculated for that system should also apply

~

to this system.

References were also provided to similar previously reviewed systems for the Refueling Floor Area High Radiation, Control Room Inlet High Radiation, and Control Room Inlet High Chlorine Level Function actuation instrumentation.

The small increases in unavailability previously calculated for the referenced systems were found acceptable for justifying similar changes in surveillance intervals and A0Ts for those systems.

Therefore, by similarity, they will also be acceptable justification for the requested similar changes in surveillance intervals and A0Ts for the MCRECS actuation instrumentation.

Section 3.9 - BWR6 Control Room Fresh Air (CRFA)

Actuation Instrumentation Changes requested include:

Change the Channel Functional Test and the trip unit calibration frequencies from monthly to quarterly and the allowed out-of-service time for surveillance testing from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. These changes are consistent with those proposed and accepted 15 a

r n-

. - -v.

r,

-,,,w

.,,-,r r

e

-e-~

_. _ _ _ _ _ _. _ ~ _ -. _

1 for other BWR6 actuation instrumentation.

Evaluation of Basis for Change i

The reference given for the reactor vessel water level 2 and the high drywell pressure inputs to the CRfA system in Ref. 5, page 5-11 was not satisfactory because it did not present any information on these two trip inputs and it presented no information which would indicate that the CRFA system initiation instrumentation was included in the analysis.

l The secondary containment statement mentions only the fuel handling area ventilation exhaust high radiation and the pool sweep exhaust rrdiation inputs to the secondary containment isolation actuation system. No fault trees or other analysis is presented in that reference for l

initiation of secoi.dary containment isolation or isolation of the HCR I

and initiation of the CRFA system.

Additional information provided by the BWROG in response to our Question Nine addressed the lack of information in the topical report for the Control Room Fresh Air (CRFA) actuation instrumentation by confirming that the same isolation logic which isolates the secondary containment isolation valves also actuates i

the CRFA system.

Also provided were logic diagrams that identify the equipment and logic configurations for the subsystems which actuate the CRFA system.

These logic configurations have been previously analyzed in similar systems for the secondary containment isolation system and i

found areptable for justifying similar requested changes in surveillance intervals and A0Ts for that actuation instrumentation.

Therefore, by similarity, the results will also be acceptable justification for the requested similar c.hanges in surveillance intervals and A0Ts for the CRFA system actuation instrumentation.

Section 3.10 - BWR4 ed BWR6 Control Rod Block Instrumentaliqu

+

Changes requested include:

Change the allowed out-of-service time for repair in Table 3.3.61, Items 5 and 6 from I hour to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and establish the allowed out-of-service time for surveillance testing at 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> by adding a note to Surveillance Requirement 4.3.6.

These requested changes are consistent with similar changes proposed and accepted for other BWR4 and BWR6 actuation instrumentation that is common to the RPS and ECCS actuation instrumentation.

l Evaluation of Basis For Change l

The basis fcr justifying these requested changes is stated tc ba l

included as part of the basis for changing the STI given in Ref. 2.

We reviewed this reference and found that it does not explicitly address i

the extension of the A0Ts.

However, in the case of the 50V level rod i

block, the same type of instrumentation is used to provide the rod block as is csed to provide the_ RPS scram signal, from Ref. 1, the effect on core damage frequency of changing the repair A0T to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the surveillance testing A0T to 6 Fours was found to be negligible for the RPS.

Since the rod block instrumentation is similar in type and 16 i

s e

configuration to that used for the RPS trip, it is reasonable to assume that the increase in rod block unavailability caused by the increases in the A0Ts for the SDV level input is also negligible.

On this basis we conclude the requested changes are acceptable.

For ~.he RCS Recirculation Flow sensors, the increase in A0Ts for test and repair was not explicitly addressed.

However, in Ref. 1, increasing the STI from 31 to 92 days does not result in a significant increase in the APRM Flow-Biased Neutron Flux scram unavailability and it is further stated that analyses indicated that increasing the test A0T to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the repair A0T to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> would produce an even smaller effect on the system unavailability.

Therefore, if these increases are acceptable for the signal input to the flow-biased APRM trip, similar test and repair A0T increases should also be acceptable for the same type of input to the rod block instrumentation since it should produce a correspondingly insignificant effect on the rod block instrumentation unavailability. On this basis we conclude the requested changes are acceptable.

17

4.<_

47..gw-f

,r, a, 4w,.---

'}

.-2,

,s.

q t' u. <,_

.1 r

.a

' s,

g I

r -,

6 g

s-2

/.,,

t t

F

' \\

5 e

W f

18 YTOYhe 3 pnp ep.% %

4 i

V.

REFERENCES l

1.

W. P. Sullivan, et al., "BWR Owners' Group Technical Specification Improvement Analysis for BWR Reactor Protection System," General Electric Company, March 1988 (NEDC-30851P-A),

1 2.

S. Visweswaran, et al., " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation," General Electric Company, October 1988 (NEDC-30851P-A, Supp.1).

3.

D. B. Atcheson, C. Ha, C. L. Larson, R. P. Raftary, W. P. Sullivan, "BWR Owners' Group Technical Specification Improvement Methodolo9y (With Demonstration for BWR ECCS Actuation Instrumentation) Part 2," General Electric Company, December 1988 (NEDC-30936P-A).

4.

L. G. Frederick, et al., " Technical Specification improvement Analysis for BWR Isolation Instrumentaticn Common to RPS and ECCS Instrumentation,".eneral Electric Company, March 1989 (NEDC-30851P A, Supp. 2).

5.

W. P. Sullivan, et al., "Techni'. Specification Improvement Analys's for BWR Isolation Actuation Instrumentation," General Electri: Comptny, July 1990 (NEDC-31677P-A).

6.

D. B. Atcheson, L. G. Frederick, W. P. Sullivan, P. T. Tran, "BWR Owners' Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation) Part 1," General Electric Comnany, December 1988 (NEDC-30936P-A).

7.

W. P. Sullivar., " Addendum to Bases For Changes to Surveillance Test Intervals And Allowed Out-Of-Service Times For Selected Instrumentation Technical Specificaiions," General Electric Company, GENE 770-06-2, February 1991.

8.

W. P. Sullivan, "BkROG Response to NRC Questions On Topical Report GENE-770-06-1," Genrral Electric Company, December 1931 (BWROG-91157).

I d

e 1

19 l

l

=. j 1

f REVIEW 0F TOPICAL REPORI GENE 770-06-1 l

TABLE OF RECOMMENDATIONS FOR PROPOSED CHANGES l

CFT Sury. Int.

Sury. Test A0T Repair A0T l

l l

System Appl Curr Prop Rec. Act.

Curr Prop Rec. Act.

Curr Prop Rec. Act.

FW/MIT BWR4 31d 92d Accept.

2 hr 6 hr

' Accept.

NR NR*

NR*

SP&DW SPR BWR4 N/A**

l EOC-RPT Both 31d 92d Accept.

2 hr 6 hr Accept.

I hr 12 hr Accept.

ATWS-RPT Both 31d 92d Accept.

2 hr 6 hr Accept.

I hr 24 hr Accept.

RCIC Inst Both N/A))

S/R Valves BWR4 N/A**

S/R Valves BWR6 31d 92d Accept.

2 hr 6 hr Accept.

NR NR*

NR* _

[LLS BWR4 31d 92d Accept.

2 hr 6 hr Accept.

NR NR*

NR*

" LLS BWR6 31d 92d Accept.

2 hr 6 hr Accept.

NR NR*

NR*

RHR Con Sp BWR6 31d 92d Accept.

2 hr 6 hr Accept.

I hr 24hr)

Accept.

FW/MTT

' BWR6 31d 92d Accept.

2 hr 6 hr Accept.

NR NR*

NR*

SPMU Sys PWR6 31d 92d Accept.

2 hr 6 hr Accept.

I hr 24hr)

Accept-MCREC BWR4 31d 92d Accept.

2 hr 6 hr Accept.

2 hr 24 hr Accept.

CRFA BWR6 31d 92d Accept.

2 hr 6 hr Accept.

NR NR*

NR*

l CR Block Both 92d NR Accept.

2 hr 6 hr Accept.

I hr 12 hr Accept.

NR = Not Requested in topical report N/A - No Auto initiation involved

- If taking channel for Surveillance does not cause loss of functions, otherwise no change.

}

))

- Refer to Review of T.R. GENE 770-06-2 U '

s s

0 BWROG'S RESPONSE TO HRC QUESTIONS ON TOPICAL REPORT GENE 770-06 1

" BASES FOR CHANGES TO SURVEILLANCE TEST INTERVALS AND ALLOWED OUT OF-SERVICE TIMES FOR SELECTED INSTRUMENTATION TECHNICAL

$PECIFICATIONS", FEBRUARY 1991 m

e e

\\

6 BWROG'S RESPONSE TO NRC QUESTIONS ON TOPICAL REPORT GENE 770-06 1

Reference:

1)

Nine (9) questions from Don Lasher, EG&G, Faxed May 13, 1991.

2)

GENE-770 06-1, " Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times for Selected Instrumentation Technical Specifications", February 1991.

3)

NEDC-308 SIP A, " Technical Specification Improvement Analysis for BWR Reactor Protection System", March 1988.

4)

NEDC-30936P-A, " Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation, Part 2",

December 1988.

5)

NEDC-30936P-A, " Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation, Part 1", December 1988.

6)

NEDC-31677P-A, lechnical Specification Improvement Analysis for BWR isolation Actuation Instrumentation *, July 1990.

E LSTION 1 For the Suppression Pool (and Drywell) Spray System (BWR4), the basis presented appears contradictory in that actuation instrumentation consisting of Drywell Pressure High; Containment Pool Pressure-High (1 don't know where this comes from); Reactor Vessel Water Level-Low Low-Low, level 1; and Timers, Systems A&B is listed but in the Basis it is stated that the BWR design for this system is manually initiated and controlled and that there are no automatic functions or initiation instrumentation. Can you clarify and correct this apparent contradiction?

RESPONSLIO 00EST10N 1 In the proposed BWR 4 Improved Standard Technical Specification (ISTS),

the suppression pool and drywell spray auto initiation signals have been removed (See Enclosure 1 for reason for removal). The only remaining auto initiation signal from Table 3.3.9-1 that will appear in the BWR 4 ISTS is for the Feedwater System / Main Turbine Trip System.

If a plant should have suppression pool and drywell spray auto initiation similar to that in a BWR 6 plant, the bases for STI and A0T changes are as provided in Section 3.7 for BWR 6 plants.

CUESTION 2 For the EOC-RPT (BWR4 and BWR6), we need more specific reference to the system or subsystems in NEDC 30851P because it appears that the logic arrangement is significantly different for these initiating signals (two-out-of-two TSV closure and two-out-of-two TCV Low 011 Pressure).

These signals are 'anded" with two channels of Turt;ine 1st Stage Pressure >30% RTP.

It is not evident that the analysis of NEDC 30851P __

i BUROG'S RESPONSE TO NRC QUESTIONS ON TOPICAL REPORT GENE-770-061 covered this particular logic configuration or that the results of that analysis bound this case.

RESPONSE TO OVESTION 2:

The logic for the turbine control valve (and turbine stop valve) trip for RPS-Scram and End of Cycle (E0C)-Recirculation Pump Trip (RPT) is presented in Enclosure 2.

The same sensors and logic relays perform both the RPS-Scram and E0C-RPT functions.

The dual trip functions are achieved by different contacts in the loolc relays.

The RPS Stram logic is 1 out of 2 channels twice required for trip.

For the turbine stop trip function, two signals from individual stop valve position switches are required to trip an individual channel.

For the turbine control valve closure, one sensor input is required to trip an indivioual channel. A trip circuit bypass is provided in each channel for reactor operation below 30% power.

For reactor operation above 30%

power, a relay contact in each channel opens, if a bypass relay fails to transfer out of the bypass position, the failure is annunciated in the control room.

Therefore, a failure to remove the bypass would be detected within a short time period.

The E0C-RPT logic is 2 out of 2 channels required to trip an individual logic Division.

Each of the two logic Divisions trip both recirculation pumps.

Failure of the t./ APT function therefore requires a failure of both logic Divisiow.

Each logic Division has contacts from two of the bypass relays.

laese contacts are in the closed position when reactor power is above 30% (trip function not bypassed). This is different than the bypass relay contacts in the RPS trip circuit which are open during operation above 30% power.

As with the RPS trip circuit, a failure to remove the bypass would be detected within a short tirae period.

A calculation of the unavailability of the two trip functions of each trip logic is provided in Enclosure 2.

The change in E0C-RPT trip function unavailability when the surveillanca interval is extended from 1 to 3 months is lower for the turbine stop valve trip function and slightly higher for the turbine control valve trip function than the same trip functions for RPS-Scram.

However, the small increase in E0C-RPT unavailability (represented by small increase risk of an MCPR violation) is judged to be uffset by the benefits associated with the similar approved STI and A0T changes for the RPS Scram function.

Therefore, it can be concluded that the STI and A0T changes for EOC-RPT trip function is bounded by the approved RPS analysis.

OVEST10N 3 For the ATWS-RPT (BWR4) as we discussed over the phone, he newer BWR4 (Hatch) logic configuration and the one you indicated is impicmented for most BWR4s, is the two-out of-two per trip system on either reactor high pressure or reactor vessel water level low with trip of either trip system tripping both Recirculation Pumps.

I l

r i

s BWROG'S RESPONSE TO NRC QUESTIONS ON TOPICAL REPORT GENE 770-06 1 i

i The topical report states that the effect of changes to STI and A0Ts for ATWS-RPT in:;trumentation on the reactivity shutdown failure frequency is negligible because the RPS failure frequency is low (5.4E-06/yr from NEDC 30851P, page 5 29) and the change in overall ATWS-RPT function unavailability due to the STI and A0T changes

(<1E-02/ demand) is small. This appears to say that even though the ATWS-RPT failure frequency may be higher, when it is combined with that i

for the RPS, the combined effect is small.

It is not clear how this combination is made or how the numbers were calculated.

Can you supply some additional information explaining how these numbers were arrived at and how they were combined with the RPS?

The last sentence on page 6 of the report seems to state that the same small change in ATWS-RPT unavailability due to STI and A0T changes can be expected for other logic designs.

It is not clear how this can result since it would appear that the logic arrangement should have a large influence on the fault trees and their evaluation which means a change in the logic configuration could result in a large change in unavailability and hence failure frequency.

This would indicate that the reliability of a system is nearly independent of its logical arrangement.

Please clarify this concern.

RESPONSE TO OVESTION 3:

It was stated in Reference 2 that the trip logic for ATWS-RPT for the Hatch 2 lead ITS plant is one out of two channels per trip system for each trip function.

Both trip systems are required to trip the two recirculation pumps. This was the configuration at the time of the analysis.

The ATWS-RPT configuration that is being installed during the current Hatch-2 refueling outage has the same type logic as the BWR 6 lead plant.

The ATWS-RPT logic for the BWR 6 lead plant (currently being installed for the BWR 4 lead plant) is two out of two channels required to trip each trip system for each trip function (low water level and high reactor 3ressure).

It was stated in Reference 2 that the change in unavailaaility when the surveillance interval is extended from 1 to 3 l

months is small (< IE-02/ demand). Calculation of this unavailability change is presented in Enclosure 3.

Unavailabilities are calculated using trip unit failure rates from original design analyses and failure rates that reflect more current operating experience.

The calculated change in unavailabilities for each trip function is approximately 2E-03/ demand based on failures rates from the original design 3nalysis and 2E-04/ demand based on failure rates reflecting current operating experience.

l The relative effect of the ATWS-RPT unavailability change on the reactivity shutdown failure frequency is also shown in Enclosure 3.

Assuming only one ATWS-RPT trip function (low water level or high reactor pressure) and the RPS failure frequencies calculated in Reference 3, the relative change in reactivity shutdown failure frequency when the ATWS-RPT channel functional test is extended from 1 to 3 mo.hs is negligible (<2.0E-08/ year).

This calculation does not include other ATWS-RPT components which are not part of the channel -

s BWROG'S RESPONSE TO NRC QUESTIONS ON TOPICAL REPORT GENE-770 06-1 ft~.;tional tests (e.g., circuit breakers). However, the model rovides adet,uate indication of the small relative effect of the surveil ance interval extension.

This negligible change in reactivity shutdown failure frequency is offset by the benefits from reduced inadvertent scrams which is discussed in Reference 3.

It is also stated in Reference 3 that a same small change in ATWS RPT unavailability due to STI and A0T extensions can be expected at other BWR plants. This doer not imply that the reliability of the ATWS-RPT is independent of logic arrangement.

However, the logic designs of the different BWR plants can be expected to have redundancy comparable to the above analyzed design based on the.specified design requirements for ATWS-RPT. A small change in ATWS-RPT unavailability can be expected if comparable redundancy exists.

W EST10N 4 For the BWR4 LLS function, it is stated that failure of the LLS function to initiate will not have an immediate adverse effect on the plant safety and therefore, that any small change in actuation instrumentation unavailability due to changes in the STI and A0Ts may be neglected.

However, the changes in the SRV LLS function actuation instrumentation unavailability caused by changes in the STI and A0Ts have not been shown to be small enough to neglect. The argument is made that the reasoning which supports making the requested changes is bounded by the analyses performed for the ECCS and isolation actuation instrumentation and further that the analyses apply because the components used in the systems are the same or similar and the number of failures needed to disable the system are comparable.

This argument as presented is unacceptable because it appears to fail to consider the manner in which those components are assembled into the LLS logic.

For example, the LLS logic is an armed logic which is not similar to any of the ECCS or isolation actuation logic. Some of the subsystems may be similar but the manner in which they are organized into the logic is different, therefore the fault trees and their evaluation may differ and the results may not be bounded by the previous analyses.

This has not been shown and neither has it been shown that the number and type of failures needed to disable the LLS actuation instrumentation is comparable to that for an identified previously reviewed ECCS or isolation actuation system. The topical report should be revised to demonstrate the similarity of this logic system to a specific reference logic system or systems.

RESPONSE _TO OutSTION 4: provides a fault tree of the low-low set (LLS) for the Hatch 2 plant. There are 11 safety / relief valves (S/RVs) at the Hatch 2 plant.

Four of these are actuated by the LLS.

Failure of the LLS function requires failure of all 4 LLS valves.

Each LLS valve is armed by a relay in a logic channel initiated by a high reactor pressure trip unit and one of the S/RV tail pipe pressure switches. Two of the valve arming logic channels receive trip signals from 5 tail pipe pressure switches.

The remaining two arming logic channels receive trip signals from 6 tail pipe pressure switches.

Each of the 11 S/RV tail pipes t - -

e BWROG'S RESPONSE TO NRC QUESTIONS ON TOPICAL REPORT GENE 770 064 have two pressure switches.

After arming, each LLS valve is actuated by twc out of two relays each initiated by a high reactor pressure trip unit.

There are a total of 22 individual pressure switches (2 per S/RV) and and 12 individual reactor pressure trip units.

A calculation of the LLS unavailability as a function of surveillance interval is given in Enclosure 4.

lwo LLS valves were assumed to open.

The change in Li-unavailability when the surveillance interval is extended from 1 to 3 months is 3.92E-05/ demand. This change is acceptably low based on the function and failure :ensequences of the LLS (i.e., reduce the number of load cycles on the containment and S/RV discharge lines and reduce the number of S/RV actuations during the plant lifetime).

Q1lESTION 5 For the BWR6 LLS function, the same arguments are made for the actuation instrumentation as were made for the BWR4 LLS actuation instrumentation, therefore the comments made for the BWR4 case apply for the BWR6 case also.

In addition, since the same initiating instrumentation may be used to actuate both the relief mode and the LLS mode, it appears that single failures could result in the loss of actuation inputs to more than one relief function.

These concerns should be addressed to first demonstrate the similarity of the LLS actuation instrumentation logic to an identified previously reviewed ECCS or isolation actuation system logic and second to analyze the apparent condition in which a single failure could result in the loss of actuation inputs to more than one relief function.

RESPONSE TO OVEST10N 5:

1. provides information on the LLSL for BWR 6 Grand Gulf plant.

There are a total of 20 S/RVs of which 6 perform the LLS function.

The seal in logic is the same for all 6 LLS valves (i.e., 1 out of 3 twice/ Division).

The LLS logic is 2 out of 2 per Division for one logic group controlling 4 LLS valves and 1 out of 1 per Division for the remaining two logic groups which each control one LLS valve.

One out of 2 Divisions are required to initiate the LLS function for both the seal in logic and LLS logic. There are a total of 12 individual reactor pressure trip units for the seal in logic (6/ Division) and 8 individual reactor pressure trip units for the LLS logic (4/ Division).

The unavailability as a function of the test interval was calculated at the LLS logic channel level.

The change in LLSL unavailability for a single logic group when the surveillance interval is extended from 1 to 3 months is 4.5E-08/ demand.

The change in LLSL unavailability for 1 out of 6 LLS valves (or 1 out of 3 logic groups) is negligible. This change is acceptably low based on the function and f ailure consequences of the LLSL (i.e., reduce the number of load cycles on the containment and S/RV discharge lines and reduce the number of S/RV attuations during the plant lifetita). l

BWROG'S RESPUNSE TO NRC QUESTIONS ON TOPICAL REPORT GENE-770-06-1

=

OVEST10N 6 For the BWR6 Relief function, several concerns with statements in the last paragraph on page 14 need to be addressed in the topical report.

First, in the level of redundancy argument, it is not clear just what is being taken as redundant.

The three actuation logic sets are not entirely redundant because in two out of three cases, two of the logic sets plus part of the third are needed to open 13 out of the 20 relief valves required to preclude reactor overpressurization.

[The safety mode may not be truly redundant because of the different (higher) opening pressures.]

Second, it is not clear how it is known that the r

relief actuation function is a small contributor to the overall S/RV function unavailability.

Third, it is not clear what "overall S/RV function unavailability" refers to, nor how it is determined.

Fourth, it is not clear how it has been determined that changes in STI and A0Ts for the S/RV actuation logic result in small contributions to the S/RV actuation logic unavailability. A previously reviewed actuation instrumentation system or subsystems have not been identifl2d which this actuation instrumentatior, or its subsystems can be compared.

An analysis or calculation of the actuation instrumentation system unavailability does not appear to have been performed. We do not appear to have a firm basis for accepting the statement that the contributions to unavailability caused by STI and f.0T increases are insignificant and therefore, are acceptable.

RESPONSE TO OVEST10N 6:

For the BWR 6 lead plant 13 of the 20 S/RVs are required to prevent reactor overpressure. The S/RVs are designed to limit the primary system pressure, including transients, to the requirements of the ASME Section III.

The ASME code allcws one half of the S/RVs to open in the safety (spring) mode and one-half to open in the pressure relief mode.

Enclosurt 6 provides a description of the pressure relief logic. One group of 9 S/RVs are actuated by 1 out of 2 logic Divisions.

Each Division is actuated by tripping 2 out of 2 reactor pressure trip unit channels. A second group of 10 S/RVs are tripped by another set of reactor pressure trip units using a similar logic configuration. One S/RV is tripped by a third set of reactor pressure trip units.

As discussed in response (5), 6 of the 20 S/RVs also have a LLS function in addition to the pressure relief function.

The LLS seal ia logic uses the same trips units as the pressure relief logic.

Because of thir ommonality, no credit was taken for the LLS relief mode when r

calcuating the pressure relief function unavailability.

A fault tree model of failure of the S/RV relief and safety mode at the logic channel level is provided in Enclosure 6.

The unavailability was calculated using trip unit failure rates from original design analyses and failure rates that reflect more current operating experience.

The calculated change in relief and safety mode unavailabilities when the surveillance test interval is extended from 1 to 3 months is approximately 7E-06/ demand based on failures rates from the original design analysis and 4E-08/ demand based on failure rates reflecting ~

e--

w-ey--'v,ww eis,ye-e-mw,-m.g e

,=-w.q-

,=n,,

-e,.--

.g,,ma--.v-w-er-

,-g--.y,,-,e,. * "

m.g--yi ymi-,y-mu-imay--+p-y-g-r Fw s

• 'w a-

c-W=-u'r 9 r wv?7='TEis'Tme-'*Pt'W ugs-vt'i'7=y

'W*1

-wr-Tim 3m-'9'--7-s-'rf'-*"

e BWROG'S RESPONSE TO NRC QUEST'005 ON TOPICAL REPORT GENE-770-06 1 current operating experience.

This increase in calculated unavailability is insignificant and is offset by the similar benefits of reduced testing discussed in Section 4.2 of the ECCS Actuation Instrumentation Analysis, Reference 4.

OUESTION 7 For the BWR6 Plant Systems Actuation Instrumentation (RHR Containment Spray System, Feedwater & Main Turbine Trip-System, and Suppression Pool Makeup System Instrumentation) provide a more specific reference system to which each of these three systems can be compared.

Also provide additional information describing the valve unavailability analyses and results.

RESPONSE TQ_0 VEST 10N 7:

a)

RHR Containment Soray System Instrumentation - The containment spray is initiated by 1 out of 2 trip systems.

Each system has a 10.85 minute timer which is initiated by either high drywell pressure or low reactor water level in a 1 out of 2 twice logic.

One RHR valve is opened and closed in each system after the time delay when both high containment pressure and high drywell pressure are present in a 1 out of twt, logic.

The reactor level and drywell prassure trip units are common to the ECCS initiation logic which were considered in the Reference 4 analysis.

A fault tree of the containment spray initiation including the valves that have to open and close is presented in Enclosure 7.

The unavailability was calculated as a function of the level / pressure trip channel surveillance interval for a single trip system and 1 out of 2 trip systems. Using a level / pressure trip channel f ailure rate based on current experience, the change in calculated unavailability when the surveillance interval is extenaed from 1 to 3 months is 8.75E-05/ demand. No credit is taken here for manual actuation if the automatic initiation logic should fail.

This small increase in calculated unavailability is insignificant and is offset by the similar benefits of reduced testing discussed in Section 4.2 of the ECCS Actuation Instrumentation Analysis, Reference 4.

b)

Feedwater and Main Turbine level 8 Trio - This trip system is not included in the the BWR 6 Improved Standard Technical Specif uation.

The reason for this is because a direct scram is provided for the level 8 trip (uses separate set of trip units and logic that was considered in the Reference 3 analysis).

For those plants that do not have a direct scram on level 8 trip (such as BWR 4 plants), the effect of changes to the surveillance intervals of the feedwater and main turbine level 8 trip system is discussed in Section 3.1 of Reference 2.

Calculations of the unavailability for this specific trip function for different surveillance intervals are provided in Enclosure 7.

The calculated change in unavailabilities when the surveillance interval is extended from 1 to 3 months is acceptably low (1.24E-04/ demand).

c)

Sucoression Popl Makeuo Systpm Instrumeatation - The same sensors that initiate the low pressure ECCS and certain primary / secondary containment isolation valves are also used in the suppression pool _

i t

BtfROG'S RESPONSE TO NRC QUESTIONS ON TOPICAL REPORT GENE-770-06-1 makeup system initiation logic.

The logic is discussed in Enclosure 7.

Also included in Enclosure 7 is calculated change in unavailapility when the surveillance intervals are changed from 1 to 3 months.

The small increase in calculated unavailab.lity is insignificant and is offset by the similar benefits of reduced testing discussed in Section 4.2 of the ECCS Actuation Instrumentation Analysis, Reference 4.

OVESTION 8 For the BWR4 Main Control Room Environmental Control (HCREC) Actuation Instrumentation, please provide specific references for the reactor vessel water level 1, high drywell pressure, and main steam line high flow inputs to the MCREC systems. The quoted reference did not address the water level 1 and main steam line high flow inputs.

More specific references should be provided for the liCREC inputs and the similarities in configuration, components and qualification addressed to allow comparison to be made and conclusions drawn.

Also, please provide additional information on how the high control room and the refueling floor area radiation functions are similar to the reactor building exhaust high radiation function to enable judgement to be made as to whether the fault trees and their evaluation are applicable to the high control room and the refueling floor area radiation functions.

P M !iS1.T0 OVEST10N 8:

The trip unit channels and logic that actuate the BWR 4 Main Control Room Environmental Control (MCREC) are provided in Enclosure 8.

MCREC is made up of two redundant independent subsystems which are initiated by Trip System logic A & B (one subsystem by logic A and one subsystem by logic B). The following is a summary of the different trips, type of logic, and where covered by previous analysis.

LOGIC PER

[1LN.1 SUBSYSJE ANALYSIS REFERENC[

LOCA 1 out of 2 Reference 5 Page B-5 twice

& Reference 6 Page 5-21 Main Steam Line 2 out of 2 Reference 6 Page 5-21 High Rad Main Steam Line 2 out of 2 Reference 6 Page 5 21 Break Refueling Floor 1 out of 1 Reference 6 Page 5-21 Area High Rad Control Room 1 out of 1 Reference 6 Page 5-21 Inlet High Rad Control Room 1 out of 1 Reference 6 Page 5-21 Inlet High Chlorine _____

A 4

BWROG'S RESPONSE TO NRC QUESTIONS ON TOPICAL REPORT GENE 770-06 1 The logic that initiates MCREC is s!nilar to the logic given in the case studies for PWR plants.

Instead of closing an inboard and outboard valve, the logic initiates redundant MCREC :;ubsystems.

The logic therefore is analogous to the logic given under the column headed

" Logic Type Per Valve Per Variable".

The only exception is the logic for a LOCA signal.

The trip units and logic for the LOCA signal is the same signal that actuates the ECCS low pressure systems analyzed in Reference 5.

The LOCA signal is also similar to the 1 out of 2 twice logic in Reference 6 except there are 2 variables instead of 4.

i MESTION 9 in the BWR6 Control Room fresh Air (CRFA) Instrumentation, the reference given for the reactor water level 2 and the high drywell pressure trip in NEDC 31677P-A, page $-11 does not present any information on these two trip inputs and it presents no information that would indicate that the CRFA system initiation was included in the analysis.

The secondary containment statement mentions only the fuel handling area ventilation exhaust high radiation and the pool sweep exhaust radiation inputs. No fault trees or cther analysis is presented in that reference for initiation of secondary containment isolation or-isolatioi, of the MCR and initiation of the CRFA system.

Please provide more complete refarences which describe the system and show the similarities between the referenced instrumentation and this actuation instrumentation so we can complete our evaluation and reach a conclusion regarding the acceptability of this instrumentation.

RESPONSE TO OVEST10N 9:

The BWR 6 CRFA is actuated by the same isolation logic that isolates the secondary containment isolation valves. The trip unit channels and logic for the BWR 6 CRFA are presented in Enclosure 9.

The BWR 6 logic 1

is the same type of logic that initiates the BWR 4 MCREC.

The same case studies given in Reference 6, page 5-21, and Reference 5 for the LOCA initiated signal (reactor Level 1 or high drywell pressure) apply for the BWR 6 CRFA.

l

. u..

..,_a-_,~_._._-_---.~.---,_,.,...---~....,.-.--,_

dd%D 4A.a AmAM L+4J wM dA 445 J4w*

r44e,4 e

&+=

0

--4.h.MA4.A_,A*-he+

nE2JhaJm. e 4-tam 4-+12,4 4akh.

MA.,_u.4@&

$m14wa

-..MS3% e

-p$344, 4 h-J.E'Aha.N &4. m a p d JA,43.__._4h-64 As.J.

?

9 I

r l

t 1

l 0

9 f

f r

i r) 6 b

7 N

h I

ENCLOSURES 1

b l

9 7

O n 6

4 i

e

=

s--w....

.-,,.2+,w

=

.c

.-w

.e,-,s.--..

v,

...,e..,ve.-,-

w-m...,

,3

,we,

l10 OS (J/'l'

}

S,

]C3 ![

f w

\\,

(

M M

M M

M g== 'g aaa aa gg/j a.:.:

I

/

e r

4 E

• s I

=

C

' to y

m k

.h F

gE E 5 v.

l "E"

y M

U I

w

.5 I[$

l' s

5 T

=

m E

5' t

y

-4 a

E M

,f.

g i

f. "

1 k

5 k

?

E b

T-T

-d W

L s

s g

b, g

,b 2

s g ggz

=

3 c

2 I

M N

g g

l gA s

hj$t E

W 8

5 W

g G

az=n a

W l

l

=

j o

e A

a l

GE-STS (BWR/4) 3/4 3-91 7)9,/7;

tac /csseo

/

(['a 24+

e DISCUSSION OF CHANGES TO STS REY. 4 (Addendum) 100 3.3.2.2:

FEEDWATER AND MAIN TURBINE TRIP INSTRUMENTATION (continued) 2.

The Level 8 Function is assumed to function in plant specific MCPR analyses (if it is not, this specification would not be app'.icable to that plant).

MCPR limits kry only required to be met when it 25% RTP, therefore this function, which only serves to support MCPR, has its applicability consistent with ripecification 3.2.2.

4.

The BWR design for suppression pool and drywell spray (where included in the design) are manually initiated and controlled systems.

There are no E!tomatic functions and therefore no corresponding instrumentation.

S.

756 tiquirem3nt for system operability in the LCO requires the same three channels to be coerable.

The specific number of channels in the FW/ turbine Level 3 trin system is a detail of the system design which is located in the Bases.

t BWR/4 3/4 3-91 7/31/89 l

o 2 -

QUESTION 2 ENCLOSURE 2 TURBINE CONTROL VALVE /STOP VAINE TRIP PAGE 1 OF 3 RPS TRIP TRIP SYSTEM A TRIP SYSTEM D I

i K10A-K10E K10B-K10F i

K9A K9B COSTACT OPENS CONTACT OPENS i

K8A

> 30 % POWER K8B

> 30 % POWER K10C-K10G K10D-K10H

{

K9C K9D.

CONTACT OPENS CONTACT OPENS

}

K8C -

> 30 % POWER K8D

> 30 % POWER 1

i m

EOC RECIRC_ PUMP TRIP a

i K10A-K10F DIVISION 1 t

K9A K9B TRIPS BOTH CONTACTS OPENS RECIRC PUMPS t

3 K8A-- K8B

> 30 % POWER I

K100- K10G DIVISION 2

)9C K9D TRIPS BOTH j

COFCACTS OPENS RECIRC PUMPS l

K8C--K8D

> 30 % POWER l

l E8A, B,C, D - CHANNEL IDGIC. REIAYS - CONTROL VALVE TRIP K10A,B,C,D,E,F,G,H - CHANNEL LOGIC RELAYS - STOP VALVE TRIO

[

K9A,B,C,D - CHANNEL LOGIC RELAYS - REACTOR PRESSURE BYPASS FOR REACTOR POWER > 30 t, - RPS IDGIC CONTACTS OPEN

- EOC-RPT IDGIC CONTACTS CLOSED l

L e

i

ENCLOSURE 2 PAGE 2 OF 3 QUESTION 2 TURBINE CONTROL VALVE /STOP VALVE TRIP REFERENCE IEEE - 352

_ TURBINE CONTROL VALVE TRIP FUNCTIOff UNAVAILABILITX RPS = 2 [ (XY)2j3, (xy)(pBP)

• I BP)

)

+ CCF ECC-RPT = (2XY) /3 + (2XY) ( 2 PBP)

(2PBP)

+

F WHERE:

X = CHANNEL FAILURE RATE (FAILURES /HR. )

Y = CHANNEL SURVEILLANCE INTERVAL (HRS.)

^

^

^

P

~

BP TURBINE STOP VALVE TRIP FUNCTION UNAVAILABILITY RPS = 2 [ (2XY) /3 + (2 XY) (P

+I BP)

)+

F BP 0

EOC-RPT = (2XY)2/3 + (2XY) (2 PBP) *I BP)

+

F WHERE:

X = CHANNEL FAILURE RATE (FAILURES /HR.)

Y = CHANNEL SURVEILLANCE INTERVAL (HRS.)

P

= BWASS FAIME PROBABIN gp CCF = COMMON CAUSE MISCAL. OF TRIP UNITS PROB.

e I

h ENCICSURE 2 PAGE 3 OF 3 QUESTION 2 - TURBINE CONTROL VALVE /STOP VALVE TRIP FAILURE RATE FAILURES /HR.

TB STOP VALVE 6.00E-06 TB CONTROL VALVE 2.00E-07 IST STAGE PRESSURE 2.00E-05 (BYPASS)

RELAY 4.00E-07

========================

Failure Rate (FR) (SENSOR + RELAY)

TB STOP VALVE 6.40E-06 TB CONTROL VALVE 6.00E-07 IST STAGE PRISSURE 2.04E-05 (BYPASS)

UNAVAILABILITY (RPS)

RPS TRIP FUNCTION 1 MONTH 3 MONTHS CHANGE TB STOP VALVE 4.41E-05 2.08E-04 1.64E-04 TB CONTPOL VALVE 2.04E-05 2.11E-05 7.70E-07 UNAVAILABILITY (EOC-RPT)

EOC-RI?

TRIP PUNCTION 1 MONTH 3 MONTHS CHANGE TB STOP VALVE 3.45E-05 1.21E-04 8.66E-05 TB CONTROL VALVE 2.08E-05 2.23E-05 1.54E-06 WHERE RPS UNAVAILABILITY =

2 * ( ( ( FR*T)

• 2 ) / 3 + ( FR*T)

• PF + ( PF)

• 2 ] + CCF (STOP VALVE) 2*[((2FR*T)*2)/3+(2FR*T)*PF+(PF)*2] + CCF (CONTROL VALVE)

FR = CHANNEL FAILURE RATE T = CMANNEL FUNCTIONAL TEST INTERVAL PF = PROBABILITY OF BYPASS FAILURE CCF = COMMON CAUSE MISCAL. OF TRIP UNITS 2.00E-05

=

EOC-RPT UNAVAILABILITY a

( (2 *FR*T)

• 2) /3+ ( 2 *FR*T) * ( 2 *PF) + ( 2 *PF)

• 2 + CCF 1ST STAGE PRESSURE

- IF ANY OUTPUT RELAY ENERGIZES (BYPASS (BYPASS)

POSITION) FAILURE WILL DE ANNTJNCIATED

i ENCLOSURE 3 PAGE 1 OF 2 QUESTION 3 - ATWS-RPT LOGIC ATWS-RPT DIV. 1 DIV. 2 TRIP BOTH K43A K43C RECIRC PUMPS

-DIV. 1 DIV. 2 TRIP BOTH K43B-

~K43D RECIRC PUMPS

/

TRIPS ON LOW REACTOR WATER LEVEL OR dIGH REACTOR PRESSURE 6

ENCICSURE 3 PAGE 2 OF 2 QUESTION 3 - ATWS-RPT LOGIC FAILURE RATES (FAILURES /HRc 1 DESIGN CURRENT ANALYSIS EXPERIENCE LEVEL 2.00E-05 5.00E-06 PRESSURE 2.00E-05 5.00E-06 RELAY 4.00E-07 4.00E-07

======= -=======,,-----=============--___

Failure Rate (FR) (SENSOR + RELAY)

LEVEL 2.04E-05 5.40E-06 PRESSURE 2.04E-05 5.40E-06 UNAVAILABILITY (ATWS-RPT) PER FUNCTION ATWS-RPT TRIP FUNCTION FUi;CTION 1 MONTH 3 MONTHS CHANGE (1) DESIGN ANALYSIS FRs LOW LEVEL 3.16E-04 2.68E-03 2.37E-03 HIGH Rx PRESSURE 3.16E-04 2.68E-03 2.37E-03 (2) OPERATING EXPR. FRs IDW LEVEL 4.07E-05 2.06E-04 1.66E-04 HIGH Rx PRESSURE 4.07E-CJ 2.06E-04 1.66E-04 WHERE:

ATWS-RPT ITNCTION UNAVAILABILITY =

(1/3)*(2*FR*T)*2 + CCF CCF = COMMON CAUSE MISCAL. OF TRIP UNITS =

2.00E-05 REACTIVITY SHUTDOWN FAILURE FREQUENCY 1.!ONTH 3 MONTHS CHANGE RPS FAILUP.E TREQ.

4.60E-06 5.40E-06 8.00E-07 (EVENTS / YEAR)

ATWS-RPT UNAVAIL.

(1) 3.16E-04 2.6SE-03 2.37E-03

(/ DEMAND)

(2) 4.07E-05 2.06E-04 1.66E-04 REACTIVITY SHUTDOWN (1) 1.45E-09 1.45E-08 1.30E-08 FREQ. (EVENTS / YEAR) (2) 1.87E-10 1.11E-09 9.28E-10 l

l 1

a ENCLOSURE 4 PAGE 1 OF 13 QUESTION 4 FAILURE RATES USED IN LLS FAULT TREE FOR HATCH 2 FAILURE TFST

((1)*(2)]/2 COMPONENT RATE (1)

INTER' VAL (2)

UNAVAIL.

RESET SWITCH 1.30E-08 730 4.75E-06 1.30E-08 2190 1.42E-05 RELAY 4.00E-07 730 1.46E-04 4.00E-07 2190 4.38E-04

/

TRIP UNIT 2.00E-05 730 7.30E-03 5

2.00E-05 2190 2.19E-02 PRESSURE SWITCH 2.00E-07 730 7.30E-05

~

2.00E-07 2190 2.19E-04 DEMAND

^

FAILURE FAILURE COMPONF.NT RATE PROB.

CCF OF SENSORS 2.00E-05 2.00E-05 DC DIVISION POWER 2.00E-03 2.00E-03 LLS SOLENOID VALVE 1.31E-03 1.31E-03 TRANSMITTER 1.30E-05 1.30E-05 FAILURE PROB. OF LOW-IDW SET TEST INTERVAL FAILURE PROB.

1 MONTH 2.66E-05 3 MONTHS 6.58E-05 CHANGE FROM 1 TO 3 MONTHS 3.92E-05 l

- o o

i Page 2 of D 4

l b

E m)s'a C s

.m u#

wN 6

^ ~ $

J d

-t I

.E y

1 -

N N

se-s o

s a

M E

5 M

e, u

>d w

W j

$ "4 CL b "n h

y, w-v f

e

) w 5

21

?

' i.

=

a-e

~

N N g-y 64 6

ag 5

EE 4

w r

E E

C g

E(

J

"" e

>d E<6 l

I l-

'r.-

-w.,

y,

.w n

,p r,

g--m-g.

g.q.

,,,.a

__.mi_,,.%_

,,p,,

p y,g

,c9,,,.,,,w y9,

_,.y--

,-,w-., -, -7 y

- o.

Page 3 of 13 6

5

$4 Vos vN w

v N

I b

b V

j o

.w

~

~

l k

"4 5

l l

~

M

~

6 Q

{e c-

?2

> d

-w at' ed$

r*\\g N

/k

~3 d$U

c. >

d Y..

i Y

>a-cs a

N E

(L N

2 g

E

= tA

>d E1 l

9 i.

I'.

I

=

Page 4 of 13

-e c

fe

.e c).

kg

• 5 i :"

1 T

p y.

pv A

C Ve N m

g 2M /1 A r b NF 5n I

s Y

6 1r

• I%

j 5

C I

ra e

-.s cg v.

Im W

@is M ~t ad Yg a

U*

6 3

c.,

N] O m

/

r

=m

?q

'x

, a K

5, tX ? i 3

re d

e e d-

.u U

90 bi!

l

[

v y-u up a

(T)

)

h8 G

C W

g t

u

}

iT "N

f e

ik

's/ sEN 2

.Q o

AzY a

. _ /. e

(

L s

t $.

s nl t

A r

F

(

t 1

4 SN$

[%~d u

/

0 A

3 tem 4

1 v F O[A -

d 1

1 0 4

6 AD W 4 A

1s 4

P r

C m

pt l

N A

F i

/1 O

s

?*5

^ IL Y(WF

)i A

I V

5 c l 0 /

2 C.

O)GE

}I A2 - b H

J 3

C (N

t8 N7 I

c A

3 0

d t

4 t

I (C

G P

F D

F 0

1 E

E 5 R T

e T

rL yU e A V F T

E S

V 5

O R

)

E fL L

T A O 23 SF A

V A

00 8

)g*I O

t T 67 L

(

AN (0t 9 0 4

A 0

/

0 2, P 4

3 WI

)D W

R E

, 0 (20 T

5 A 0 3

s 0 -

k

_( m 3 0

Y A

~

A $

i LS RT ll K!

L A

A A

0 f

F 4

4 O3 A 0 3 N 0

k f.

4 0

P 6

f K

A(

4 R

LS t

E L FI Gw 4

A F

l f

0 1

in l$

af 10 FIG 4 IU f 10 %

"ILAY " 370 M L AY "334^

F AIL S [FT N y,,_;j f Alt $

K 314 AI Qf]h/

k 31 B

r. 4 s

I GB pit Ar vi7tp Rf L AY K 3 70A Rit Av t. 314 A tezet MAstip T AIL S F A!LS (YT N F Alt.S Ort N Ikip (Nil ( All 5 k 370'.[@o[)

RK 314A ([d/3 h Me3k(b)

PK SA 146E-04 t46E 04 09730 sFFOM f IG 6) m Nt?OC h DM IR ANiMil v

I All 3 M*

on zacs I w -05

[o j

r.gur e 6 L OV - L t]V SET F AIJL i IPLE FflP HAIOf 2 l

i

^A 3

10 IIG f [\\

r pt s ss t y t-._'

p_, -

1Ft {Tf[}

(Al.%TA}

bg3, NhX, Fi{.5$ w[

(LAv & Jt.)A 5%dif(H F Att $

F AfL5 (PTN gyq N 1

s[g3qg) e N WL (bk/Wh 7 30t -o,

/

_._.L._...

(kJ6'8h E'll AT F I4ftA GitAy e.13 3A I Alli F Allt ty'(N g

n t

(IA) zn

[M3/98 oE 6N h)*0A "4

(F RCH T !G. *,)

3 sq.,94 C*

(K97o$)

n'. -+

c m., m a)

LOV-LOV SEI f~AULi TPt E l' flR liAT Cil 2

E' Monday September 23, 1991 15:08 LL$1.RPT Page:

1 C:\\BWR3G\\ GENT?70\\LLSI.CAF CAFTA Fault Tree Report 9 03-91 1:43 Pag a

1 Tree list:

LLS AND Olvt DIV2 DIV1 09 DIVIP DIV1R LL$1 CCFS(h5 LL51 AND C8A CHC DIV2 OR DIYlP DIV2R LL52 CCFSEh3 LL52 AMD CHB CHD CHA OR K341A K342A K313A t013B CMB OR K341B K342B K313B F013G CHC OR K371A K372A K314A F013F CHD OR K371B K372B K3146 F0130 K341A OR N521A h120A NB20A RK341 A K341B OR N521B N120B N6202 RK3418 K371A OR NE21C N120C hS20C RK371A K371B DR NE21D N1200 N0200 RK3718 K342A OR nE22A N122A RK342A K342B OR N022B h!22B RK3(28 K372A OR hS22C N122C RK372A K372B OR kE22D N122D RK372B K313A OR K340A RK313A TPA r313S OR K340B RK313B TPB K314A OR K370A RK314A TPC K314B OR K3703 RK314B TPD K34DA 07 RK340A hE2

• h120A K3408 Ok RK340B N620B N1208 K370A 0 ',

RK370A h620C N120C K370B OR RK3700 hf-?00 N1200 TPA AND N302A K314A1 TPB AhJ N301A K3J451 TPC AND N3020 K313Al TPD AND N301t K31381 R302A OR N302A1 K14A h301C OR h301C1 K14C K314A1 OR K370A RK314A K314B1 04 K370B RK314B K313Al OR K340A RK313A K31381 OR K340B RK313B Monday September 23, 1991 15:08 LLSI.RPT Page:

1

Monday Sept ember 23, 1991 15:08 LLS1.RFT Page 11 of 13 p,ge:

3 C:\\BVtDG\\ GENE 770\\LLSI.CAF CAFTA Fault free Report 9 03-91 1:43 Pag e

3 Gate descriptions:

Gate hane Description CHA CHANNEL A FAlis CHB CHANNEL B FAILS CHC CH&hNEL C FAILS CHD CHAhNEL D FAILS C1V1 DIY1510N 1 LDGlC FAILS CIVZ DIV1510N 2 LOGIC FAILS K313A RELAY K313A FAILS K313Al RELAY K313A FAILS OPEN K313B RELAY K313B FAILS K31381 RELAY K3135 FA!LS OPEN K314A RELAY K314A FAILS K314A1 RELAY K314A FAILS OPEN K314B RELAY K314B FAILS K314B1 RELAY K314B FAILS OPEh K340A RELAY K340A FAILS K34DB RELAY K3408 FAILS K341A RELAY K341A FAILS K3418 RELAY K3419 FAILS K342A RELAY K342A FAILS K342B RELAY K3428 FAILS K370A RELAY K370A FAILS K370B RELAY K370B FAILS K371A RELAY K371A FAILS K3719 RELAY K3718 FAILS K372A RELAY K372A FAILS K3728 RELAY K372B FAILS LLS FAILURE OF LOW-LDV SET LOGIC LLSI LLS LOGIC 1 FAILS LL52 LLS LOGIC 2 FAILS h301C N301C PRESSURE SV FAILS N302A h302A PRESSURE SW FAILS TFA PRESSURE SV LOGIC A FAILS TPB PRESSURE SV LDGIC B FAILS TPC PRES SW LOGIC C TPD PRES SW LOGIC 0 Konday Septem;er 23. 1991 15.00 LL51.RPT Page:

3

4 I"E" UI 1 handay September 23, 1991 15:08 LLSI.RPT Page: G C:\\PVROG\\ GENE 770\\LLSI.CAF CAFTA Fault Tree Report 9-03-91 1:43 Paq o

6 Basic Event Descriptions:

Basic Event hare Prob.

Description CCFSENS 2.00E-05 CCF OF SENSOR Div1P 2.00E-03 DIV I DC POVER FAILS DIV1R 4.75E-05 DIV ! LLS RESET FAILS OlV2P 2.00E-03 Div 11 DC POVER FAILS O!V2R 4.75E-06 OlV !! LLS RESET FAILS F013B 1.31E-03 LLS VALVE F013B FAILS F013D 1.31E-03 LLS VALVE F013D FAILS F013F 1.31E-03 LLS VALVE F013F FA!LS F013G 1.31E-03 LLS VALVE F013G FAILS K14A 1.46E-04 RELAY K14A FAILS OPEN K140 1.46E-04 RELAY K140 FAILS OPEN

$120A 1.30E-05 N120A TRANSMITTER FAILS M1208 1.30E-03 N120B TRANSMITTER FAILS N12DC 1.30E-05 N12DC TRANSMITTER FA!LS N1200 1.30E-05 N1200 TRANSMITTER FAILS k122A 1.30E-05 N122A TRANSMITTER FA!LS N122B 1.30E-05 N122B TRANSMITTER FAILS N122C 1.30E-05 N1220 TRANSMITTER FAIL $

N122D 1.30E-05 61220 TRANSMITTER FAILS N301A 7.30E-05 N301A PRESSURE SVITCH FAILS OPEN h301C1 7.30E-05 N30101 FRES$URE SWITCH FAILS OPEN N302A1 7.30E-05 N302Al PRESSURE SWITCH FAILS OPEN N3020 7.30E-05 N302C PPESSURE SWITCH FAILS OPEN NE20A 7.30E-03 NS20A KASTER TRIP UNIT FAILS hS208 7.30E-03 N6208 MASTER TRIP UNIT FAILS N620C 7.30E-03 NE20C MASTER TRIP UNIT FAILS

~

N5200 7.30E-03 N6200 MASTER TRIP UNIT FAILS NE21A 7.30E-03 hS21A SLAVE TRIP UNIT FAILS N621B 7.30E-03 N621B SLAVE TRIP UNIT FAILS N621C 7.30E-03 NR21C SLAVE TRIP UNIT FAILS N621D 7.30E-03 N5210 SLAVE TRIP UNIT FAILS N622A 7.30E-03 NS224 MASTER TRIP UNIT FAl!S NS22B 7.30E-03 N5228 HASTER TRIP UNIT FAILS NB22C 7.30E-03 NE22C MASTER TRIP UNIT FAILS h622D 7 30E-03 NS220 MASTER TRIP UNIT FAILS RK313A 1.45E-04 RELAY K313a. FAILS OPEN RK3133 1.4SE-04 RELAY K313B FA!LS OPEN RK314A 1.46E-04 RELAY K314A RAILS OPEN RK3148 1.46E-04 RELAY K314B FAILS C:EN RK340A

1. 46E-04 RELAY K340A FAILS OPEN RK3408 1.4EE-04 RELAY K3408 FAILS OPLN RK341A 1,46E-04 RELAY K341A FAILS OPEN RK341B 1.46E-04 RELAY K341B FAILS OPEN RK342A 1.46E-04 RELAY K342A FAILS OPEN RK342B 1.45E-04 RELAY K342P FAILS OPEN Wonday September 23, 1931 15:08 LLSI.RPT Page:

6

4 0

Monday September 23, 1991 15:08 List.RPT Page 13 of 13 p,,;

7 C:\\BVROG\\ GENE 770\\LLS1.CAF CAFTA Fault Tree Report s-03-91 1:43 Peg e

7 Baste Event Descriptions:

Basic Event Name Prob.

Des;rtption R G70A 1.46E-D4 RtLAY K370A FAIL 5 OPIN RO708 1.46E 04 RELAY G 70B FAILS OPit RG71A 1.46E-04 CELAY G 71A FAILS OPEN RG718 1.45E-04 RELAY 0716 FAILS OPEN RO72A 1.4EE-04 RELAY G72A FAILS OPEN RG728 1.4BE-04 RELAY G 72B FA!LS OPEb Monday September 3, 1991 15.08 LLSI.RPT Page:

7

l 4

O ENCLOSURE 5 PAGE 1 OF 2 QUESTION 5 GRAND GULF LIE >/ RELIEF LOGIC i

o TOTAL OF 20 S/RVs o

6 OF THE S/RVs PERFORM A LLS FUNCTION LLS FOUCTION j

SEAL-IN LOGIC LLS LOGIC LOGIC DIV 1 DIV 2 DIV 1 DIV 2 i

LLS VALVES EACH DIVISION IlJs ILig IEE 23Jfg F047D,G 1 OUT OF 3 TWICE N668A N668B N618A N618B F051A,F SEAL-IN IOGIC N669A N669B N618E N618F N670A N670B 2 OUT OF 2 LLS LOGIC N668E N668F N669E N669F N670E N670F F051D 1 OUT OF 3 TWICE N668A N668B N616E N616F SEAL-IN LOGIC N669A N669B N670A N670B 1 OUT OF 1 LLS LOGIC N668E N66BF N669E N669F N670E N670F F051B 1 OUT OF 3 TWICE N668A N668B N617A N617B SEAL-IN LOGIC N669A N669B N670A N670B 1 OUT OF 1 LLS LOGIC N668E N668F N669E N669F N670E N670F

4

?.

ENCLOSURE 5 PAGE 2 OF 2 QUESTION 5 CRAND GULF BWR 6 LOW-LOW SET FUNCTION UNAVAILABILITY FAILURE RATE FAILURES /HR.

PRESSURE TRIP UNIT 2.00E-05 RELAY 4.00E-07

=============================

FAULT TREE INPUTS Failure Rate STI (HOURS)

UNAVAIL.

LLS LOGIC CRANNEL 2.04E-05 730 7.45E-03 (TRIP UNIT + RELAY) 2190 2.23E-02 2.00E-03 DIVISION POWER 2.00E-03 SUPPLY 2.00E-05 CCF OF SENSORS 2.00E-05 LLSL UNAVAILABILITY I MONTH 3 MONTHS CHANGE 1 LOGIC GROUP 2.40E-05 2.40E-05 4.48E-08 1 OUT OF 6 LLS VALVES 2.40E-05 2.40E-05 NEGLIGIBLE (1 OUT OF 3 LOGIC GROUPS)

4 ENCICSURE 6 PAGE 1 OF 3 QUESTION 6 GRAND GULF LLS/ RELIEF LOGIC o

TOTAL OF 20 S/RVs o

6 S/RVs HAVE LLS FUNCTION NO. IN Rx PRESSURE S/RV LOGIC GROU2 GROUP TRIP UNITS LOGIC F041 A,B,C,D,E,F,G,K 9

N670 A,B,E,F 2 OUT OF 2/DIV.

F051 C 1 OUT OF 2 DIV.

F047 A,C,H,L 5

N669 A,B,E,F 2 OUT OF 2/DIV.

~

F051 K 1 OUT OF 2 DIV.

F047 D,G 4

N669 A,B,E,F 2 OUT OF 2/DIV.

F051 A,F OR 1 OUT OF 2 DIV.

LLSL LLSL (SEE ENCL. 5)

(SEE ENCL. 5)

F051 B 1

N669 A,B,E,F 2 OUT OF 2/DIV.

OR 1 OUT OF 2 DIV.

LLSL LLSL (SEE ENCL. 5)

(SEE ENCL. 5)

F051 D 1

N668 A,B,E,F 2 OUT OF 2/DIV.

OR 1 OUT OF 2 DIV.

LLSL LLSL (SEE ENCL. 5)

(SEE ENCL. 5) 20 NOTE:

TRIP UNITS N668 A,B,E,F PROVIDE THE SEAL-IN N669 A,B,E,F FUNCTION FOR THE LLSL.

N670 A,B,E,F 13 OUT OF 20 S/RVs ARE REQUIRED TO PREVENT REACTOR OVERPRESSURE.

ASME CODE ALIDWS 1/2 TO OPEN ON RELIEF MODE (I.E., AT LEAST 7 S/RVs) AND 1/2 TO OPEN ON THE SAFETY MODE (I.E., AT LEAST 6 S/RV).

S ENCLOSURE 6 PAGE 2 OF 3 QUESTION 6 GRAND GULF LLS/ RELIEF LOGIC FAILURE OF S/RV RELIEF &

SAFETY MODE m

FAILURE OF RELIEF FAILURE OF RELIEF FAILURE OF LOGIC GROUP 1 LOGIC GROUP 2 BOTH RELIEF

& SAFETY MODE

& SAFETY MODE LOGIC GROUPS O

b b

i RELIEF LOGIC RELIEF LOGIC RELIEF LOGIC GROUP 1 GROUP 2 GROUP 1 FAILURE FAILURE FAILURE l

[1]

[2]

[1]

FAILURE OF FAILURE OF RELIEF LOGIC SAFETY SAFETY GROUP 2 MODE MODE FAILURE l

[3]

[3]

[2]

[1] RELIEF LOGIC GROUP 1 FAILURE (9 S/RVs AFFECTED)

FAILURE OF Rx PRESSURE TRIP UNIT CHANNELS N670AM E N670E 6@ N670B E N670F

[2] RELIEF LOGIC GROUP 2 FAILURE (10 S/RVr AFFECTED)

FAILURE OF Rx PRESSURE TRIP UNIT CHANNELS N669AM E N669E AND N669B E N669F

[3]

FAILURE TO ACTUATE AT LEAST 6 S/RVs IN SAFETY MODE

'4 1-ENCLOSURE 6 PAGE 3 OF 3 QUESTION 6 GRAND GULF LLS/ RELIEF LOGIC FAILURE RATE (FAILURES / HOUR)

DESIGN CURPINT ANALYSIS EXPERIENCE PRESSURE TRIP UNIT 2.00E-05 5.00E-06 RELAY 4.00E-07 4.00E-07

============================_=======

Failure Rate (FR) (TRIP UNIT + RELAY)

PRESSURE TRIP CHANNEL 2.04E-05 5.40E-06 UNAVAILABILITY - S/RV RELIEF FUNCTION SURVEILIANCE INTERVAL 1 MONTH 3 MONTHS CHANGE RELIEF MODE (SINGLE LOGIC GROUP)

DESIGN ANALYSIS FR 3.16E-04 2.68E-03 2.37E-03 CURRENT EXPR. FR 4.07E-05 2.06E-04 1.66E-04 RELIEF & SAFETY MODE DESIGN ANALYSIS FR 1.06E-07 7.24E-06 7.14E-06 CURRENT EXPR. FR 2.47E-09 4.68E-08 4.43E-08 WHERE:

RELIEF MODE UNAVAILABILITY (SINGLE LOGIC GROUP) =

(1/3)*(2*FR*T)*2 + CCF RM(1) = RM(2)

=

CCF =

COMMON CAUSE MISCAL. OF TRIP UNITS 2.00E-05 / DEMAND

=

RELIEF & SAFETY MODE UNAVAILABILITY =

RM (1) *SM (1) + RM (2) *SM(2 ) + RM (1) *RM (2)

SM(1) = SM(2) = UNAVAILABILITY OF S/RV SAFETY MODE ACTUATION

= <1.00E-05: DEMAND

4 QUESTION 7 PLANT SYSTEMS ACTUATION ENCLOSURE 7 A)

BWR 6 CONTAINMENT SPRAY FAIL TO ACTUATE PAGE I 0F 6 Y

CONTAINMENT SPRAY O

i I I

l SYSTEM A SYSTEM B l

FAILS FAILS w

VALVES TIMER A(B)

F024A(B) &

FAILS F028A(B) FAIL O

s l

l l

[

l LOGIC IA(B)

TIMER LOGIC 2A(B)

LOGIC 3A(B)

VALVES LOGIC 4A(B)

FAILS K93A(B)

FAILS FAILS F042A(B) &

FAILS L

FAILS F028A(B) FAIL c

O O

F3 O

l LEVEL TU DW PRE 55.

LEVEL TU

'DW PRESS.

CONT. PRESS.l CONT. PRESS

'DW PRESS.

DW PRESS.

I N69IA(B)

TU N694A(B)

N691E(F) 10 N694E(F)

TU N662A(C)

TU N662B(D)

TU N694A(B)

TU N694E(F)

FAILS FAILS FAI;S FAILS FAILS FAII.S FAILS FAILS i

l m

M J

ENCLOSURE 7 PAGE 2 OF 6 QUESTION 7 PLANT SYSTEMS ACTUATION A)

BWR 6 CONTAINMENT SPRAY FAILURE RATE (FAILURES / HOUR)

DESIGN CURRENT ANALYSIS EXPERIENCE LEVEL / PRESSURE 2.00E-05 5.00E-06 TRIP UNIT RELAY 4.00E-07 4.00E-07

u===========================================

Failure Rate (FR) (TRIP UNIT + RELAY)

LEVEL / PRESSURE 2.04E-05 5.4OE-06 TRIP CHANNEL UNAVAILABILITY - CONTAINMENT SPRAY AUTO INITIATION SURVEILLANCE INTERVAL 1 MONTH 3 MONTHS CHANGE SINGLE TRIP SYSTEM DESIGN ANALYSIS FR 6.29E-03 1.36E-02 7.31E-03 CURRENT EXPR. FR 6.02E-03 1.11E-02 5.11E-03 1 OUT OF 2 TRIP SYSTEMS DESIGN ANALYSIS FR 5.96E-05 2.05E-04 1.45E-04 CURRENT EXPR. FR 5.62E-05 1.44E-04 8.75E-05 WHERE:

SINGLE SYSTEM (TRAIN) UNAVAILABILITY =

2 * (2/3 ) * (FR*T) *2 + CCF +VLV + TIMER SYS(1) = SYS(2)

=

VLV = PROB OF FAILURE TO OPEN ONE VALVE AND TO CLOSE ANOTHER

= 2*(1.6E-06/HR)*(2160 HRS)/2 TIMER = TIMER 'UNAVAIL. = (6.77E-06/HR) * (TEST INTERVAL)/2 FR = LEVEL OR PRESSURE TRIP CHANNEL FAILURE RATE (FAIL./HR)

T = SURVEILLANCE INTERVAL (HOURS)

CCF = COMMON CAUSE MISCAL. OF TRIP UNITS = 2.00E-05 / DEMAND 1 OUT OF 2 SYSTEMS UNAVAILABILITY = SYS(1)*SYS(2) + CCF

s ENCLOSURE 7 PAGE 3 OF 6 QUESTION 7 PLANT SYSTEMS ACTUATION FEEDWATER/ MAIN TURBINE TRIP t

LEVEL 8 EE.I.TSHE.E K6 2 4 A --

N6?4B -------

TWO OUT OF THREE TO TRIP K62 4C --

l l

i i

4 1

1

\\

ENCLOSUhE 7 PAGE 4 OF 6 QUESTION 7 PLANT SYSTEMS ACTUATION FEEDWATER/ MAIN TURBINE TRIP FAILURE RATE (FAILURES / HOUR)

DESIGN CURRENT ANALYSIS EXPERIENCE LEVEL 8 2.00E-05 5.00E-06 TRIP UNIT RELAY 4.00E-a/

4.00E-07

================-===u===============

Failure Rate (FR) (TRIP UNIT + RELAY)

LEVEL 8 2.04E-05 5.40E-06 TRIP CHANNEL UNAVAILABILITY - FEEDWATER/ MAIN TURBINE TRIP ON LEVEL 8 SURVEILMNCE INTERVAL 1 MONTH 3 MONTHS CHANGE 2 OUT OF 3 DESIGN ANA~ LYSIS FR 2.42E-04 2.02E-03 1.77E-03 CURRENT EXPR. FR 3.55E-05 1.60E-04 1.24E-04 WHERE:

FEEDWATER/ MAIN TURBINE LEVEL 8 TRIP UNAVAILABILITY =

(FR*T)*2 + CCF FR = FAILURE RATE (FAILURES /HR)

T = SURVEILLANCE INTERVAL (HRS)

CCF = COMMON CAUSE MISCAL. OF-TRIP UNITS = 2. 00E-05 / DEMAND

a 4

ENCLOSURE 7 PAGE S OF 6 QUESTION 7 PLANT SYSTEMS ACWATION SUPPRESSION POOL MAKEUP INITIATION TRIP UNIT R N691 A(B)

Rx LEVEL 1

- EITHER -

N694 A(B)

DRYWELL PRESSURE HIGH -

BOTH N691 E(F)

Rx LEVEL 1

- EITHER -

N694 E(F)

DRYWELL PRESSURE HIGH --

- EITHER -

LPCS (RHR B/C) MANUAL INITIATION

- BOTH -

N600 A(B)

SUPPRESSION POOL LEVEL - LOW --

N600 C(D)

SUPPRESSION POOL LEVEL - LOW ANY ONE MANUAL INITIATION EITHER N682 A(B)

Rx LEVEL 2

- EITHER N650 A(B)

DRYWELL PRESSURE HIGH -

I MANUAL INITIATION

- BOTH 30 MINUTE -

TIMER N682 D(C)

Rx LEVEL 2

- EITHER --

N650 D(C)

DRYWELL PRESSURE HIGH -

MANUAL INITIATION

' NOTES:

LETTERS IN PARENTHESES REPRESENT TRIP SYSTEM B TRIP SYSTEM A OPENS VALVES F001A AND F002A TRIP SYSTEM B OPENS VALVES F001B AND F002B i

i ENCLOSURE 7 PAGE 6A 0F 6 QUESTION 7 PLANT SYSTEMS ACTUATION SUPPRESSION P00L MAKEUP INITIATION UNAVAILABILITY - SUPPRESSION POOL MAKEUP AUTO INITIATION FAILURE RATE (FAILURES / HOUR)

DESIGN CURRENT ANALYSIS EXPERIENCE LEVEL / PRESSURE h$bbEbb b$bbEb6 TRIP UNIT Failure Rate (FR) (TRIP UNIT + RELAY)

LEVEL / PRESSURE h$bkEbb b$4bEbb 5

TRIP CHANNEL SURVEILLANCE INTERVAL 1 MONTH 3 MONTHS CHANGE LOW SUPPR POOL LEVEL INITIATION - SYSTEM A(B)

DESIGN ANALYSIS FR 3.75E-03 5.52E 03 1.77E-03 CURRENT EXPR. FR 3.54E-03 3.66E-03 1.24E-04 ANTICIPATORY TIME DELAY INITIATION - SYSTEM A(B)

DESIGN ANALYSIS FR 6.14E-03 1.23E-02 6.12E-03 CURRENT EXPR. FR 6.01E-03 1.10E-02 5.02E-03 LOW SUPPR POOL OR ANTICIPATORY TIME DELAv INITIATION. SYSTEM A (B)

DESIGN ANALYSIS FR 3.52E-03

3. 4E-03 1.69E-05 CURRENT EXPR. FR 3.52E-03 3.53E-03 1.01E-06 WHERE:

LOW SUPPR POOL LEVEL SYSTEM A(B) INITIATION UNAVAILABILITY LSPL 2*(I/3)*(FR*T)*2 + (1/3)*(FR*T)*2 + CCF + VLV I.48E-04 +

7.39E-05 +

2.00E-05 +

3.50E 3.75E-03 (1 MONTH) 1.33E-03 +

6.65E-04 +

2.00E-05 +

3.50E 5.52E-03 (3 MONTHS)

VLV - PROB. OF FAILURE TO OPEN TWO VALVES l

2*(1.6E-06/HR)*(2190 HRS)/2 =

3.50E-03 FR = LEVEL OR PRESSURE TRIP CHANNEL FAILURE RATE (FAILURES /HR) l T - SURVEILLANCE INTERVAL (HOURS) l CCF.

COMMON CAUSE MISCAL. OF TRIP UNITS =

2.00E-05 / DEMAND (CONTINUED ON NEXT PAGE) l

4

)

I ENCLOSURE 7 PAGE 6B 0F 6 QUESTION 7 PLANT SYSTEMS ACTUATION SUPPRESSION POOL MAKEUP INITIATION (CONTINUED)

ANTICIPATORY TIME DELAY SYSTEM A(B) INITIATION - ATD 2*(1/3)*(FR*T)*2 + CCF + TIMER + VLV 1.48E-04 +

2.00E-05 +

2.47E-03 +

3.50E 6.14E-03 (1 MONTH) 1.33E-03 +

2.00E-05 +

7.41E-03 +

3.50E 1.23E-02 (3 MONTHS)

TlHER UNAVAIL. - (6.77E-06/HR)*(TEST INTERVAL)/2 - 2.47E-03 (1 MONTH)

- 7.41E-03 (3 HONTHS)

LOW SUPPR FOOL OR ANTICIPATORY TIME DELAY INITIATION - SYSTEM A (B)

(LSPL-VLV-CCF)*(ATD-VLV-CCF)+VLV+CCF -

(

3.75E 3.52E-03 )*( 6.14E 03 - 3.52E-03 )

+

3.52E-03

- 3.52E-03 (1 MONTH)

(

5.52E 3.52E-03 )*(

1.23E 3.52E-03 )

+

3.52E-03

- 3.54E-03 (3 MONTHS)

M

i ENCLOSURE 8 PAGE 1 OF 2 QUESTION 8 i

BWR 4 MCRECS INSTRUMENTATION

_OCA OR MCRECS ACTUATION N691A LEVEL 1

-- EITHER ---

TRIPS N694A DRYWELL PRESSURE -

MCRECS

-- BOTH A

N691C LEVEL 1

- EITHER ---

=

N694C DRYWELL PRESSUFE --

N691B LEVEL 1

-- EITHER ---

TRIPS N694B DRYWELL PRESSURE --

MCRECS

--- BOTH B

N691D LEVEL 1

- EITHER ---

N694D D'RYWELL PRESSURE -

HIGH RADIATION IN STEAM LINE K603A HIGH MSL RADIATION - q TRIPS F--

BOTH MCRECS K603C HIGH MSL RADIATION - J A

K603B HIGH MSL RADIATION ---

TRIPS BOTH MCRECS K603D HIGH MSL RADIATION ---

B MhlN STEAM LINE_E.BEAJs N686A - N687A - N688A - N689A - STEAM LINE FLOW -

TRIPS

- BOTH MCRECS N686C - N687C - N688C - N689C - STEAM LINE FLOW -

A N686B - N687B - N688B - N689B - STEAM LINE FLOW --

TRIPS

- BOTH MCRECS N686D - N687D - N688D - N689D - STEAM LINE FLOW -

B

(

)

ENCLOSURE 8 PAGE 2 OF 2 QUESTION 8 BWR 4 MCRECS INSTRUMENTATION HIGH RADIATION / CHLORINE LEVEL E002A REFUELING FLOOR AREA RAD.

TRIPS MCRECS

]

R615A CONTROL ROOM AIR INLET RAD.

---

ANY ONE A

NO22A CONTROL ROOM AIR INLET CHIDRINE --

K002D REFUELING FLOOR AREA RAD.

TRIPS MCRECS R615B CONTROL ROOM AIR INLET RAD.

---

ANY ONE A

NO22B CONTROL ROOM AIR INLET CHLORINE --

g l

l ENCLOSURE 9 PAGE 1 OF 2 QUESTION 9 SECONDARY CONTAINMENT ISOLATION ACTUATION N650B DRYWELL PRESSURE HIGH --

- ANY ONE ---

N682B RPV LEVEL 2 TRIP MANUAL INITIATION ---

SYSTEM

-- BOTH 2

N650C DRYWELL PRESSURE HIGH --

-- ANY ONE --

N682C RPV LEVEL 2 MANUAL INITIATION -----

N650A DRYWELL PRESSURE HIGH --

- ANY ONE ---

N682A RPV LEVEL 2 TRIP MANUAL INITIATION -----

SYSTEM

-- BOTH 1

N650D DRYWELL PRESSURE HIGH -

N682D RPV LEVEL 2

--- - ---- ANY ONE ---

MANUAL INITI ATION ----

T. RIPS PERFOPMEt dY EACH SYSTEM o

CLOSES CONTAINMENT COOLING SYSTEM DAMPERS o

CLOSES AUX. BLDG. FUEL HANDLING AREA VENT. SYS. DAMPERS o

CLOSES AUX. BLDG. VENT. SYS. DAMPERS o

STARTS STANDBY GAS TREATMENT SYSTEM o

CLOSES CONTROL ROOM VENT. SYS. DAMPERS o

STARTS CONTROL ROOM EMERGENCY FILTRATION SYSTEM o

CLOSES SECONDARY CONTAINMENT ISOLATION VALVES (SEVERAL)

EACH TRIP SYSfEM ACTUATES EITHER AN INBOARD OR OUTBOARD ISOLATION VALVE / DAMPER OR EITHER ONE OF REDUNDANT SYSTEMS.

u.

_-__mm.-__.__-_

e ENCLOSURE 9 PAGE 2 OF 2 QUESTION 9 SECONDARY CONTAINMENT ISOLATIOR K617B FUEL HANDLING ARIA VENT. ---

EXH. RAD. - HIGH

-- BOTH - -

TRIPS K617C FUEL HANDLING AREA VENT.

3 SYSTEM EXH. RAD. - HIGH

-- EITHER 2

K618B FUEL HANDLING AREA POOL SWEEP EXH. RAD. - HIGH

-- BOTH ---

K618C FUEL HANDLING AREA POOL SWEEP EXH. RAD. - HIGH K617A FUEL HANDLING AREA VENT.

EXH. RAD. - HIGH

-- BOTH ---

TRIPS K617D FUEL HArIDLING AREA VENT.

SYSTEM EXH. RAD. - HIGH

-- EITHER 1

K618A FUEL HANDLING AREA POOL SWEEP EXH. RAD. - HIGH

-- BOTH ---

K618D FUEL HANDLING AREA POOL ---

SWEEP EXH. RAD. - HIGH CONTROL ROOM FRESH AIR INSTRUMENTATION K621B CONTROL ROOH VENT. RAD. - HIGH ---

TRIPS

--- BOTH SYSTEM K621C CONTROL ROOM VENT. RAD. - HIGH ---

2 K621A CONTROL ROOM VENT. RAD. - HIGH ---

TRIPS

--- BOTH SYSTEM K621D CONTROL ROOM VENT. RAD. - HIGH ---

1

= e..-n_*,

._+~ ~

.m

.m

-.,a--

y,

3 h

?. i';

ENCLOSURE 2 k

e

- 8 S

/

, - PROPOSED MODIFICAT]DNS TO THE BWR ACTUATION INSTRUMENTATION

- STANDARD TECHNICAL SPECIFICATION $

.g" e

6 9 9

[

l' r

1

a INSTRUMENTATION 3/4.3.9 PLANT SYSTEMS ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.9 The plant systems actuation instrumentation channels shown in Table 3.3.9-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.9-2.

MPLICABILI_TY:

As shown in Table 3.3.9-1.

ACTION:

a.

With a plant system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column cf Table 3.3.9-2, declare the channel inoperable and either place the inoperable channel in thL tripped condition until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip setpoint value, or declare the associated system inoperable, b.

For the suppression pool (and drywell) spray system:

1.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place at least one inoperable channel in the tripped condition within one hour or declare the afsociated system inoperable.

2.

With the number of OPERABLE channels less than recuired by the Minimum 0PERABLE Channels per h ip 9/ stem requirement f or both trip systems, declare the an xim system inoperable.

c.

For the feedwater system / main tur n e t-ip systeu:

1.

With the number of OPERABLE chaanels one 1ers the -equired by the Minimum OPERABLE Channels requirement, restore 6he inoperable channel to OPERABLE status within 7 days or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2.

With the number of OPERABLE channels two less than rt red by the Minimum OPERABLE Channels requirement, restore at east one of the inoperable channels to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

l l

GE-STS (BWR/4) 3/4 3-89 i

~

t D

INSTRUMENTATION SURVEILLANCE REQUIREMENTS i

4.3.9.1 E6th plant jem actuation instrumentation channel shall be demonstrated 09ERAB

) the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and C i NNEL CALIBRATION operations f <

the OPERATIONAL CONDITIONS and at the frequencies shown ir. Table 4.3.9.1-1.

4.3.9.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

,cmmmW~vWK

% A CHANNEL IAAY BE PLAGD lN AN INOPEPABLE S~INTOS FOR UP TO 6 HOURS FOR REQU1 RED SURNE\\LLANCE W\\THOUT PLAUNG7HE TR\\P S)'STEtA lN THE TR\\PPED COND n\\DM myy xv' us GE-STS (BWR/4) 3/4 3-90

e TABLE 3.3.9-1 PLANT SYSTEMS ACTUATION INSTRUMENTATION APPLICABLE OPERATIONAL 9

CONDITIONS k

TRIP FUNCTION MINIMUM O

OPERABLE CHANNELS PER TRIP SYSTEM l

1.

SUPPRESSION POOL (AND GEYWELL) SPRAY SYSTEM 1

1, 2, 3 Drywell Pressure-High a.

1 1,2,3 l

b.

Containment Pool Pressure-High Reactor Vessel Water Level - Low Low Low, level 1 1

1, 2, 3 c.

d.

Timers 1

1, 2, 2 1)

System A 1

2, 3 1

{

2)

System B w

MINIMUM OPERABLE CHANNELS 2.

FEEDWATER SYSTEM / MAIN TURBINE TRIP SYSTEM I

Reactor Vessel Water Level-High, level (8) a.

t

e M

TABLE 3.3.9-2 4

y

[

PLANT SYSTEMS ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT val.UE 3,,

i 1.

SUPPRESSION POOL (AND DRYWELL) SPRAY SYSTEM 3

a.

Drywell Pressure-High

$ (1.69) psig

$ (1.89) psig b.

Containment Pressure-High 5 (35) psig.

$(

) psig c.

Reactor Vessel Water Level - Low Low Low, Level 1 1 -(

) psig 3 -(

) psig j

d.

Timers 4

1)

System A

$ (12) minutes 5 (13.2) minutes 2)

System B

$ (14) minutes 5 (15.4) minutes 2.

FEEDWATER SYSTEM / MAIN TURBINE TRIP SYSTEM I

a.

Reactor Vessel Water Level-High, Level (8)

$ (54.5) inches *

$ (56.0) inches Y$

^See Bases Figure B 3/4 3-1.

a

]

ci7 TABLE 4.3.9.1-1 (Centinued)

[

PLANT SYSTEMS ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS

'k CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 1.

SUPPRESSION POOL (AND ORYWELL) SPRAY SYSTEM a.

Drywell Pressure-High (NA)

-tMt &

(Q) 1, 2, 3 b.

Containment Pressure-High (NA)

-(M} Q (Q) 1, 2, 3 c.

Reactor Vessel Water Level-Low Low Low, level 1 (NA)

-(M ) C L (Q) 1, 2, 3 d.

Timers (NA) fMt G }

(R) 1, 2, 3 2.

FEEDWATER SYSTEM / MAIN TURBINE TRIP SYSTEM a.

Reactor Vessel Water level-High, (NA)

(M) G.

(R) 1 Level (8) w 8

4 i

il

o e

\\,

INSTRUMENTATION END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION

)

LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle recirculation pump trip (EOC-RPT) system instrumentation channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table.'.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3.3.4.2-3.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to (30)% of RATED THERMAL POWER.

ACTION:

a.

With an end-of-cycle recirculation pump trip system instrumentation channel trip sotpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value, b.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel (s) in the tripped condition within c.

With the number of OPERABLE channels two or more less than raquired by the Minimum OPERABLE Channels per Trip System requirement for one trip system and:

1.

If the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel lace both inoperable channels in the tripped condition withi l

12. BOUR$

2.

If the inoperable channels include two turbine contror va ve channels or two turbine stop valve channels, declare the trip r

system inoperable, d.

With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or (take the ACTION required by l

Specification 3.2.3) (reduce THERMAL POWER to less than (30)% of RATED THERMAL POWER within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />).

With both trip systems inoperable, restore at least one trip system e.

l to OPERABLE status within one hour or (take the ACTION required by l

Specification 3.2.3) (reduce THERMAL POWER to less than (30}% of RATED THERMAL POWER within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />).

i l

GE-STS (BWP/4) 3/4 3-40

• .z s

INSTRUMENTATION I

SURVEILLANCE REQUIREMENTS 4.3.4.2.1 Each end-of-cycle recirculation pump trip tyrtem instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations et the frequencies shown in Table 4.3.4.2.1-1.

4.3.4.2.2.

LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

4.3.4.2.3 The END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME of each trip function shown in Table 3.3.4.2-3 shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least the logic of one type of channel input, turbine control valve f3st closure or turbine stop valve closure, such that both types of channel inputs are tested at least once per 36 months.

(The time allotted for breaker arc suppression,

(

) ms, shall be verified by test at least once per 60 months.)

L 0

l I

i l

l l

GE-STS (BWR/4) 3/4 3-41

-, - - -. ~.

i S4 TABLE 3.3.4.2-1

-4

)[

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION eR d3 MINIMUM OPERABLECHANNE(g)

TRIP FUNCTION PER TRIP SYSTEM 5}

1.

Turbine Stop Valve - Closure 2

I}

2.

Turbine Control Valve-Fast Closure 2

R' kb

[

(* A trip system may be placed in an inoperable status for up tov @=

ours for required surveillance provided j,

that the other trip system is OPERABLE.

I This function shall be automatically bypassed when turbine first stage pressure is less than or equal to (

) psig, equivalent to THERMAL POWER less than (30)% of RATED THERMAL POWER.

i a

O TABLE 3.3.4.2-2 in

7 END-OF-CYCLE RECIRCULATION PUMP TRIP SETPOINTS m

!j ALLOWABLE q

TRIP FUNCTION TRIP SETPOINT VALUE

=

1.

Turbine Stop Valve-Closure 1 (5)% closed

$ (7)% closed 2.

Turbine Control Valve-Fast Closure

> (500) psig

> (414) psig iR.

' O 4

h

c TABLE 3.3.4.2-3 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESP 0ASE TIME

[

TRIP FUNCTION RESPONSETIME(MillesecondM Ev 1.

Turbine Stop Valve-Closure 1 (100)

O 2.

Turbine Control Valve-Fast Closurc 5 (100)

Y t

~'

b-

.JN i

TABLE 4.3.4.2.1-1 c)

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM SURVEILLANCE REQUIREMENTS CHANNEL j[

FUNCTIONAL CHANNEL gr TEST CALIBRATION ad TRIP FUNCTION

25*} C2 R

1.

(urbine Stop Valve-Closure

2b ) (3[

(R) 2.

Turbine Control Valve-Fast Closure

(^ Including trip systtm logic testing.)

T.

O t

l l

ti

o.

6

{

3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION-

-ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION d

3.3.4.1 The anticipated transient without scram recirculation pump trip (ATWS-RPT) system instrumentation channels shown in Table 3.3.4.1-1 shall be OPERABLE with their trip setpoints set consistent with values shown in the Trip Setpoint column of Table 3.3.4.1-2.

APPLICABILITY:

OPERATIONAL CONCITION 1.

ACTION:

a.

-With an ATWS-RPT system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.1-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.

b.

With the number of OPERABLE channels one less than required by t!.;

Minimum OPERABLE Channels per Trip System requirement for one or both Trip System (s), restore the inoperable channel (s) to OPERABLE status within 14 days or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c.

With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE channels per Trip System requirement for one-trip system and:

l 1.

If the inoperable channels consist of one reactor vessel water level channel and one reactor vessel pressure channel, place l

both inoperable channels in the tripped condition

• within $

or declare the trip system inoperable.

24 HOUR S 2.

If the inoperable channels include two reactor vesse water level channels or two reactor vessel pressure channels, declare the trip system inoperable.

d.

With one trip system inoperable, restore the inoperaole trip system l

to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least STARTUP within l

the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

e.

With both trip systems inoperable, restore et least one trip system to OPERABLE status within one hour or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

• The inoperable channels need not be placed in the tripped condition where this would cause the Trip Function to occur.

GRAND GULF-UNIT I 3/4 3-37 Amendment No. 41 r

l

- ~ - -

d, i

l' 1

INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.4.1.1 Each ATWS recirculation pump trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHr WEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies sho.n in Table 4.3.4.1-1.

4.3.4.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

1 l

i 1

GRAND GULF-UNIT 1 3/4 3-37a Amendment No. 41

4.

INSTRUMENTATION TABLE 3.3.4.1-1 ATVS REClRCULATION PUMP TRIP SYSTEM INSTRUMENTATION MINIMUMOPERABLECHAN((g5PER 1 Rip rVNCTION TRIP SYSTEM 1.

Rea: tor Vessel Water Level -

Lo= Low, level 2 2

2.

Reactor Vessel Pressure - High 2

6 HOURS (a) One channel may be placed in an inoperable status for up to 0 h*wu+ for required surveillance provided the other thannel is OPERABL m

GRAND CULF-UNIT 1 3/4 3-38

d.

4 T_ABLE 3.3.4.1-2 ATW5 RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS TRIP ALLOWABLE TRIP FUNCTION SETPOINT VALUE 1.

Reactor Vessel Water Level -

-> -41.6 inches *

-> -43.8 inches Low Low, Level 2 2.

Reactor Vessel Pressure - High 1 1095 psig

< 1102 psig

'See Bases figure B3/4 3-1.

I l

(

l' i'

l GRAND GULF-UNIT 1 3/4 3-39 Amendment No. 41 DEC 3 01.

,.-,.+,,.~.,.,.,,,__,,,,,_,_,,,,..._____m._,,....._,_.___..._,,,,#,.,m.

.,.,... ~,,,

,.,-.._,.,.._,~.-.m._...

= _.

It o-INSTRUMENTAV10N TABLE 4.3.4.1-1 ATW5 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL TRIP FUNCTION CHECK TEST CAL 1 BRAT 10N 1.

Reactor Vessel Water Level -

5

• G.

R*

Low Low, level 2 2.

Reactor vessel Pressure - High 5

-M: G.

R*

' Calibrate trip unit at least once per days.

92

)

GRAND GULF-UNIT 1 3/4 3-40

!i e

j 1

REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES SAFETY / RELIEF VALVE 5 LIMITING CONDITION FOR OPERATION 3.4.2.1 (At least (two) reactor coolant system code safety valves and) the safety valve function of at least (11) (of the following) reactor coolant system safety / relief valves shall be OPERABLE with the specified code safety valve function lift settings:*

(2) safety valves @ (1146) psig 11%

(3) safety-relief valves @ (1175) psig +1%

(3) safety-relief valves @ (1185) osig T1%

(3) safety-relief valves @ (1195) psig 71%

(2) safety-reliefvalves@(1205)psig{1%

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a._

With (one or more of the above required reactor coolant systen code safety valves or with) the safety valve function of one or more of the above reauired safety / relief valves inoperable, be in at least HOT SHUT 00kW within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTOOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

With one or more (code safety valves or) safety / relief valves stuck open, provided that suppression pool average water temperature is less than (95)*F, close the stuck open (code safety valves and/or) safety relief valvels);

if unable to close the stuck open valve (s) within 2 minutes or if sup-pression pool average water temperature is (95)"F or greater, place the reactor mode switch in the Shutdown position.

c.

With one or more safety / relief valve (tail pipe pressure switches)

(acoustic monitors) inoperable, restore the inoperable (switch (es))

-(monitor (s)) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

GE-STS (BWR/4) 3/4 4-5

~,

~

O 4

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (4.4.2.1.1 (The code safety valve function of each of the above required safety /

relief valves shall be demonstrated OPERABLE by verifying that the bellcws on the stiety/ relief valves have integrity, by instrumentation indication, at-least i

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

4.4.2.1.2 The (tail pipe pressure switch) (acoustic monitor) for each safety /

relief valve shall be demonstrated OPERABL with the setpoint er fjed to be-

/,

((20) 1 (5) psig) by performance of a:

4y a.

CHANNEL (FUNCTIONAL TEST) (CHECK) at least once per 3* days, and a b.

CHANNEL CALIBPATION at least once per 18 months (*).

(^The provisions cf Specification 4.0.4 are not applicable provided the Surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.)

e A CRAMNRMAY BE PLALED IN AN INOPEPABW STATUS FOR b? 10 (o HOVR5 FOR RE(lu\\RfD SVR\\E\\tLANCE NNGAOVT RPUE TRUTRW 5YSTEN\\ IN 7RE ~ TOPPED CONDU \\QN.

4 v-L 3/44-6 GE-STS (BWR/4)

O e

REACTOR COOLANT SYSTEM SAFETY / RELIEF VALVES LOW-LOW SET FUNCTION LIMITING CONDITION FOR OPERATION 3.4.2,2 The relief valve function and the low-low set function of the following reactor coolant system safety / relief valves Aall be OPERABLE with the following settings:

Low-Low set Function Relief Function Setpoint* (psig) i 1%

Setpoint* (psi,g) 1%

Valve No.

Open Close Open Close (1033)

(926)

(1073)

(936)

(1113)

(946)

(1113)

(946)

~

(1113)

(946)

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a.

With the relief valve function and/or the low-low set function of one of the above required reactor coolant system safety / relief valves inoperable, restore the inoperable relier valve function and low-low set function to OPERABLE status within 14 6ays or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in CCLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

With the relief valve function and/or the low-low set function of more than one of the above required reactor coolant system safety /relie M alves inoperable, be in at least HOT SHUTOOWN witnin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLO SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.2,2.1 The relief valve function and the low-low set function pressure actuation instrumentation shall be demonstrated OPERABLE b, performance of a:

x.y.

a.

CHANNEL FUNCTIONAL TEST, including calibration of the r p unit, at least once per b.

CHANNEL CALIBRATION, LOGIC SYSTEM FUNCTIONAL TEST and simulated automatic

~

operation of the entire system l

4 W A CRANNEL MAY BE PLATED IN AN INOPECLABLE 5'IATVL FOR VP 70 6 HOVRS TOR REGV\\TLED quRVERu;tsitE WfTHOVT PLAC)NG 1RE "m.i? SYSTEtA itt TRE 7R\\PPED LOND) TION.

w

^The lif t setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

GE-STS (BWR/4) 3/4 4-7

_~.

=.

REACTOR C00LAN7 SYSTEM 3/4.4.2 SAFETY VALVES SAfrTY/ RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2.1 Of the following safety / relief valves, the safety valve function of at least (6) valves and the relief valve function of at least (5) valves other than those satisfying the safety valve function requirement shall be OPERABLE with the specified lift settings:

Number of Valves Function Setpoint* (psig) i 1%

(7)

(Safety)

(1165)

(5)

(Safety)

(1180)

(4)

(Eafety)

(1190)

(1)

(Relief)

(1103)

(8)

(Relief)

(1113)

(7)

(Relief)

(1123)

APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a.

With the safety and/or relief valve function of one or more of the above required safety / relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

With one or more safety / relief valves stuck open, provided that suppression pool average water temperature is less than (105) F, close the stuck open tafety/ relief valve (s); if unable to close the open valve (s) within 2 minutes or if suppression pool average water temperature is (105) r or greater, place the reactor mode switch in the Shutdown position.

c.

With one or more safety / relief valve (tail pipe pressure :;wi6ches) (acoustic monitors) inoperable, restore the inoperable (switch (es)) (monitor (s)) to OPERABLE status within 7 days or be in at least HOT SHUT 00WN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.2.1.1 The (tail pipe pressure switch). acoustic monitor) for each safety /

relief valve shall be demonatrated OPERABLE with the setpoi gerified to be

((20) t (5) psig) (

) by performance of a:

2.

CHANNEL (FUNCTIONist TEST) (CHECK) at least once pn days, and a a.

CHANNEL CAL g 0N at least once per 18 months.

b.

4.4.2.1.2 The relief e function pressure actuation instrui.ientation shall be demonstrated OPERABL by performance of a:

a.

CHANNEL,FUNCTI L TEST

'luding calibration of the trip unit, at least once per et ays. 92 b.

CHANNEL CALIBRAT ON, LOG CNlYSTEM FUNCTIONAL TEST and simulated automatic operation of the entire system at least once per 18 months.

The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.

The provisions of Specification 4.0.4 are not applicable provided the surveillance is-performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is MXX Cu Ot01\\.

y6 p 6

wouru von RecttwREo suRVER.GNCE WGBCATT PLNJE *THE WJP 4

s o

REACTOR COOLANT SYSTEM SAH.TY/ RELIEF VALVES LOW-LOW SET FUNCTION LIM: TIN CONDITION FOR OPERATION 3.4.2.2 The relief valve function and the icw-low set function of the following reactor coolant system safety / relief valves shall be OPERABLE with the following settings:

Low-Low Set function Relief Function Setpoint* (psig) i 1%

Setpoint* (psig) 1%

Valve No.

Open Close Open Close (1033)

(926)

(1073)

(936)

(1113)

(946)

(1113)

(946)

~

~

(1113)

(946)

APPLICABILITY:

OPERATIONAL CONDITICNS 1, 2 and 3.

ACTION:

a.

With the relief valve function and/or the low-low set function of one of the above required reactor coolant system safety / relief valves inoperable, restore the inoperable relief valve function and the lor low set function to OPERABLE status within 14 days or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTOOWN within the folicwing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

With the relief valve function and/or the low-low set function of more than one of the above required reactor coolant systM safety / relief valves inoperable, be in at least HOT SHUTDCWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4 SURVEILLANCE REQUIREMENTS 4.4.2.2.1 The relief valve function and the low-low set function pressure actuation instrumentation shall be demonstrated OPERABL formance of a:

a.

CHANNEL F NCTIONAL TEST, including calibration of the trip unit, at least once pe 3

s.

.9A b.

CHANNEL CALI~BRATION, LOGIC SYSTEM FUNCTIONAL TEST and simulated automatic operation of the entire system at least onc per 18 months.

"The lift setting pretsure shall correspond to ambient conditions of the valves j

at nominal operatin temperature s

l WM A CRANNEL /AAY BE PLALED IN AN INOPERABLE STATUS FOR UP TO l

6 HOVR5 FOR REQV\\ RED SURVEtL\\.ANLE WITHOUT PLAC!NG'THE TRIP SYSTEtA iM ~TRY. TR.iPPED COND\\~T\\ON

' ^

• GE-5Td (u

c ILSTRUMENTATION

.s/4.3.8 PLANT SYSTEMS ACTUATION INSTRUMENTATION ESTRUMENTATION 3/4.3.8 PLANT SYSTEM 3 ACTUATION INSTRUMENTATION gIMITINGCONDITIONFOROPERATION i

3.3.8 The plant systems actuation instrumentation channels shown in Table i

3.3.8-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.8-2.

APPLICABILITY:

As shown in Table 3.3.8-1.

ACTION:

a.

With a plant system actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.8-2, declare the channel inoperable and take the ACTION required by Table 3.3.8-1.

b.

With one or more plant systems actuation instrument channels in-operable, take the ACTION required by Table 3.3.8-1.

SURVEILLANCE REQUIREMENTS 4.3.8.1 Er.ch plant system actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.8.1-1, 4.3.8.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

l l

GRAND 6 TAI-UNIT 1 3/4 3-92

'M

TABLE 3.3.8-1 m

5?g PLANT SYSTIMS ACTUATION INSTRUMENTATION

' E-MINIMUM APPLICABLE 5

OPERABLE CHANN OPERATIONAL PERTRIPSYSTEMg g

TRIP FUNCTION CONDITIONS ACTION Z

1.

CONTAINMENT SPRAY SYSTEF g

a.

Drywell Pressure-High 2

1,2,3 130 b.

Containment Pressure-High 1

1,2,3 131 c.

Reactor Vessel Water Level-Lew Low Low, level 1 2

1,2,3 130 d.

Timers

1) ~ Systan A 1

1,2,3 131

2) System B 1

1,2,3 131 j

2.

FEEDWATER SYSTEM / MAIN TURBINE TRIP SYSTEM w

a.

Reactor Vessel Water Level-High, level 8 3

1 132 3.

SUPPRESSION POOL MAKEUP SYSTEM a.

~rywell Pressure - High (ECCS) 2 1,2,3 135 b.

Drywell Pressure - High (RPS) 2 1,2,3 135 c.

Reactor Vessel Water Level - Low Low Low, Level 1 2

1,2,3 135 d.

Reactor Vessel Water Level - Low Low, Level 2 2

1, 2, 3 135 e.

Suppression Pool Water Level - Low Low 1

1,2,3 133 f.

Suppression Pool Makeup limer 1

1,2,3 133 g.

SPMU Manual Initiation 2

1,2,3 134 rHOUR o

(a)A channel may be placed in an inoperable status for un to 2 ' = = during periods of required surveillance provided at least one other OPERABLE channel n e same trip system is monitoring that parameter.

s e

I TABLE 3.3.8-1 (Continued)

PLANT SYSTEMS ACTUATION INSTRUMENTATION ACTION ACTION 130 -

a.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement, laceAeinoperablechannelinthetrippedconditionwithin 24 HOURS e = 1 otherwise declare the associated containment spray system inopera,ble and take the action required by Tech-nical Specification 3.6.3.2.

b.

With the number of OPERABLE channels two less than required:

by the Minimum OPERABLE channels per Trip System require-ment, declare the associated containment spray system inoperable and take the action required by Technical Specification 3.6.3.2.

ACTION 131 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement, restore the inoperable channels to OPERABLE status within one hour; other-wise, 6eclare the associated containment spray system inoperable and take the action required by Technical Specification 3.6.3.2.

ACTION 132 -

For the feedwater system / main turbine trip system:

a.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels requirement, restore the inoperable channel to OPERABLE status within 7 days or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With the number of OPERABLE channels two less than reouired by the Minimum OPERABLE Channels per Trip System require-ment, restore at least one of the inoperable channels to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

L ACTION 133 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement, declare the associated suppression pool makeup system inoper e and take the action required by Specification 3.6.3.4.

p ACTION 134 -

With the number of OPERABLE channels less than regtire y the Minimum OPERABLE Channels per Trip System requirer t, restore the inoperable channels tn OPERABLE status withi curs; other-wise, declare the associated suppression pool makeup system inoperable and tcke the action required by Specification 3.6.3.4.

ACTION 135 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

a.

With one channel inoperable, place _the inoperable channel in the tripped condition within n=, HuBor d lar4~t h associated system (s) inoperable.

HOURS b.

With more than one channel inoperable, deci e

associated system (s) inoperable.

GRAND GULF-UNIT 1 3/4 3-94

j TALE 3.3.8-2 x

[

- PLANT SYSTEMS ACTUATID'l INSTRUMENTATION SETPOINTS Cy ALLOWABLE g_ TRIP FUNCTION TRh SETPOINT VAUiE

[

1.

CONTAINMENT SPRAY SYSTEM a.

Drywell Pressure-High

< 1.39 psig

< 1;44 psig b.

Containment Pressure-High i 7.84 psig

-7 8.34 psig c.

Reactor Vessel Water level-Low Low Low,-Level 1

> - 150.3 inches 1 - 152.5 inches d.

Timers

1) System A 10.85 + 0.10 minutes 10.25 - 0.00, + 1.18 minutes

2) System B 10.8510.13 minutes **

10.26 - 9.0G, + 1.18 minutes 2.

FEEDWATER SYSTEM / MAIN TURBINE TRIP SYSTEM wA Reactor' Vessel Wa' r Level-High, level 8 5 53.5 inches

• i S4.1 inches a.

w 3.

SUPPRESSION POOL MAKEUP SYSTEM a.

Drywell Pressure - High (ECCS) i 1.39 psig i 1.44 psig b.

Drywell Pressure - High (RPS) i 1.23 psig i 1.43 psig c.

Reactor Vessel Water level - Low Low Low, Level 1 1 -150.3 inches

• 1-152.5 inches d.

Reactor Vessel Water Level - Low Low, Level 2 3 -41.6 ix oes*

1-43.8 inches e.

Suppression Pool Water Level - Low Low 1 17 ft 5 inches 1 17 ft 2 inches f.

Suppression Pool Makeup Timer i 29.0 minutes 1 29.5 minutes g.

SPMU Manual Initiation NA NA

'*See' Bases Figure B 3/4 3-1.

• fgjnt for Syst'n p'is the sum of E12-K0938 plus E12-K116.

E12-K116 is not to gg 10.00 seconds.

e ii

Q 2W u a; 7

,5 s

<msn-w mm mm m

m m

m m

mm gas m

se 44 ea a

ae ce n, -.

C--

R$.

m.m c

=

E g

-8

.a -

m-E E CC Ct:Cr M

K K

M CaC E

CF m e, mw v.=

g Cu m

J W

i et 60 66 ct 6aa 666 l e

m-

-j !

g,gie H

$# 1 4

4#

t-

s e

4 W

.a.

m mm 3

g gg mm mm m

m m

m m

m mm 5"

~

E c

a a

N e

.=

w

~

s.

E.

v

=

v m

E

.!=

S 2

t W

E

%E 5

3 W.

w w

• ~ W

.E.e W

1. ==

.C

.c W

W

@)

O W

Z f-to m

On J

A L

C k

L m

b b

m.

x

.- u' &. -

5.

L

.J

.,as,

&W-n W

W W

a.

M.

3-E W

Z3e>

f *=

• e 4

8 eh W

' *=

M e en W

. 3M 3

r==

r*

ed m

• s=

e=

>=

~ s e. 3,

X W

W Al to Wm w

E o

)

a-m Db-e==

M 03 3

r= *==

m O

O C fg a

m& W 3 es g

en en W as W

m A e=e

>=

m en G m

en en to >

sn N ea C

4)eh m J an J

W W

en SA to G

C. -

• • =

+

43C L C &#

W b

b W

W 8'*

O O -

C Ch.

A

>3

>W sh.. A

>W

• c=

av=

3 3

m O

th en se C c=

A.

J

>=

L e=

Paa e=

3.g 3'

b b&

to g en -

~%

=

g h

3 g t"N., 8 5 W 3 D;

.3 2 E

~h h W a W a Ra g g t.a e

'c W*

.C, W

m W

W e

• =*

t W kg W

M E E

W Q

Q CE J Rac J to J m m

+a h $

f W

g w o a

E u

4AJ + W5 a

8 4

.d a

d.

41.

- a,1 4k.

2_::

~

p-(*

Nb Ms gue

~ " "

• _ ' -

te.spe GRAND GULF-UNIT 1 3/4 3-96~

?

".s.-

_n.,-

s-. _, -.,,,

,.-+,.w.-.-

,,.n_.,w.,

_em,

.,.,,.a

e INSTRUMENTATION MAIN CONTROL ROOH ENVIRONMENTAL CONTROL SYSTEM (MCRECS) ACTUATION INSTRUMENTATION LIMITING CONDIT'ON FOR OPERATION 3.3.6.7 The MCRhis actuation instrumentation channels shown in Table 3.3.6.7-1 shall be OPERABLE, with their trip setpoints set consistent with tr.a values shown in the Trip Setpoint column of Table 3.3.6.7-2.

APPLICABIQT,1: As shown in Table 3.3.6.7-1.

ACTION:

As shown in Table 3.3.6.7-1.

SURVEILLANCE REQUIREMENTS 4.3.6.7 Each MCRECS actuation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHCCK, CHANNEL FUNCTIONAL TEST, and r"ANNEL e

CALIBRATION operations during the OPERAlIONAL CONDITION and e se frequencies shown in Table 4.3.6.7-1.

HATCH - WIT 2 3/4 3-58 A,nendfrent No. 71

-.e--

~

b:

4 r

. 'i I

1(E t L3.6. 7-1 t satEET 1 Or 21 i' i

-e n

MCRICS AcTuAfl01s leStatstEnfATlost 2

s-MilllMUM SOUMBfR APPt t cAet t c

OrtRAstE CHAsIIIELS OPERAT IO8044.

l' 2

IRif FUlsCTipil PER TRIP SYST[Miellbl COII0f ilost ACTIOtt y

i i.

1.

Reacter vesses weer teves -

2 i '

ro 2821-19691 A, S C, 9 1, 2, 3 52 Low tow' tow (tevei 1) (e)

..j 2.

Drywes t Pressure - 8tigh (c) 2 1,2,3 52 2 Ell-se694 A. S. C, e 3.

Desin Stese Line Radiation - Nigh (c) 2 1, 2, 3.'**

S3 2Dll-R603 A, S C, B-4.

Mein Stees Llne Flow - Nigh (c) 2/llns 1,2,3 53-

+

2821-01646 A, B, C D 2821-II647 A, 3, C O l

. :r ij 2821-80634 A, S,.C, O j

4 2921-88649 A, 3, C, 8 t

5.

Aerueting Tsoor Aree Redistien - Nigh (c) 1 2D21-F002 A, D 1, 2, 3, 5, *

$4 t

't M

2 6.

Control Room Air inlet Radiation - Migli (c) 1, 2, 3, S.

• S4 t

'{

1Z41-R61S A, 8 h

9 f

3 3

i f

{

2 g

l 4

I i

f l

l f

i 1

s l',

I i ;

-l t

4 l'T

.t l

2:

i 0

2 1

j O$

l f ; 't -

ji )

i t

2 I

i i

i a

6m ym TA8t.E 3.3.6.7-1 (SHEET 2 0F 2)

MCRECS. ACTUATION INSTRUMENTATIO4 ACTION ACTION 52 -

Take the ACTION required by Specification 3.3.3.

j ACTION $3: -- Take the ACTION required by Specification 3.' 2.

ACTION 54 -

I I

With one of'the required radiation monitors inoperable, restore the a.

monitor to OPERABLE status within 7 days or, within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, initiate and maintatn-opw4 tion of the MCRECS in the pressurization mode of operation, b.

With no radiation monitors'0pERABLE, within-1 hour initiate and maintain operation of the MCRECS in the pressurization mode of 1

operation, The provisions of Specification 3.0.4 are not applicable.

c.

4

' NOTES-When handling irradiated fuel in secondary containment, 6 HOURS A caannel may be placed in an inoperable status for up t a.

? e -; for required surveillance without placing the trip system in t e tr pped condition,'provided at least one other OPERA 8LE channel in the'same trip system is monitoring that parameter.

b.

With a design _providing only one channel per trip system, an troperable channel need not be placed in the tripped condition where this would-cause the Trip Function to occur. -In these cases, JS1:0N required by Table 3.3.6.7-1 for that Trip Fun.h' neta channel shall be restored to OPERABLE status withi

-%r the

+ -

s-4.akaa.

24 HOURS Actua% ; S; MCRECS in the-control room pressurization c.

d.

(DeleteQ Within-24 hours.rior to the planned start of the hydrogen injection test e.

with the reac'-

ower at greater than 20 per:ent rated-power, the normal full power -

Jasion backgrnund level;and associated trip setpoints may be

ch: aged.

1 on a casculated value of'the radiation level expected during the test.

> backe ound radiation level and associated trip setpoints may. be adh.ced during the test based on either calculations or measurements of actual radiation. levels resulting fros' hydrogen injection. The background radiation level shall-be determined and asso'ciated trip setpoints.shall be set within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of re-establishing normal. radiation levels after completion of hydrogen injectjon and prior to estab' Aing reactor pcwor levelsletow 20 piircent rated power.

HATCH - UNIT 2 3/4 3-58b Amendment No. 96 CEP 1 ' 1968 4

1

+e.

~

~ ~ ~ ~ '

~ *

..e,.~.4 m~

m I'

4

,s b43 2.a.6.7-1 iE

-4 HCRECS ACTt!ATION INSTRUMENTAT1061 SETrolm nZ TRif FUNCTION TRIP SETFOINI

/LLLWABLE VALUE c

z 1.

Reactor Vessel Water Level -

2 -113 inches 2 - 113 inches i.

Q Low Low Low (Level 1)

N

2. 4 Drywell Pressure - High 1 1.92 psig i 1.92 psig 3.

Main Steam Lins Radiation - High 1 3 x rull-power hackground*

5 3 x rull-power bacliground' 4.

Main Steam Line Flow - High 5 1381 rated elv.e 5 138% rated flow 5.

Reruellog floor Area Radiattlon - High 120 mr/ hour 5 20 er/ hour I

6.

Control Room Air inlet i 1 er/ hour 1 1 er/ hour Radiation - High N

l em, n

u e

l W

n l

t 1

l l'.

R m

$ $ *Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the planned start of the hydrogen injection test with the reactor power et greater than 70-percert j c.

rated power, the normal full powr reeletion background level and associated trlp setpoints esy be changed based on a

$

• calculated value of the radiation level expected during the test.

The background radiation level and associated trip F*

to o.etpoints may be adjusted during the test based on olther calculations or measurements or actual radiation levels t'

resulting from hydrogen inJactJon. The bacligrourJ radiation level shall be determined and associated trip sitpoints

$ z shall be set within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> er re-establishing normal radiation levels after completion or hydrogen injecteon and prior e

co o to estebIlshing reactor power levels belev 20-percent rated power.

+

• e t

0 1A8tE 4.3.6.7-1 l

h

-4 11GLECS ACTUATION INSTitut9EMIAI19M SURVEf LLANCE REQUlf LM[H.[$

9 CHAststL OffRAIiOMAL 4

CHANNEL FttNCTICMAL CHAMit[L COMOIIIONS IM M4iOf

.[

c-T R I P (3 rdil.lg s

.. CHECK YE$T Call BRAT I04 5URVEILLAlgEE Rf TIRED i

2 y

1.

Reactor Vessol Water Level -

S

-M-O R

I, 2, 3 Low Low Low (Level 1)

N 2.

Dryweli Pressure - High S

-M-- b R

1, 2, 3 W"'

R 1, 2, 3 8

3,-

Main Steen Line Asdistion - High D

ti.

Main Store Line Flow - High 5

-M-d R

1, 2, 3 5.

Refueling floor Area Radiation -

D

-M8**Q q

1, 2, 3, 5

• High-

-M

• 8 Q d

R 1, 2, 3, 3* e 6.

Control Room Air intet NA Radiation - High N.

Y E

e When handling Is radiated fuel in the secondary contaltunent.

3

a. Instrument slignment using a standard current sou rce.

e t

f

'k I

.n..

Wh 3

!) e,

l h;

l pH i

/

Ch' M

,Il 4

I 3 6

- -.-.j,

6-IV INSTRUMENTATION 3/4.3.6 CONTROL ROD BLOCK INSTRUMF,NTATION LIMITING CONDITION FOR OFiiRA110N 3.3.6.

The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values i

shown in the Trip Setpoint column of Table 3.3.6-2.

)

APPLICABILITY: As shown in Table 3.3.6-1.

ACTION:

i a.

With a control rod block instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.6-2, declare the channei inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value, b.

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, take the ACTION required by Table 3.3.6-1.

SURVEILLANCE REQUIREMENTS e

a UA 4.3.6 Each of the above required control rod block tripf systems and instrumentation channels shall be demonstrated OPERABLE %y the performance of the CHANNEL CHECK, CHANNE' FUNCTIONAL TEST and C.;ANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1.

<+v vm

~

• ACHRMNEL MAy BE PLnLED IN AN INOFERQGif STATV5 f0Q V9 To 6 %DMRt foQ RI@hnfo SunNihu.AM E WG90VT 9L94\\NL ME M.W SySTENs \\tbW -MAPPED COND\\TiOtd, l

9ROJ1DED AT LEAST ONE GNER OPERAGLE CHAluNEL IN TBE CPNCTf& SYSEIA 6 MONTT0Q\\%MT PARATtCTER.

~ +

~m l

GE-STS (BWR/4) 3/4 3-51 l

l

6-TABLE 3.3.6-1 g

CONTROL ROD BLOCK INSTRUMENTATION A

MINIMUM APPLICABLE d

OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION PER 1 RIP FUNCTION CONDITIONS ACTION

^

1.

ROD BLOCK MONITOR (a)-

2 1*

60 A

a.

Upscale b.

Inoperative 2

1*

60 c.. Downscale 2

1*

50 2.

APRM a.

Flow Biased Neutrcn Flux -

Upscale 4

1 61 b.

Inoperative 4

1, 2, 5 61 c.

Downscale 4

1 61 d.

Neutron Flux - Upscale, startup 4

2, 5 61 3.

SOURCE RANGE MONITORS Detector not full in(b) 3 2

61 w

e 1

2 5

61 T

b.

Upscale (c) 3 2

61 m

2 5

61 Inoperative (c) 3 2

c.

d.

Downscale( )

3 2

4.

INTERMEDIATE RANGE MONITORS Detector not full in (I"))

6 2, 5 61 a.

b.

Upscale 6

2, 5 61 Inoperatig) 6 2, 5 61 c.

Downscale d.

6 2, 5 61 5.

SCRAM DISCHARGE VOLUME e.

Water Level-High (2) 1, 2, 5**

62 b.

Scram Trip Bypass (2)

(1, 2,) 5**

62 6.

REACTOR COOLANT SYSTEM REC 1RCULATION FLOW a.

Upscale 2

1 62 b.

Inoperative 2

1 62 c.

(Comparator) (Downscale) 2 1

62

jy 4

r TABLE 3.3.6-1 (Continued)

CONTROL 20') BLCCV. INSTRUMENTATION ACTION ACTION 60 -

Declare the RBM inoperable and take the ACTION required by Specification 3.1.4.3.

With the number of OPERABLE Channels:

ACTION 61 a.

One less than required by the M011 mum OPERABLE' Channels per Trip Function requirement, restore the inoperable cnannelto OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour, b.

Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.

With the number of DPERABLE channels less than required by the ACTION 62 Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within ew-*mm 12 ' HOURS NOTES With THEP. MAL POWER > (30)% of RATED THERMAL POWER.

With more than one control rod withdrawn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

The RBM shall be automatically bypassed when a peripheral control rod is a.

selected (or the reference APRM channel indicates less than (303% of RATED THERMAL POWER).

b.

This function shall be autt.matica11y bypassed if detector count rate is

> 100 cps or the IRM channels are on range (3) or higher.

c.

Tnis function shall be automatically bypassed when the associated IRM channels are on range 8 or higher.

d.

This function shall be automatically bypassed when the IRM channels are on range-3 or higher.

~

This function shall be automatically bypassed when the IRM channels are e.

on range 1.

GE-STS (BVR/4) 3/4 3-53

i E

TABLE 3.3.6-2 CONTROL R0D DLOCK INSTRUMENTATION SETPOINTS qj hj TRIP FUNCTION TRIP SETPOINT ALLOWA8LE VALUE jf 1.

R0D BLOCK MONITOR Ro a.

Upscale 5 0.66 W + (40)%

$ 0.66 W + (43)%

d; b.

Inoperative NA NA c.

Downscale 1 (5)% of RATED THERMAL POWER 1 (3)% of RATED THERMAL POWER 2.

APRM a.

Flow Biased Neutron Flux -

Upscale

< 0.66 W + (42)%*

5 0.66 W 6 (45)%*

b.

Inoperative NA NA c.

Downscale 1 (5)% of RATED THERMAL POWER 1 (3)% of RATED THERMAL POWER d.

Neutron Flux - Upscale, Startup 5 (12)% of RATED THERMAL POWER

$ (14)% of RATED THERMAL POWER 3.

SOURCE RANGE MONITORS a.

Detector not full in NA NA 5

5 b.

Upscale 5 (2 x 10 ) cps (5 x 10 ) cp3 w

c.

Inoperative NA NA 30 d.

Downscale 1 (3) cps 1 (2) cps w

E 4.

INTERMEDIATE RANGE MONITORS a.

Detector not full in NA NA b.

Upscale 5 (108/125) divisions of 5 (110/125) divisions of full scale full scale c.

Inoperative NA NA d.

Downscale 1 (5/125) divisions of 2 (3/125) divisions of full scale full scale 5.

SCRAM DISCHARGE VOLUME a.

Water Level-High

<(

) inches 5(

) inches b.

Scram Trip Bypass NA NA 6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscale 5 (108/125) divisions of 5 (111/125) divisions of full scale full scale b.

Inoperative NA NA c.

(Comparator)'(Downscale) 5 (10)% flow deviation 5 (11)% flow deviation

• The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W).

The trip setting of this function must Le maintained in accordance with Specification 3.2.2.

I m,

o TABLE 4.3.6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS NOTES:

Neutron detectors may be excluded from CHANNEL CALIBRATION.

a.

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.

Includes reactor manual control multiplexing system input.

With THERMAL POWER > (30)% of RATED THERMAL POWER.

With more than one control rod withdrewn.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

i l

i I

l l

I l

l l

l l

l l

l l

GE-STS (BWR/4) 3/4-3-56 l

.