ML18093A272

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)

Text

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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in HOT STANI:BY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANI:BY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the reactor trip breakers.

DESIGNATION P-6 P-7 SA LEM -UN IT 1 REACTOR TRIP SYSTEM INTERLOCKS CONDIT ION AND SET PO INT FUNCTION 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. With 2 of 4 Power Range Neutron Flux Channels > 11% of RATED THERMAL POWER or 1 of 2 Turbine impulse chamber pressure channels > a pressure equivalent to 11% of RATED THERMAL POWER. 3/4 3-7 P-7 prevents or defeats the automatic block of reactor trip on: Low flow in more than one primary coolant loop, reactor coolant pump undervoltage and frequency, pressurizer low pressure, pressurizer high level, and the opening of more than one reactor coolant pump breaker. \

Ul :i::i t""I trj ::s: c:: TABLE 3.3-1 (Continued) z H REACTOR TRIP SYSTEM INSTR LMENTAT ION 8 I-' MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE F UNCT 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 21. Reactor Trip Breakers 2 1 2 1,2 and* 1### I "" 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in HOT STANCBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANCBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and/or open the reactor trip breakers.

DES !GNAT ION P-6 P-7 SA LEM -UN IT 2 REACTOR TRIP SYSTEM INTER LOCKS CONDIT ION AND SET PO INT FUNCTION 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. With 2 of 4 Power Range Neutron Flux Channels > 11% of RATED THERMAL POWER or 1 of 2 Turbine impulse chamber pressure channels > a pressure equivalent to 11% of RATED THERMAL POWER. 3/4 3-7 P-7 prevents or defeats the automatic block of reactor t ri p on: Low fl ow in mo re than one primary coolant loop, reactor coolant pump undervoltage and frequency, pressurizer low pressure, pressurizer high level, and the opening of more than one reactor coolant pump breaker.

(/) t"i trj :s: c:: TAB LE 3. 3-I (Continued) z H 1-3 REACTOR TRIP SYSTEM INSTR LMENTAT ION N MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE e F UNCT IONA L UN IT OF CHANNELS TO TRIP OPERAS LE MOIES ACTION 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 t ,* '.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. 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 del ta-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 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 Of the reactor protectiai system are maintained for a sin;Jle lcq> less of flow.

I t TRIP RCPl BREAKER TRIP RCP2 BREAKER TRIP RCP3 BREAKER TRJP RCP"4 BREAKER RCPI BRKR, OPEN RCP2 BRKR. RCP3 BRKR. RCP"4 BRKR. FLOW LOOP I SEE FUNCTIONAL DIAGRAM DAAWING NO. 22111412-B-cm.42 JI Ill FLO'll LOOP 2 SEE FUNCTIONAL DIAGRAM ORA'lllNG NO. 22111412*B*'l542 II Ill REACTOR TRIP (SHEET NO. 2) OPEN FLOW LOOP J SEE FUNCTIONAL DlAGRAM DRAWING NO. 22111"413-B-q5"42 II Ill l -EXISTING LOW FLOW AND REACTOR COOLANT PUMP BREAKER POSITION REACTOR TRIP LOGIC FOR SALEM UNlTS 1 AND 2. OPEN FLD'll LOOP 4 . SEE FUNCTIONAL DIAGRAM DRA'lllNG NO.

II Ill REACTOR TRIP (SHEET NO. Zl OPEN I

• TRIP RCPJ BREl'IKER TRIP RCP2 BREAKER TRIP RCP3 BREAKER TRIP RCP4 BREAKER RCPI BRKR. P2 BRKR. OPEN RCP3 BRKR. RCP4 BRKR. FLOW LOOP I SEE FUNCTIONAL DIAGRAM DRAWING NO. 220412-B*Cl542 II Ill OPEN FLOW LOOP 2 SEE FUNCTIONAL DIAGRAM DRAWJNG NO. 228<412-8-'Hl-42 II Ill REACTOR TRIP <SHEET NO. 2l FLOW LOOP 3 SEE FUNCTIONAL DIAGRAM DRAWING NO. 22"413-B-q542 II III OPEN FIGURE 2 -PROPOSED MODlFICATlONS TO THE SALEM REACTOR COOLANT PUMP BREAKER POSITION REACTOR TRIP LOGIC. FLOW LOOP 4 SEE FUNCTIONAL DIAGRAM DRAWING NO.

II III P-7 REAC'TOR 'TRIP* !SHEET NO. 2) OPEN I '

ATTACHMENT 0'3 NUCLEAR SAFETY EVALUATION CHECK LISTS FOR DELETION OF RCP BREAKER POSITION TRIP r I SF.CL-87-219 . 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 lOCFRSO. 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) (4.2) (4.3) (4.4) (4.5) (4.6) (4. 7) Yes_ No__x_ Will the probability of an accident previously evaluated in the FSAR be increased?

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

Yes_ No_K_ May the possibility of an accident which is different than any already evaluated in the FSAR be created? Yes_ No_K_ Will the probability of a malfunction of equipment important to safety previously evaluated in the FSAR be increased?

Yes_ No__x_ Will the consequences of a malfunction of ment inportant to safety previously evaluated in the FSAR be increased?

Yes_ No_x_ May the possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR be created? Yes_ No_K_ Will the margin of safety as defined in the bases to any technical specification be reduced? Page 1 of 2

. .,...16D *If the answers to any of the above questions are unknown, indicate

• under 5) REMARKS and 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 lOCFR.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 lOCFR50 Appendix B and the specific requirements of IEEE-279-1971.

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

FOR FSAR. . Section: ____ Pages: ____ Tables: _____ Figures: ____ _ Reason for/ Description of Change: Prepared by (Nuclear Safety) :--'G=*=E'"""".

-..w..p-=-

---

(?. d:] Coordinated with Engineer(s)

_c=."--"'Albens====i.__

__

Coordinated Group Manager(s)

Nuclear Safety Group Page 2 of 2 Date: Date: Date: Date: .r--1'1.:..

g7 .S-14-87 '>t/NJ1 ATTACHMENT 4' RECOMMENDED CHAPTER 7 CHANGES TO REFLECT THE DELETION OF RCP BREAKER POSITION TRIP Designation P-4 P-6 P-7 P-8 P-10 SGS-UFSAR TABLE 7.2-2 INTERLOCK CIRCUITS Derivation Reactor trip ,_ 1/2 Neutron flux (intermediate range) above setpoint 2/2 Neutron flux (intermediate range) below setpoint 3/ 4 Neutron flux . (power range) below setpoint (from P-10) and 2/2 Turbine impulse chamber pressure below setpoint (from P-13) 3/4 Neutron flux (power range) below setpoint 2/4 Neutron flux (power range) above setpoint Function 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 Allows manual block of source range reactor trip Def eats the block of source range reactor trip Blocks reactor trip on: flow or reactor coolant breakers open in more one loop, undervoltage, underfrequency, turbine trip, pressurizer low pressure, and pressurizer high level Low pump than Blocks reactor trip on low flow eF FeaeteF eeeleftt -ptHHp bFeekeF epen in e single leap Allows manual block of power range (low setpoint) reactor 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 Revision 6 February 15, 1987 r 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 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

• ALSO CAUSES A REACTOR TRIP. * ---* -**-

• --*-*-----------* --

Reactor Trip 9. Monitored electrical supply to reactor coolant pumps: 9A. Undervoltage 9B. Underfrequency 9C. Reactor coolant pump breakers 10. Safety injection signal (actuation)

SGS-UFSAR TABLE 7.2-1 (Cont) Coincidence Circuitry and Interlocks 1/2 taken twice, interlocked with P-7 1/2 taken twice, interlocked with P-7 P7 Interlocked with.P 8 PS Low pressurizer pressure (2/3) or 2/3 high containment pressure; or 2/3 differential steam line pressure signals of one line compared with the other three 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 Comments 1/2 twice underfrequency signals trip all reactor coolant pumps and directly actuate reactor trip: interlocked with P-7. (Opening of the coolant pump breakers will also actuate a reactor trip) Blocked below P-7. Open breaker in 1 loop t->.BOVE P-;. permitted helnw P 8. Trips main feedwater pumps. Closes all feedwater control valves. Closes feed-:o* water pump discharge valves and initiates Phase A isolation.

Initiates turbine trip. Revision 6 February 15, 1987