ML20149E679
| ML20149E679 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 12/30/1987 |
| From: | Lainas G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20149E685 | List: |
| References | |
| NUDOCS 8801130441 | |
| Download: ML20149E679 (21) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION g
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MISSISSIPPI POWER & LIGHT COMPANY SYSTEM ENERGY RESOURCES, INC.
SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 41 License No. NPF-29 1.
The Nuclear Regulatory Comission (the Comission) has found that A.
The application for amendment by System Energy Resources, Inc.,
' (thelicensee),datedAugust 13, 1987, as revised October 23, 1987, November 25, 1987 Decent >er 22, and December 27, 1987, complies with the standards and req)uirements of the Atomic Energy Act of 1954, as amended (the Act, and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon
)
defense and security or to the health and safety of the public; and E.
The issuance of this amend,ent is in accordance with 10 CFR Part 51 l
of the Comission's regulations and all applicable requirements have i
been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Facility Operating License No. NPF-29 is hereby L
amended to read as follows:
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8801130441 871230 I
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(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environ-mental Protection Plan contained in Appendix B, as revised through Arendment No. 41, are hereby incorporated into this license. System Energy Resources, Inc. shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Gus C. Lainas, As 1stant Director for Region II Reactors Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: December 30, 1987 1
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l ATTACHMENT TO LICENSE AMENDMENT NO. 41 3
FACILITY OPERATING LICENSE NO. NPF-29 i
l DOCKET NO. 50-416 o
i Replace the following pages of the Appendix A Technical Specifications with the attached pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
The corresponding overleaf page(s) have been provided to maintain document completeness.
Remove Insert 3/4 1-17 3/41-17(overleaf) 3/4 1-18 3/4 1-18 3/4 1-19 3/4 1-19 3/4 1-20 3/4 1-20 3/4 3-37 3/4 3-37 3/4 3-37a 3/4 3-38 3/4 3-38 (overleaf) 3/4 3-39 3/4 3-39 3/4 3-40 3/4 3-40 (overleaf) t 3/4 6-41 3/46-41(overleaf) 3/4 6-42 3/4 6-42 B 3/4 1-3 B3/41-3(overleaf)
B 3/4 1-4 5 3/4 1-4 l
B '!/4 1-4a B 3/4 1-4a i
B 3/4 3-1 B 3/4 3-1 (overleaf) l B 3/4 3-2 B 3/4 3-2 B 3/4 3-3 B 3/4 3-3 E 3/4 3-3a B 3/4 3-3a l
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1
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.4.2 The RPCS shall be demonstrated OPERABLE by:
Verifying the OPERABILITY of the rod pattern controller function a.
when THEWL POWER is less than the low power setpoint by selecting and attempting to move an inhibited control rod:
1.
After withdrawal of the first insequence control rod or gang for each reactor startup.
2.
As soon as the rod inhibit mode is automatically initiated at the RPCS low power setpoint, 20 +15, -0% of RATED THERMAL POWER, during power reduction.
3.
The first time only that a banked position, N1, N2, or N3, is reached during startup or during power reduction below the RPCS low power setpoint.
b.
Verifying the OPERABILITY of the rod withdrawal limiter function when THERMAL POWER is greater than or equal to the low power set-point by selecting and attempting to move a restricted control rod in excess of the allowable distance:
1.
As each power range above the RPCS low power setpoint is entered during a power increase or decrease.
2.
At least once per 31 days while operation continues within a given power range above the RPCS low power setpoint.
c.
Verifying each RPCS bypass switch is in the unbypassed position or is in compliance with ACTION b.3 of this specification:
1.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
Prior to a control rod movement, except by scram, following a power reduction to less than or equal to the low power setpoint.
GRAND GULF-UNIT 1 3/4 1-17 Amendment No. 28 t
NAR 5 997
REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 Two standby liquid control system subsystems shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2 and 5*.
ACTION:
a.
In OPERATIONAL CONDITION 1 or 2:
1.
With one system subtystem inoperaole, restore the inoperable subsystem to OPERABLE status within 7 days or be in at least H0T SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
With both standby liquid cor, trol system subsystems inoperable, restore at least one-subsystem to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least H0T SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
In OPERATIONAL CONDITION 5*:
1.
With one system subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 30 days or insert all insertable control rods within the next hour.
2.
With both standby liquid control system subsystems inoperable, insert all insertable control rods within one hour.
SURVEILLANCE REQUIREMENTS 4.1.5 Each standby liquid control system subsystem shall be demonstrated OPERABLE:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that; 1.
The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.
2.
The available volume of sodium pentaborate solution is greater than or equal to 4530 gallons.
l l
3.
The heat tracing circuit is OPERABLE by determining the temperature of the pump suction piping is within the limits of 1
Figure 2.1.5-1.
- With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
GRAND GULF-UNIT 1 3/4 1-18 Amendment No. 41
.J
REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.
At least once per 31 days by; 1.
Starting both pumps and recirculating demineralized water to the test tank.
2.
Verifying the continuity of the explosive charge.
3.
Determining that the available weight of sodium pentaborate is greater than or equal to 5800 lbs and the concentration of boron l
l in solution is within the limits of Figure 3.1.5-1 by chemical l
analysis.*
4.
Verifying that each valve, manual, power operated or automatic, in the flow path that is not locked, sealed, or otherwise l
secured in position, is in its correct position.
l c.
Demonstrating that, when tested pursuant to Specification 4.0.5, l
the minimum flow requirement of 41.2 gpm at a pressure of greater than or equal to 1300 psig is met, without actuation of the pump l
relief valte.
d.
At least once per 18 months during shutdown by; 1.
Initiating one of the standby liquid control system subsystems, including an explosive valve, and verifying that a flow path from the pumps to the reactor pressure vessel is available by pumping demineralized water into the reactor vessel.
The replacement charge for the explcsive valve shall l
be from the same manufactured batch as the one fired or from another batch which has been certified by having one of that
)
batch successfully fired.
Both system subsystems shall be tested in 36 months.
2.
Demonstrating that the pump relief valve opens within 3% of l
the system design pressure and verifying that the relief valve does not actuate during recirculation to the test tank.
3.
- Demonstrating that all heat traced piping between the storage tank and the reactor vessel is unblocked by pumping from the storage tank to the test tank and then draining and flushing the piping with demineralized water.
4.
Demonstrating that the storage tank heater is OPERABLE by
(
verifying the expected temperature rise for the sodium pentaborate solution in the storage tank after the heater is energized.
- This test shall also be performed anytime water or boron is added to the solution or when the solution temperature drops below the limit of Figure 3.1.5-1.
- This test shall also be performed whenever both heat traciog circuits have i
been found to be inoperable and may be performed by any series of sequential, I
overlapping or total flow path steps such that the entire flow path is included.
GRAND GULF-UNIT 1 3/4 1-19 Amendment No. 41
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do'3HnlVH3dW31 GRAND GULF UNIT l.
3/4 1-20 AMENDMENT NC. 41
3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 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 Secpoint column of Table 3.3.4.1-2.
APPLICABILITY:
OPE ~.;TIONAL CONDITION 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 the 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 onc trip system and:
1.
If the inoperable channels consist of one reactor vessel water level channel and one reactor vessel pressure channel, place both inoperable channels in the tripped condition
- witnin one hour or declare the trip system inoperable.
2.
If the inoperable channels include two reactor vessel water level channels or two reactor vessel pressure channels, declare the trip 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 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 />, e.
With both trip system inoperable, restore at least one trip system to OPERABLE status wit in 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 1 3/4 3-37 Amendment No. 41
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 CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.4.1-1.
4.3.4.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall he performed at least once per 18 months.
GRAND GULF-UNIT 1 3/4 3-37a Amendmen; No. 41
INSTRUMENTATION TABLE 3.3.4.1-1 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION MINIMUMOPERABLECHANgggSPER TRIP FUNCTION TRIP SYSTEM 1.
Low Low, Level 2 2
2.
Reactor Vessel Pressure - High 2
(a) One channel may be placed in an inoperable status 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 required surveillance provided the other channel is OPERABLE.
GRAND GULF-UNIT 1 3/4 3-38
-~
TABLE 3.3.4.1-2 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS TRIP ALLOWABLE TRIP FUNCTION SETPOINT VALUE 1.
> -41.6 inches *
> -43.8 inches low Low, Level 2 2.
Reccto. Vessel Pressure - High 5 1095 psig i 1102 psig
^See Bases Figure B3/4 3-1.
GRAND GULF-UNIT 1 3/4 3-39 Amendment No. 41
INSTRUMENTATION TABLE 4.3.4.1-1 ATWS RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL FUNCTIONAL CHANNEL TRIP FUNCTION CHECK TEST CALIBRATION 1.
5 M
R*
Low Low, Level 2 2.
Reactor Vessel Pressure - High 5
M R*
"Calibrate trip unit at least once per 31 days.
GRAND GULF-UNIT 1 3/4 3-40
i TABLE 3.6.4-1 (Centinued)
CONTAINMENT AND ORWELL ISOLATION VALVES SYSTEM AND PENETRATION VALVE NUMBER NUMBER Containment (Continued)
RHF. Pump "B" Test E12-F335 67(0)(C)
Line RHR "B" Test Line E12-F290B-B 67(0)(d)
To Suppr. Pool Inst. Air to ADS P53-F006 70(I) (d)
LPCS Relief Valve E21-F018 71A(0)
Vent Header RHR Pump "C" E12-F025C 71B(0)(d)
Relief Valve Vent Header RHR "C" Relief E12-F406 71B(I)(c)
Valve Vent Hdr.
to Suppr. Pool
& Post-Acc.
Sample Return RHR Shutdown E12-F036 73(0).
Vent Header RHR Shutdown E12-F005 76B(0)
Suction Relief Valve Disch.
RHR Heat Ex. "A" E12-F055A 77(0)(d)
Relief Vent Header RHR Heat Ex. "A" E12-F103A 77(0)(d)
Relief Vent Header RHR Heat Ex. "A" E12-F104A 77(0)(d)
Relief Vent Header 89(I)((c SSW "A" Supply P41-F169A 92(I) C SSW "B" Supply P41-F169B Ctzt. Leak Rate M61-F015 110A(I)
Test Inst.
Ctat. Leak Rate M61-F014 110A(0)
Test Inst.
Ctat. Leak Rate M61-F019 110C(I)
Test Inst.
Ctat. Leak Rate M51-F018 110C(0)
Test Inst.
Ctat. Leak Rate M61-F017 110F(I)
Test Inst.
Ctat. Leak Rate M61-F016 110F(0)
Test Inst, b.
Drywell LPCI "A" E12-F041A 313(I)
LPCI "B" E12-F041B 314(I)
LPCI "B" E12-F236 314(0)
CRD to Rectre.
833-F013A 326(I)
Pump A Seals GRAND GULF-UNIT 1 3/4 6-41 Amendment No. 24 NOV 12 Ei
TABLE 3.6.4-1 (Continued)
CONTAINMENT AND ORYWELL ISOLATION VALVES SYSTEM AND PENETRATION VALVE NUMBER NUMBER Drywell (Continued)
CR0 to Recirc.
B33-F017A 326(0)
Pump A Seals Instrument Air P53-F008 335(I)
Standby Liquid C41-F007 328(I)
Control Standby Liquid C41-F006 328(0)
Control Standby Liquid C41-F218 328(I)
Control-Drain Cont. Cooling P42-F115 329(I)
Water Supply Drywell Chilled P72-F147 332(I)
Water Supply Condensate Flush B33-F204
.333(I)
Conn.
Condensate Flush B33-F205 333(0)
Conn.
Combustible Gas E61-F002A 339(0)
Control Combustible Gas E61-F002B 338(0)
Control Combustible Gas E61-F004A 340(0)
Control Combustible Gas E61-F004B 340(0)
Control Upper Containment G41-F265 342(0)
Pool Drain CRD to Recirc.
B33-F013B 346(I)
Pump B Seals CRD to Recirc.
B33-F017B 346(0)
Pump B Seals Service Air PS2-F196 363(I)
Cont. Leak Rate M61-F021 438A(I)
Test Inst.
Cont. Leak Rate M61-F020 438A(0)
Sys.
BLIND FLANGES Cont. Leak Rate NA 40(I)(0)
Sys.
Cont. Leak Rate NA 82(I)(0)
Sys.
l Containment NA 343(I)(0)
Leak Rate System i
GRAND GULF-UNIT 1 3/4 6-42 Amendment No. 41 1
REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued)
Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR.
The overtravel position fea-ture provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criti-cality after completing CORE ALTERATIONS that could have affected the control rod coupling integrity.
The subsequent check is performed as a backup to the initial demonstration.
In order to ensure that the control rod patterns can be followed and therefore that other parameters are within their limits, the control rod position indication system must be OPERABLE.
The control rod housing support restricts the outward movement of a con-trol rod to less than 3 inches in the event of a housing failure.
The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any dam-age to the primary coolant system.
The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequer.t as to cause excessive wear on the system components.
3/4.1.4 CONTROL R00 PROGRAM CONTROLS The rod withdrawal limiter system input power signal orginates from the first stage turbine pressure.
When operating with the steam bypass valves open, this signal indicates a core power level which is less than the true core power.
Consequently, near the low power setpoint and high power setpoint of the rod pattern control system, the potential exists for nonconservative control rod withdrawals.
Therefore, when operating at a sufficiently high power level, there is a small probability of violating fuel Safety Limits dur-ing a licensing basis rod withdrawal error transient. To ensure that fuel Safety Limits are not violated, this specification prohibits control rod with-drawal when a biased power signal exists and core power exceeds the specified level.
Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth 1
enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident.
The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal.
When THERMAL POWER is greater than 20% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in j
a peak enthalpy of 280 cal /gm.
Thus requiring the rod pattern controller function to be OPERABLE when THERMAL POWER is less than or equal to 20% of RATED THERMAL POWER provides adequate control.
GRAND GULF-UNIT 1 B 3/4 1-3 Amendment No.
20 l MAR 5 IE/
REACTlVfTY LONTROL SYSTEMS BASES CONTROL R00 PROGRAM CONTROLS (Continued)
The RPCS provides automatic supervision to assure that out-of-sequence rods will not be withdrawn or inserted.
A rod is out of sequence if it does not meet the criteria of the Banked Position Withdrawal Sequence as described in the FSAR.
The RPCS function is allowed to be bypassed in the Rod Action Control System (RACS) if necessary, for example, to insert an inoperable con-trol rod, return an out-of-sequence control rod to the proper in-sequence position or move an in-sequence control rod to another in-sequence position.
The requirement that a second qualified individual verify such bypassing and positioning of control rods ensures that the bases for RPCS limitations are not exceeded.
In addition, if THERMAL POWER is below the low power setpoint, additional restrictions are provided when bypassing control rods to ensure operation at all times within the basis of the control rod drop accident analysis.
The analysis of the rod drop accident is presented in Section 15.4 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3.
The RPCS is also designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power dnsity during higher power operation.
A dual c hnnel system is provided that, above the low power setpoint, restricts the withdrawal distances of all non peripheral control rods.
This restriction is greatest at highest power levels.
3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bring-ing the reactor from full power to a cold, xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern.
To meet this objective it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 90 to 120 min-utes.
A minimum available quantity of 4530 gallons of sodium pentaborate solution containing a minimum of 5800 lbs. of sodium pentaborate is required to meet a shutdown requirement of 3%.
There is an additional allowance of 165 ppm in the reactor core to account for imperfect mixing and leakage.
The l
time requirement was selected to override the reactivity insertion rate due to cooldown following the xenon poison peak and the required pumping rate is 41.2 gpm.
The minimum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted.
The tempera-ture requirement is necessary to ensure that the sodium pentaborate remains in solution.
1.
C. J. Paone, R. C. Stirn and J. A. Woolley, "Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NE00-10527, March 1972 2.
C. J. Paone, R. C. Stirn and R. M. Young, Supplement 1 to NE00-10527, July 1972 3.
J. M. Haun, C. J. Paone and R. C. Stirn, Addendum 2, "Exposed Cores,"
Supplement 2 to NEDO-10527, January 1973 GRAND GULF-UNIT 1 B 3/4 1-4 Amendment No. 41
REACTIVITY CONTROL SYSTEMS BASES STANDBY LIQUID CONTROL SYSTEM (Continued)
With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, oper-tion of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.
Relief valves are provided on the SLCS pump discharge piping to protect the SLCS pump and piping from overpressure conditions.
Testing of the relief valve setpoint and verifying that the relief valve does not open during steady state operation of the SLCS pumps demonstrates OPERABILITY of the relief valve.
The relief valves are ASME Class 2 valves and, as such, have a i 3% tolerance in the opening pressure from the set pressure, per the ASME Code (Section III -
Division 1 Subsection NC-7614.2(b), 1974 Edition).
Surveillance requirements are established on a frequency that assures a high reliability of the system.
Once the solution is established, boron con-centration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.
Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.
Compliance with the NRC ATWS Rule 10CFR50.62 has been demonstrated by means of the equivalent control capacity concept using the plant specific minimum parameters.
This concept requires that each boiling water reactor must have a standby liquid control system with a minimum flow capacity and boron content equivalent in control capacity to 86 gpm for 13% weight sodium pentaborate solution (natural boron enrichment) used for the 251-inch diameter reactor vessel studied in NEDE-24222, Reference 4.
The described minimum sys-tem parameters (82.4 gpm, 13.6% weight with natural boron enrichment) provides an equivalent control capacity to the 10CFR 50.62 requirement.
The techniques of the analysis are presented in a licensing topical report NEDE-31096-P, Reference 5.
Only one subsystem is needed to fulfill the system design basis, and two subsystems are needed to fulfill ATWS rule requirements.
An SLCS subsystem consists of the storage tank, one divisional pump, explosive type valve, and associated controls, and other valves, piping, instrumentation, and controls necessary to prepare and inject neutron absorbing solution into the reactor.
4.
"Assessment of BWR Mitigation of ATVS, Volume II," NEDE-24222, December J
1979.
5.
L. B. Claasen et al., "Anticipated Transients Without Scram, Response to NRC ATWS Rule 10CFR50.62," G. E. Licensing Topical Report prepared for the BWR Owners' Group, NEDE-31096-P, December 1985.
GRAND GULF-UNIT 1 B 3/4 1-4a Amendment No. 41
3/4.3 INSTRUMENTATION I
BASES 3/4.3.1. REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates I reactor scram to:
a.
Preserve the integrity of the fuel cladding.
b.
Preserve tr.e integrity of the reactor coolant system.
Minimize the energy which must be absorbed following a loss-of-coolant c.
accident, and d.
Prevent inadvertent criticality.
This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of main-tenance.
When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.
The reactor protection system is made up of two independent trip systems.
There are usually four channels to monitor each parameter with two channels in each trip system.
The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram.
The bases for the trip settings of the RPS are discussed in the bases for Specification 2.2.1.
The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are com-plated within the time limit assumed in the accident analysis.
No credit was taken for those channels with response times indicated as not applicable.
Response time may be demonstrate # by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined.
Sensor response time verificatien may be demonstrated by either (1) inplace, ensite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.
3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to sitigate the consequences of accidents by prescribing the OPERABILITY trip set-points and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance.
Scoe of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety.
Negative barometric pressure fluctuations are accounted for in the trip setpoints and allowable values specified for drywell pressure-high. The setpoints of other
' instrumentation, where only the high or low end of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.
Except:for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the i
sensors are connected.
For D.C. operated valves, a 3 second delay is assumed before the valve starts to move.
For A.C. operated valves, it is. assumed that GRAND GULF-UNIT 1 8 3/4 3-1
INSTRUMENTATION BASES ISOLATION ACTUATION INSTRUMENTATION (continued) the A.C. power supply is lost and is restored by startup of the emergency diesel generators.
In this event, a time of 10 seconds is assumed before the valve starts to move.
In addition to the pipe break, the failure of the D.C. operated valve is assumed; thus the signal delay (sensor response) is concurrent with the 10 second diesel startup.
The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 10 second delay.
It follows that checking the valve speeds and the 10 second time for emergency power establishment will establish the response time for the isolation functions.
However, to enhance overall system relia-bility and to monitor instrument channel response time trends, the isolation actuation instrumentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or greater than the drift allowance assumed for each trip in the safety analyses.
3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control.
This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection.
Negative baro-metric pressure fluctuations are accounted for in the trip setpoints and allow-able values specified for drywell pressure-high.
Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or greater than the drift allowance assumed for each trip in the safety analyses.
3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram recirculation pump trip (ATWS-RPT) system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient.
The response of the plant to this postulated event has been evaluated in General Electric Company report NEDC-32408 dated March 1987.
The results of the analysis show that the Grand Gulf ATWS-RPT design provides adequate protection for these events in which the normal scram paths fail.
The ATVS-RPT provides fully redundant trip of the recirculation pump motors so that the pumps coast down to zero speed.
This trip function reduces core flow creating steam voids in the core, thereby decreasing power generation and limiting any power or pressure excursions.
The Grand Gulf A WS-RPT design provides compliance with the requirements of the NRC ATWS Rule 10CFR50.62.
I GRAND GULF-UNIT 1 B 3/4 3-2 Amendment No. 41
INSTRUMENTATION BASES RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION (Continued)
The ATWS-RPT and Alternate Rod Insertion (ARI) system use common setpoints and trip channels (transmitters and trip systems).
Therefore, the ARI trip function and the RPT trip function will be initiated simultaneously.
The instrumentation setpoints for the RPV pressure and water level trip channels are established such that the normal scram paths for these variables would already be initiated.
The end-of-cycle recirculation pump trip (E0C-RPT) system is a part of the Reactor Protection System and is an essential safety supplement to the reactor trip.
The purpose of the E0C-RPT is to recover the loss of thermal margin which occurs at the end-of-cycle.
The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivity.
Each E0C-RPT system trips both recirculation pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events.
The two events for which the E0C-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.
A fast closure sensor from each of two turbine control valves provides input to the E0C-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second E0C-RPT system.
Similarly, a closure sensor for each of two turbine stop valves providcs input to one E0C-RPT system; a closure sensor from ea:h of the other two stop valves provides input to the other E0C-RPT system.
For each EOC-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves.
The opera-tion of either logic will actuate the E0C-RPT system and trip both recircula-tion pumps.
Each E0C-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled.
The manual bypasses and the automatic Operating Bypass at less than 40% of RATED THERMAL POWER are annunciated in the control room.
The automatic bypass setpoint is feedwater temperature dependent due to the subcooling changes that affect the turbine first-stage pressure-reactor power relationship.
For RATED THERMAL POWER operation with feedwater tempera-ture greater than or equal to 420 F, an allowable setpoint of 1 26.9% of control valve wide open turbine first-stage pressure is provided for the bypass func-tion.
This setpoint is also applicable to operation at less than RATED THERMAL POWER with the correspondingly lower feedwater temperature.
The allowable set-point is reduced to 1 22.5% of control valve wide open turbine first-stage pres-sure for RATED THERMAL POWER operation with a feedwater temperature between 370 F and 420 F.
Similarly, the reduced setpoint is applicable to operation at less than RATED THERMAL POWER with the corresponding lower feedwater temperature.
The E0C-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e., 190 ms.
Included in this time are:
the response time of the sencor, the response time of the system logic and the breaker interruption time.
Breaker interruption time includes both breaker response time and the manufacturer's design arc suppression time of 12 ms.
4 GRAND GULF-UNIT 1 B 3/4 3-3 Amendment No. 41 l
INSTRUMENTAT10N BASES RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION (Continued)
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or greater then the drift allowance assumed for each trip in the safety analyses.
3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without providing actuation of any of the emergency core cooling equipment.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and.the Allowable Value is equal to or greater than the drift allowance assumed for each trip in the safety analyses.
3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the require-ments of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/4.2 Power Distribution Limits.
The trip logic is arranged so that a trip in any one of the inputs will results in a control rod block.
The OPERABILITY of the control rod block instrumentation in OPERATIONAL CONDITION 5 is to provide diversity of rod block protection to the one-rod-out interlock.
GRAND GULF-UNIT 1 B 3/4 3-3a Amendment No. 41 "l
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