ML20214F170

Responds to NRC 870407 Request for Addl Info Re 870106 Application for Amends to Licenses DPR-24 & DPR-27 Concerning RCS Low Flow Trip Surveillance Testing.Procedures ICP 2.1,ICP 2.3 & ICP 2.17 Encl

ML20214F170
Person / Time
Site:	Point Beach
Issue date:	05/15/1987
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	Wagner D NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20214F173	List: ML20214F170 ML20214F176 ML20214F187 ML20214F193
References
TAC-64688, TAC-64689, VPNPD-87-205, VPNPD-87-52, NUDOCS 8705220435
Download: ML20214F170 (8)

Text

~

IMSCORSin Electnc ma coumr 231 W. MICHIGAN,P.O. BOX 2046 MILWAUKEE,Wl53201 (414)221-2345 VPNPD-87-205 VPNPD-87-52 May-15, 1987 U. S. NUCLEAR REGULATORY COMMISSION Document Control Desk Washington, D. C. 20555 Attention: Mr. David Wagner , Project Manager Gentlemen:

DOCKETS 50-266 AND 50-301 JANUARY 6, 1987 AMENDMENT REQUEST REGARDING RCS LOW-FLOW TRIP SURVEILLANCE TESTING (TAC'S 64688 AND 64689)

POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 This letter is in response to the further information requested in T. G. Colburn's letter to C. W. Fay dated April 7, 1987, l regarding our license amendment request for reactor coolant system low-flow trip surveillance testing.

l Question 1 1 Please provide a copy of the electrical schematics and logic drawings containing these relays, their contacts, bistables, and a discussion showing how testing the double loss of flow trip at power would actually initiate a reactor trip.

Response

For brevity, this discussion is limited to one train of reactor protection circuitry and associated testing of the reactor coolant flow logic circuits. The discussion, however, is equally applicable to the other train of reactor protection circuitry.

-The attached " Logic Test Switch Table" shows that selector low switch S1 does not accommodate testing of the "A" plus "B" reactor coolant flow to reactor trip circuitry. Position 16 allows testing the "A" loop low-flow logic, while position 17 allows testing the "B" loop low-flow logic.

O C3 !

8705220435 870515 4 r PDR ADOCK 05000266 p PDR

\

(

I NRC Document Control Desk May 15, 1987 Page 2 Figure 1 shows the-reactor coolant flow and reactor coolant

. pump loss of power to reactor trip logic circuits and the P7 and P8 permissive bypasses of various portions of that logic.

Note that P7 and P8 are the "at power" trip interlock and single loss-of-flow trip interlock respectively as described in Technical Specification 15.2.3.2.

When selector switch S1 is rotated to position 16,' contacts P16-Dl, P16-D2, and P16-D3, shown on Figure 2, close. This action allows the "Y" relays to become energized when logic test switches S2, S3, and S4 are closed. It also permits-the-testing of the two out of three logic combinations. When the "Y" relay energizes, contacts 2 and 6 and contacts 7 and 3 open causing the "X" relay to deenergize (Figure 2). Figure 1 shows how the "X" relay contacts (PC-411X through FC-415X) make up the required two out of three logic. When two out of three "X" relays deenergize, the associated contacts open causing relays RC-1 and RC-2 to deenergize, and RC relay contacts 4 and 8 and contacts 1 and 5 in the reactor trip circuit to open. If reactor power is above P8 (49 percent reactor power), the RT (reactor trip) relays deenergize and trip the reactor trip breaker. The "B" loop lcw reactor coolant flow is tested in the same manner except the rotary selector switch is placed in position 17, and relays RC-3 and RC-4 are deenergized.

-Figure 1 also shows that the "A" plus "B" low reactor coolant flow trip is actuated by relay contacts 1 and 5 on RC-1, i.e.,

< when reactor power is less than 10 percent, this reactor trip is defeated.- There is no available logic test circuitry on selector switch S1 to perform a test of simultaneous low flow in "A" plus "B" loops..

Another test device installed in this circuitry is the bistable test switch (Figure-3). The bistable test switch deenergizes l the associated "X" relays in both trains of reactor protection.

l If reactor power is above P8, operating two out of three bistable test switches in the same loop will result in a single-loop loss of flow reactor trip from the unbypassed train of reactor protection (only one train of reactor protection may be bypassed during testing).

l As shown in Figures 1 and 3, the same bistables and relays which generate the single loss of flow trips also generate the l

trip for simultaneous loss of flow in both loops. These l

bistables and relays are tested monthly. Only the simultaneous opening of contacts 1 and 5 on relays RC-1, RC-2, RC-3, and RC-4 are not verified by the normal monthly tests.

I

. l

'NRC~ Document ~ Control Desk-May 15, 1987 Page'3 l

Question 2-Please provide a copy of your current procedure for testing the reactor coolant low-flow trip circuitry. Has the procedure always specified that the-logic testing for low flow both loops be performed.on a refueling interval fregaency?

Response

Enclosed are copies of current test procedures ICP 2.1,

" Periodic Test Reactor Protection and Safeguards Analog Channels I Through-IV;" ICP 2.3, " Periodic Test Reactor.

Protection System Logic;" and ICP 2.17, " Periodic Test Reactor Protection System Logic (Post Refueling).". ICP 2.1 is used for monthly testing of the reactor protection bistables, including the loss of reactor coolant flow trip bistables. (See procedure steps 7.59, 7.67,_7.160, 7.241, 7.249, and 7.322 for details.) ICP 2.3 is used for monthly reactor protection relay logic testing. (See steps 5.2.21-and 5.2.22.) ICP 2.17 is for-the post-refueling startup reactor protection relay logic testing. ICP 2.17 was issued July 1986 and includes testing of the reactor trip from simultaneous low flow. (See step-5.2.24.) To our. knowledge, other than post-maintenance testing-performed after relay changeout procedures, the simultaneous loss of flow for both loops trip test as described in ICP 2.17 was not performed prior to issuing the procedure. The testing that-was performed on single loss of flow reactor trip, however, did verify that the relay coils associated with contacts 1 and 5 operated correctly.

Question 3 10 CFR 50, Appendix A, General Design Criterion 21, states in part that the-reactor protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation. Further guidance on this requirement is given in Regulatory Guide 1.'22, " Periodic Testing of Protection System Actuation Functions." Discuss this requirement with respect to the loss of flow trip (both loops) for Point Beach Unit 1 and 2.

Response

The regulatory position contained in Safety Guide 22 of February 17, 1972, provides for acceptable methods of testing protection systems during reactor operation. Sections D.2.b and D.4 have applicable criteria permitting testing of the simultaneous loss of flow to two loops reactor trip logic in the manner being used presently at Point Beach. All actuation devices and all actuated devices which are part of the reactor trip logic for loss of flow in both loops are being tested in

accordance with the guidance in these sections of the Safety Guide. (The Safety Guide was the predecessor of Regulatory Guide 1.22.)

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

s- =.

'NRC Document Control' Desk May 15, 1987 Page"4 The Point' Beach Nuclear Plant (PBNP) design predates Safety' Guide 22. . We, nonetheless, believe PBNP meets the intent of Safety Guide 22. LPBNP rarely operates at less than 50 percent power,1at which the loss of flow to both' loops trip signal:is

~

active.

Safety Guide 22.primarily discusses testing of' trip functions active during~ normal operating conditions.- Thistis consistent with our other-Technical Specifications not requiring. testing of low-power trips and permissives at power levels above their-setpoints. .This is, therefore, our basis for not generating a trip signal from loss of flow to both loops during normal-full-power operation.

The monthly tests of the involved bistables and relays provide-a very low probability that this loss-of-flow trip would fail to-function properly because all components except the relay contacts are tested. ;The additional testing now done each' refueling outage under ICP 2.17 further reduces the probability of failure of the relay contacts. Modifications of.the existing reactor protection system to allow testing at full-power is. impractical'and of limited value.because the trip function is not operable-at normal power levels.at PBNP.

In. addition to meeting the criteria of Safety Guide 22 by.

testing the actuating devices and the actuated equipment, we.

have started testing the relay contacts' associated with the i loss of flow to both loops reactor trip circuit. It is tested

.at refueling intervals by performance of ICP~2.17 which was issued in July 1986. Our' Technical' Specification change request will better define-the purpose and' frequency of performing that portion of ICP 2.17 that tests.these relay contacts. -

'If you have any questions, please'do not hesitate to contact us.

Very truly yours, ,

s c. , uf' f/'

C.-W. Fay l

Vice President Nuclear-Power Enclosure i Copies to NRC Regional Administrator, Region III

!. NRC Resident Inspector e

i b E l- ,, . _ . . -

..,,,.,___._.,__-,-.___.s . , , - - . , . . . , _ _ , _ _ . . . _ . _ _ , _ _ . . _ _ . _ . . ., . . - - _ . _ . - .

r*:

r LOGIC TEST SWITCH TABLE S1l SELECTOR SWITCH LOGIC TEST SWITCHES Position Function 'S2 S3 S4 S5 S6' l' High Flux-Source Range NC31D NC32D 1/N33A 1/N31B . 1/N32B 2 High Flux-Int. Range.. NC35F NC36F 1/N38A 1/N35A 1/N36A 3 PWR Range High Flux Low Trip

_ NC41P- NC42P' NC43P NC44P 1/N47A' 4 Nuclear Power >X% NC41M NC42M NC43M NC44M

'5 Intermediate Range P6 NC35D NC36D 1/N39A - 1/N39B 6 Pressurizer Low Press PC429E PC430H PC431J PC4,49A

'7 Overtemp Delta T- TC405C TC406C TC407C .TC40BC 8 Overpower Delta T- TC405A TC406A TC407A TC408A

-9 Low SG FW Flow /Le tel FC466C FC467B LC462C. 'LC463E 10 . Low SG FW Flow / Level LC472C LC473E FC476C FC477B 11 PWR Range High Flux.

High Trip. , NC41R NC42R NC43R NC44R

12. Nuclear Power >Z% NC41S NC42S NC43S NC44S 13 Nuclear- Power >Y% NC41N NC42N NC43N NC44N-

14. 4160. Bus voltage. 273/A01 274/A01 273/A02 274 /A02 15 Pressurizer High Level LC426A LC427A LC428A 16 Reactor. Coolant Flow FC411 FC412 FC413 17- Reactor Coolant Flow- FC414 FC415 FC416 18 Pressurizer High Press PC429A PC430A PC431A.

19 Turbine Auto Stop 63/AST3 63/AST4 63/AST5 20 LO-LO SG Water. Level LC461B LC462A LC463C 21 LO-LO SG Water Level LC472A LC473C LC471B

-22 Loss of Circ Water Pumps A52-06 A52-12 23 Reactor Coolant Loss of Power RCPA RCPB 24 Close Stop Valves 33AC-SL '33AC-SR 25 Turbine Power >X% PC485A PC486A 26 Safety Injection SIllX-A SIllX-B 27 High Condenser Press PC484A

-28 Spare 29 Spare 30 Spare TRAIN "A" ONLY (TYPICAL)

g ,..

i e s FIGURE 1 REACTOR COOLANT LOOP A LOOP 8 REACTOR COOLANT FLOW LOOP A RCP KR. R KR. p p i

I H E I g g

'3 -

@1 @t 1 @1 @1 @1 e4#4#4 e . .

> 34&434 1r 1r't IP r 1r r, ir 1r 1r 2/3 2/3 j

P7 P8 1r if it it P8

+ ,,, ,, ,,

1rit 1r 1r 1r 1r P7 1r 1r l

l 4r g r 1r 1r .gj wo

~

if 1r BOTH PUMPS EITHER PUMP TO RE ACTOR TO RE ACTOR

• TRIP CKT. TRIP CKT.

REACTOR COOLANT FLOW LOGIC L

7 - a l Y ret,1Y - TEST R ELA Y FIGURE 2 ,

X . nil.4f - LOGIC AEL A Y ,

SI - ROTARY SELCTOR TEST SwirCH (30 POSITION SWlfCH) 52;S3 S4- LOGIC TEST SWITCH P - WOTARY SELECTOR SWITCH PC SITION O -ROTARY SELECTOB SWITCH OECK 4t/ /

t?'  %,

u. / c,.

St FC g,To57pygg fc

'i H ' + pc -q 4II t f16 02 *

\I f//

IY CZ f '

lix 7-$' 3 '

E fc /To bA'Y pc

'il2 - BISTABLE sqqg.g

.- FC it2-

\. di;f Plt 03 -

.O lln 7 j

si ,.c To y,&y3- cs

.: il3-Y f'}[q',"y" 1 $X

\ $lJy a f17-98 .

^

!!* 3

/ ,,

Si fc To 'fl 41 y I4*Y 'y'fI#"y' q, 4I4*X

\ FC e z.

fn-a2 -: \f'Tl ,

l ln- / 3 e, St rc To elf-y c or.y BIsrnou (g.

~

~'

fC\ 11.s~ 4t hY '*

rs7-o3 ii, fl yu sl FL Tol#*Y ' f 4tc.-Y BISTA8J 4lt,-X 3 55 fc YI6 g ,

\ tis-y :s, N

s2 <

S3 e se l ' .. 1 " ,. 1" , .9 ,

14 26 x SEE LOGIC TEST' SWITCH TABLE THIS DRANING IS NOT A COMPLETE CIRCulT WIRING SCHEh1E. ETS PL4Rf0$6 IS TO SHOW THE AffLICABLE LOGIC TEST swlTCHis AND THEIR CONTACTS OPERATioll lN THE RE/1CTOR COOL ANT FLOW t 06 tC RELAY CIRCulTS.

TR A I N 'A ONL Y

r.

' i- FIGURE 3 . -

125 VOC _.

99 I  ??-l 11 Ii -

Sy # 1 I5" 11 17 Pt E P ?'2

Ir ql aeI RC-l RC-3 I l if AceAx yg v!!r rI iv v ilT-- ,h.

G5 RCfAx Reer 5

-S ,l!s s!!, her -

a:-z ec 1 ac3 ggax LOD'P 'A Lob B Y '

ti RT7 4 RT8 - REACTOR TRIP RELA YS nr Loofs A+a RCPAX+KfBX- REACTOR C00LAtfT Pt/MP LDSS OF POWER RELAY CONTACTS 49 -K 4 2.x 1 3 U- SI 'l FC s'? ct & C2 Vilf Q/

$ ct & C 2.

/ 315 'f l I f W LOOP A REACTOR COOLANT FLOW dy-x vff-x. '

ll tI siIS Ell1 o' VI -T II cl C2 yllt g FC

[ ll Cl M C2

~k s i 'S 4 IIF $ ~.s LOOP B REACTOR COOLANT FLOW TRhIN "A" nNL Y