Text

Proposed Tech Specs,Changing Reactor Coolant Pump Breaker Position Trip Logic from 1 Out of 4 Logic Above 36% Power Level (Permissive P-8) to 2 Out of 4 Logic Above 11% Power (Permissive P-7)
	ML18093A272
Person / Time
Site:	Salem
Issue date:	07/24/1987
From:	; Public Service Enterprise Group
To:	;
Shared Package
ML18093A271	List: ML18093A270; ML18093A272;
References
	NUDOCS 8708030397
	Download: ML18093A272 (19)
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ATTACHMENT 1 RCP BREAKER POSITION REACTOR TRIP LOGIC CHANGE TECHNICAL SPECIFICATION MARK-UPS 8708030397 870724 PDR ADOClr\ 05000272 p POO I

LIMITING SAFETY SYSTEM SETTINGS BASES Safety Injection Input from ESF If a reactor trip has not already been generated by the reactor protective instrumentation, the ESF automatic actuation logic channels will initiate a reactor trip upon any signal which initiates a safety injection. This trip is provided to protect the core in the event of a LOCA. The ESF instrumentation channels which initiate a safety injection signal are shown in Table 3.3-3.

Reactor Coolant Pump Breaker Position Trip The reactor Coolant Pump Breaker Position Trip is an anticipatory trip which provides reactor core protection against DNB resulting from the opening of two or more pump breakers above P-7. This trip is blocked below P-7. The open/close position trip assures a reactor trip signal is generated before the low flow trip set point is reached. No credit was taken in the accident analyses for operation of this trip. The functional capability at the open/close position settings is required to enhance the overall reliability of the Reactor Protection System.

SA LEM - UN IT 1 B 2-8

TABLE 3.3.1 (Continued)

ACTION 9 With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in HOT STANI:BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per specification 4.3.1.1.

ACTION 10 Deleted ACTION 11 With less than the Minimum Number of Channels OPERABLE, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 12 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANI:BY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

REACTOR TRIP SYSTEM INTERLOCKS DESIGNATION CONDIT ION AND SET PO INT FUNCTION P-6 With 2 of 2 Intermediate Range P-6 prevent or defeats the Neutron Flux Channels < 6 x 10-ll the manual block of source amps. range reactor trip.

P-7 With 2 of 4 Power Range Neutron P-7 prevents or defeats the Flux Channels > 11% of RATED automatic block of reactor THERMAL POWER or 1 of 2 Turbine trip on: Low flow in more impulse chamber pressure channels than one primary coolant

> a pressure equivalent to 11% of loop, reactor coolant pump RATED THERMAL POWER. undervoltage and under-frequency, pressurizer low pressure, pressurizer high level, and the opening of \

more than one reactor coolant pump breaker.

SA LEM - UN IT 1 3/4 3-7

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c:: TABLE 3.3-1 (Continued) z H

8 REACTOR TRIP SYSTEM INSTR LMENTAT ION I-'

MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCT IONA L UN IT OF CHAN NE LS TO TRIP OPERABLE MOC£S ACTION

18. Turbine Trip

a. Low Autostop Oil Pressure 3 2 2 1 7#

b. Turbine Stop Valve Closure 4 4 4 1 7#

19. Safety Injection Input from SSPS 2 1 2 1,2 1

20. Reactor Coolant Pump Breaker 1/breaker 2 1/breaker 1 11 w Position Trip (above P-7) per aper-

.......... ati ng loop

""w I 21. Reactor Trip Breakers 2 1 2 1,2 and* 1###

"" 22. Automatic Trip Logic 2 1 2 1,2 and* 1

LIMITING SAFETY SYSTEM SETTINGS BASES Undervoltage and Underfreguency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection against DNB as a result of loss of voltage or underfrequency to more than one reactor coolant pump. The specified setpoints assure a reactor trip signal is generated before the low flow trip setpoint is reached. Time delays are incorporated in the underfrequency and undervoltage trips to prevent spurious reactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 0.9 seconds.

for underfrequency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequency trip setpoint is reached shall not exceed 0.3 seconds.

Turbine Trip A Turbine Trip causes a direct reactor trip when operating above P-9. Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient. No credit was taken in the accident analyses for operation of these trips. Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System.

Safety Injection Input from ESF If a reactor trip has not already been generated by the reactor protective instrumentation, the ESF automatic actuation logic channels will initiate a reactor trip upon any signal which initiates a safety injection. This trip is provided to protect the core in the event of a LOCA. The ESF instrumentation channels which initiate a safety injection signal are shown in Table 3.3-3.

Reactor Coolant Pump Breaker Position Trip The Reactor Coolant Pump Breaker Position Trip is an.anticipatory trip which provides reactor core protection against DNB resulting from the opening of two or more pump breakers above P-7. This trip is blocked below P-7. The open/close position trip assures a reactor trip signal is generated before the low flow trip setpoint is reached. No credit was taken in the accident analyses for operation of this trip. The functional capability at the open/close position settings is required to enhance the overall reliability of the Reactor Protection system.

SALEM - UNIT 2 B 2-7

TAB!£ 3.3.1 (Continued)

ACTION 9 With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in HOT STANCBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per specification 4.3.1.1.

ACTION 10 Deleted ACTION 11 With less than the Minimum Number of Channels OPERAB1£, operation may continue provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 12 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERAB1£ status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANCBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the reactor trip breakers.

REACTOR TRIP SYSTEM INTER LOCKS DES !GNAT ION CONDIT ION AND SET PO INT FUNCTION P-6 With 2 of 2 Intermediate Range P-6 prevent or defeats the Neutron Flux Channels < 6 x 10-ll the manual block of source amps. range reactor trip.

P-7 With 2 of 4 Power Range Neutron P-7 prevents or defeats the Flux Channels > 11% of RATED automatic block of reactor THERMAL POWER or 1 of 2 Turbine t ri p on: Low fl ow in mo re impulse chamber pressure channels than one primary coolant

> a pressure equivalent to 11% of loop, reactor coolant pump RATED THERMAL POWER. undervoltage and under-frequency, pressurizer low pressure, pressurizer high level, and the opening of more than one reactor coolant pump breaker.

SA LEM - UN IT ~- 2 3/4 3-7

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c:: TAB LE 3. 3-I (Continued) z H

1-3 REACTOR TRIP SYSTEM INSTR LMENTAT ION N

MINIMUM FUNCT IONA L UN IT TOTAL NO.

OF CHANNELS CHANNELS TO TRIP CHANNELS OPERAS LE APPLICABLE MOIES ACTION e

18. Turbine Trip

a. Low Autostop Oil Pressure 3 2 2 I 7#

b. Turbine Stop Valve Closure 4 4 3 I 6#

I9. Safety Injection Input from SSPS 2 I 2 I,2 I w

20. Reactor Coolant Pump Breaker I/breaker 2 I/breaker I 11 w Position Trip (above P-7) per aper-I ati ng loop 2I. Reactor Trip Breakers 2 I 2 I,2 and* I###

22. Automatic Trip Logic 2 I 2 I,2 and* I

. I ATTACHMENT Z..

NON-LOCA TRANSIENT EVALUATION FOR THE DELETION OF RCP BREAKER POSITION TRIP

'.Ihe reactor coolant p:mp breaker position reactor trip is primarily an enticipatOcy trip. If an RCP breaker trips, the lCM flCM trip setpoint is typically reached two to three secx:inds after the RCP begins coastirg down.

~us the breaker qi.en signal WCAlld trip the plant in anticipaticri of the lCM flCM reactor trip. I.cM flow in one of the four reactor coolant lc:qs above the P-8 setpoint and low flow in two of the four reactor coolant lc:qs between P-8 ard P-7 will trip the reactor. :Reactor trip ai lCM flow is blocked bel.CM P-7. ~ existirq RCP breaker position reactor trip loqic is identical to the low flow reactor trip logic (i.e. blocked below P-7, 2 of 4 between P-7 and P-8 and l of 4 above P-8) * 'lhe existirq low flow and RCP breaker position trip logic is sb:7wn in Figure 1.

~ prcposed logic is sJnm in Figure 2. 'llle cruy c::hailJe in the result.irg protection is that reactor trip ai one RCP breaker open signal \hell qierat.irg above the P-8 setpoint will rrM be blocked (i.e., two breakers open will be required at all p::rwers above P-7 to trip the reactor)

If the RCP is coastiD3 down (i.e. the breaker c:pen signal is real) , then reactor trip will occur cm lCM flCM. '!he existirq FSAR analysis shows that for one reactor coolant p.mp coastirg down, adequate protection is provided by the lCM flCM trip.

In the event of a sin;Jle loop less of flow, the lCM flCM reactor trip is the design protection, and it meets the design requirement of maint:ainin; the minim.ml mBR above the l:imit value. If no credit is taken for the lCM flCM xeactcr trip, a reactor trip ai either cwertemperature or ~

delta-T wc:W.d terminate the acx:ident before ma OCOJrS in a significant port.ion of the core. ~ analysis shcMs that the hot spot clad tenrierature (ai the inner clad surface) remains well bel.CM the meltllg point.

'lhree redurmnt flow dlanne.ls are provided for each loc::p fer the low flCM reactor trip. Above P-8, less of flotr1 in 1Jrrf e11e loop, as sensed by two of the three dlannels, actuates a reactor trip. For the c:werteJiperature ard ovezpc:Mer delta-T reactor trips, cme channel per loop is prcwided, and overtarperature or oveJ:pCWe.r delta-T sensed by ertJ two dlannels trips the xeactor. .

~ deletion of reactor trip an a sinqle RCP breaker position signal above P=S, as desc:r.ibsd above, affects only the coincidence logic of the RPS ard does not degrade either its perfonnanoe er ~onnanoe to system

functional requirements. Also the dlan:;1e provides a reducticn in cballerges (i.e. sp.Jrioos trips) to the reactor trip system.

For all ncn-I.OCA safety analyses, the lcqic charge \rttl.ch results in the deletia'l of reactor trip al a sin;Jle reactor coolant p.mp breaker open above P-8 is acx::eptable and the existirq FSAR Loss of Flow Analyses are applicable. 'lhe diversity and~ Of the reactor protectiai system are maintained for a sin;Jle lcq> less of flow.

TRIP RCPl TRIP RCP2 TRIP RCP3 TRJP RCP"4 BREAKER BREAKER BREAKER BREAKER RCPI BRKR, RCP2 BRKR. RCP3 BRKR. RCP"4 BRKR.

OPEN OPEN OPEN OPEN FLOW LOOP I FLO'll LOOP 2 FLOW LOOP J FLD'll LOOP 4 .

SEE FUNCTIONAL DIAGRAM SEE FUNCTIONAL DIAGRAM SEE FUNCTIONAL DlAGRAM SEE FUNCTIONAL DIAGRAM DAAWING NO. ORA'lllNG NO. DRAWING NO. DRA'lllNG NO.

22111412-B-cm.42 22111412*B*'l542 22111"413-B-q5"42 22111413-B-~2 JI Ill II Ill II Ill II Ill I

t REACTOR TRIP REACTOR TRIP (SHEET NO. 2) (SHEET NO. Zl Fl~E l - EXISTING LOW FLOW AND REACTOR COOLANT PUMP BREAKER POSITION REACTOR TRIP LOGIC FOR SALEM UNlTS 1 AND 2.

TRIP RCPJ TRIP RCP2 TRIP RCP3 TRIP RCP4 BREl'IKER BREAKER BREAKER BREAKER RCPI BRKR. P2 BRKR. RCP3 BRKR. RCP4 BRKR.

OPEN OPEN OPEN OPEN FLOW LOOP I FLOW LOOP 2 FLOW LOOP 3 FLOW LOOP 4 SEE FUNCTIONAL DIAGRAM SEE FUNCTIONAL DIAGRAM SEE FUNCTIONAL DIAGRAM SEE FUNCTIONAL DIAGRAM DRAWING NO. DRAWJNG NO. DRAWING NO. DRAWING NO.

220412-B*Cl542 228<412-8-'Hl-42 22"413-B-q542 229413-B-~2 II Ill II Ill II III II III I I P-7 REACTOR TRIP REAC'TOR 'TRIP*

<SHEET NO. 2l !SHEET NO. 2)

FIGURE 2 - PROPOSED MODlFICATlONS TO THE SALEM REACTOR COOLANT PUMP BREAKER POSITION REACTOR TRIP LOGIC.

ATTACHMENT 0 '3 NUCLEAR SAFETY EVALUATION CHECK LISTS FOR DELETION OF RCP BREAKER POSITION TRIP

SF.CL 219 .

CU.Stamer Reference No(s).

Westinghouse Reference No (s) *

(Change Control or RFQ As Applicable)

FCN-m:r0-40514 WESTINGHOUSE NUCLEAR SAFEI'Y EVAIUATION CHECK LIST

1) NUCLEAR PIANT(S) Salem Generating Station Unit 2 (PN.J)

2) CHECK LIST mCABIB 'ro:Modification of RCP Breaker Position Trip

3) 'Ih.e safety evaluation of the revised procedure, design change or modification required by 10CFRSO. 59 has been prepared to the extent required and is attached. If a safety evaluation is not required or is inconplete for any reason, explain on Page 2.

Parts A and B of this Safety Evaluation Check List are to be completed only on the basis of the safety evaluation perfonned..

CHECK LIST - PARI' A (3.1) Yes_lL No_ A change to the plant as described in the FSAR?

(3.2) Yes_ No__x_ A change to procedures as described in the FSAR?

(3. 3) Yes_ No__x_ A test or experiment not described in the FSAR?

(3. 4) Yes__x_ No_ A change to the plant technical specifications (Appendix A to the Operating License)?

4) CHECK LIST - PARI' B (Justification for Part B answers l1UlSt be included.

on page 2.)

(4.1) Yes_ No__x_ Will the probability of an accident previously evaluated in the FSAR be increased?

(4.2) Yes_ No_K_ Will the consequences of an accident previously evaluated in the FSAR be increased?

(4.3) Yes_ No_K_ May the possibility of an accident which is different than any already evaluated in the FSAR be created?

(4.4) Yes_ No_K_ Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

(4.5) Yes_ No__x_ Will the consequences of a malfunction of equip-ment inportant to safety previously evaluated in the FSAR be increased?

(4.6) Yes_ No_x_ May the possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR be created?

(4. 7) Yes_ No_K_ Will the margin of safety as defined in the bases to any technical specification be reduced?

Page 1 of 2

If the answers to any of the above questions are unknown, indicate

under 5) REMARKS and ~lain below.

If the answer to any of the above questions in 4) cannot be answered in the negative, based on written safety evaluation, the change cannot be approved without an application for license amendment submitted to the NRC pursuant to 10CFR.50.59.

5) REMARKS:

'!he following summarizes the justification upon the written safety evaluation, (*) for answers given in Part B of the Safety Evaluation Check List: .

'!he reactor trip from reactor coolant punp (RCP) breaker p:sition is mcxlified to actuate on the opening of any* two breakers above the P-7

. interlock setpoint. *

'!he changes only .affect the coincidence logic of the Solid State Protection System (SSPS) and does not degrade either its performance or confo:nnance to system design requirements. 'Ibis specific change to the design of the SSPS has undergone Westinghouse analysis and review per the general requirements of 10CFR50 Appendix B and the specific requirements of IEEE-279-1971.

(*) Reference to document(s) containing written safety evaluation:

.,...16D FOR FSAR. ~

. Section: _ _ _ _ Pages: _ _ _ _ Tables: _ _ _ _ _ Figures: _ _ _ __

Reason for/ Description of Change:

Prepared by (Nuclear Safety) :--'G=*=E'"""".-=I.ang="--........&...,Q?~*-..w..p-=------Date: .r--1'1.:.. g7

(?. d:]

Coordinated with Engineer(s) :_c=."--"'Albens====i.___ _~==---...;....:..L~-Date: .S-14-87 Coordinated Group Manager(s) =~n~.N~*~Ka!::!;;t~z:!.-.-_~---=.;...:_~~~ Date: ¢~h7 Nuclear Safety Group Manager:-'P~-~J~*-M~o::::rr=.=..=i.:::s~.::__-~~~='--Date: ~

'>t/NJ1 Page 2 of 2

ATTACHMENT 4' RECOMMENDED CHAPTER 7 CHANGES TO REFLECT THE DELETION OF RCP BREAKER POSITION TRIP

TABLE 7.2-2 INTERLOCK CIRCUITS Designation Derivation Function P-4 Reactor trip Actuates turbine trip Close main feedwater valves on T below setpoint avg Prevents opening of main feedwater valves which were closed by safety injection or

,_ high steam generator water level P-6 1/2 Neutron flux Allows manual block of source (intermediate range reactor trip range) above setpoint 2/2 Neutron flux Def eats the block of source (intermediate range reactor trip range) below setpoint P-7 3/ 4 Neutron flux . Blocks reactor trip on: Low (power range) flow or reactor coolant pump below setpoint breakers open in more than (from P-10) and one loop, undervoltage, 2/2 Turbine underfrequency, turbine impulse chamber trip, pressurizer low pressure below pressure, and pressurizer setpoint (from high level P-13)

P-8 3/4 Neutron flux Blocks reactor trip on low (power range) flow eF FeaeteF eeeleftt below setpoint - ptHHp bFeekeF epen in e single leap P-10 2/4 Neutron flux Allows manual block of power (power range) range (low setpoint) reactor above setpoint trip Allows manual block of intermediate range reactor trip and intermediate range rod stops (C-1)

Blocks source range reactor trip (back-up for P-6) 1 of 4 SGS-UFSAR Revision 6 February 15, 1987

Low Reactor Coolant Flow Trip This trip protects the core from DNB following a loss-of-coolant flow. The means of sensing loss-of-coolant flow are described below.

Low Primary Coolant Flow Trip A loop low flow signal is generated by two-out-of-three low flow signals per loop. Above the P- 7 setpoint (approximately 10 percent of full power) low flow in any two loops results in a reactor trip. Above the P-8 setpoint (approximately 60 percent of full power) low flow in any loop results in a reactor trip.

Reactor Coolant Pump Breaker Position Trip*

INSERT A HERE One open brea-ker-s-ignal is gene-rated for each rea-etor eoolant-

---p.wnp-. Above the F-7 s@tpoii:i.t th@ r@actor txips oi:i. two opei:i.

bre-ak-er signals, Above the p....g-se-t-point the reactor trips on

~-e-e-p~-kcr signal."

Reactor Coolant Pump Undervoltage and Underfrequency Trips There is one underfrequency and one undervoltage sensor per bus.

A 1/2 logic taken twice underfrequency signal-directly trips all of the reactor coolant pumps, and also produces a direct reactor trip (interlocked by P-7). (An indirect trip is produced by the pump breaker-position trip.) For undervoltage protection, there

. is an undervoltage sensoi;- on each of the four busses. Reactor trip above P-7 is actuated by a 1/2 logic taken twice.

All of these low reactor coolant flow trips are blocked below the P-7 setpoint (approximately 10 percent power).

7.2-23 SGS-UFSAR Revision 6 February 15, 1987

INSERT A OPENING OF TWO.REACTOR COOLANT PUMP BREAKERS ABOVE THE P-7 INTERLOCK SETPOINT, WHICH IS INDICATIVE OF AN IMMINENT LOSS OF COOLANT FLOW~

ALSO CAUSES A REACTOR TRIP. *

---* -**- -*--*---~~---- *--*-*- ------- ---* --

TABLE 7.2-1 (Cont)

Reactor Trip Coincidence Circuitry and Interlocks Comments :o*

9. Monitored electrical supply to reactor coolant pumps:

9A. Undervoltage 1/2 taken twice, interlocked with P-7 9B. Underfrequency 1/2 taken twice, interlocked 1/2 twice underfrequency signals trip all with P-7 reactor coolant pumps and directly actuate reactor trip: interlocked with P-7.

(Opening of the coolant pump breakers will also actuate a reactor trip)

P7 9C. Reactor coolant pump Interlocked with.P 8 an~ PS Blocked below P-7. Open breaker in 1 loop t->.BOVE P-;.

breakers permitted helnw P 8.

10. Safety injection signal Low pressurizer pressure (2/3) or Trips main feedwater pumps. Closes all (actuation) 2/3 high containment pressure; feedwater control valves. Closes feed-or 2/3 differential steam line water pump discharge valves and initiates - -

pressure signals of one line Phase A isolation. Initiates turbine compared with the other three trip.

lines; or 2/4 high steam flow in coincidence with 2/4 low T or avg

,2/4 low steam line pressure; or manual 1/2 (See 7.2 System 2 of 7 SGS-UFSAR Revision 6 February 15, 1987

ML18093A272

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