ML20034C649
| ML20034C649 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 04/24/1990 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20034C650 | List: |
| References | |
| NUDOCS 9005040281 | |
| Download: ML20034C649 (9) | |
Text
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UNITED $TATES
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NUCLEAR REGULATORY COMMISSION l
3 WAsHINoTON, o, C,20566 t
S o.....,0 SYSTEM ENERGY RESOURCES. INC.. et al.~
DOCKET'NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 t
AMENDMENT TO FACILITY OPERMING LICENSE l
Amendment No. 67 License No. NPF-29 1.
The Nuclear Regulatory Commission (the Commission) has found that A.
The application for amendment by System Energy Resources, Inc.,
(the licensee), dated June 30, 1988, as supplemented by letter dated February 19, 1990, complies with the standards and requirements of i
the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; l
B.
The facility will operate in conformity with the appitcation, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) 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 Commission's regulations;-
l 0.
The issuance of this amendment will not be inimical to the common l
defense and security or to the health and safety of the public; and E.
The issuance of this' amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
i 9005040281 900424 PDR ADOCK 05000416 P
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2.
Accordingly, the license is anonded by changes to the Technical Specifications, as indicated in the attachment to this license amendmentt and paragraph 2.C.(2) of facility Operating License No. NPF.29 is herehy amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environnentai Protection Plan contained in Appendix B, as revised through Amendment No. 67, are hereby incorporated into this license.
System Energy Resources, Inc. shall operate the facility in accordance with the Technical Specifications and the Environnental Protection Plan..
3.
This license amendment is effective as of its date of issuance.
FOR TliE lilCLEAR REGULATORY C0FMISSION Original Signed By:
Elinor G. Adensam Director Project Directorate 111 Division of Reactor Projects.1/11 Office of Nuclear Reactor Regulation 4
Attachment:
Changes to the Technical Specifica tions Date of Issuance: April 24, 1990
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ATTACMENT TO LICENSE AMENDNENT NO. 67 FACILITY OPERATING LICENSE NO. NPF 29 DOCKET NO. 50-426 Replace the following pages of the Appandix A* Technical Specifications with the attached peces.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Reseve Insert 3/4 3-1 3/4 3 1 3/4 3-5 3/4 3-5 3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-8 B3/4 3-1 B3/4 3 1 B3/4 3-2 83/4 3-2 l
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3/4.3 INSTRUMENTATION i
r 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor' protection system instrumentation channels shown in Table 3.3.1-1 shallibe OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.
l APPLICABILITY:
As shown in Table 3.3.1-1.
ACTION:
With the number of OPERABLE channels.less than required by the Minimum I
a.
OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel and/or that trip system in the tripped condition
- within twelve hours.
The provisions of Specification 3.0.4 are not l
applicable.
b.
With the number of OPEPABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for both trip systems, place at least one trip system ** in the tripped condition within one hour and take the ACTION required by Table 3.3.1-1.
SURVEILLANCE REQUIREMENTS 4.3.1.1 Each reactor protection system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.
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4.3.1.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
4.3.1.3 The REACTOR PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall b,, demonstrated to be within its limit at least once per 18 months.
Each test shall include at least one chan-nel per trip system such that ell channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip system.
- An inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur.
In these cases, the inoperable channel shall be restored to OPERABLE status within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or the ACTION required by l
Table 3.3.1-1 for that Trip Function shall be taken.
- The trip system need not be placed in the tripped condition if this would cause the Trip Function to occur.
When a trip system can be placed in the tripped condition without causing the Trip Function to occur, place-the trip system with the most inoperable channels in the tripped condition; if both systems have the same number of inoperable channels, place either trip system in the tripped condition.
GRAND GULF-UNIT 1 3/4 3-1 Amendment No. 67 t
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TABLE 3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS f
(a) A channel may be placed'in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for l
required surveillance without placing the trip system in the tripped condi-tion provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
(b) The " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn
- per Specification 3.9.2 and shutdown margin demonstrations performed per Specification 3.10.3.
i (c) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.
(d) This function is not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.
(e) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.
(f) This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required.
(g) With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
(h) This function shall be automatically bypassed when operating below the appropriate turbine first stage pressure setpoint of:
(1) < 26.9%** of the value of turbine first-stage pressure at valves wide open (VWO) steam flow when operating with rated feedwater temperature of greater than or equal to 420'F, or i
(2) < 22.5%** of the value of turbine first-stage pressure at VWO steam Tiow when operating with rated feedwater temperature between 370'F and 420'F.
- Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
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- Allowable setpoint values of turbine first-stage pressure equivalent to THERMAL POWER less than 40% of RATED THERMAL POWER.
GRAND GULF-UNIT 1 3/4 3-5 Amendment No. 67
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TABLE 4.3.1.1-1 E
REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIRENENTS Er, CHANNEL OPERATIONAL d.
CHANNEL FUNCTIONAL CHANNEL CONDITIC4tS FOR WHICH g
FUNCTIONAL UNIT CHECK TEST CALIBRATION (3)
SURVEILLANCE REQUIRED 1.
4 a.
Neutron Flux - High S/U,5,(b)
S/U, W R
2 S
W R
3,4,5 b.
Inoperative NA W
NA 2,3,4,5 2.
Average Power Range Monitor:( )
a.
Neutron Flux - High, S/U,5,(b)
S/U, W SA 2
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Setdown S
W SA 3, 5 w1 b.
Flow Biased Simulated Therma 1 Power - High 5, D(h)
Q W(d)(e), SA, R(i) 1 w
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c.
Neutron Flux - High S
Q W
, SA 1
d.
Inoperative NA Q
NA 1, 2, 3, 5 3.
Reactor Vessel Steam Dome l
Pressure - High S
Q R(g) 1,2(j) 4.
Low, level 3 S
Q R(g) 1, 2 k
5.
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R
.High, Level 8 5
Q R
1 S
6.
Main Steam Line Isolation Valve - Closure NA Q
R 1
7.
Main Steam Line Radiation -
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O High 5
Q R
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Drywel? Pressure - Hi6h ~
'S Q
R 1, 2 7
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TABLE 4.3.1.1-1 (Continued)
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REACTOR PROTECTION SYSTEM INSTRUNENTATION SURVEILLANCE REQUIREMENTS o5 5
f CHANNEL OPERATIONAL
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g CHANNEL
'QWCTIONAL_
CHANNEL CONDITIONS FOR WHICH
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FUNCTIONAL UNIT CHECK
,.TtST CALIBMTION' SURVEILLANCE REQUIRED
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Scram Discnarge Voludater-W.-
Level - High i
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Transmitter / Trip Hnit S_
Q R(g) 1, 2, 5 1) yI s
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b.
Float Switch MA Q
R 1,2,5(1) b RI9} ;
10.
Turbine Stop Valve - Closure S
Q I
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- 11. Turbine Control Valve Fast Closure Valve Trip System Oil Pressure - Low S
Q R(g)
I 12.
ReactcVi4 ode Switch X
Shutdown Position NA R
NA 1,2,3,4,5 e.a2 43. Manual Scram NA W
NA 1, 2, 3, 4,5 i
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(a) ~ Neutron detectors may be exC W ed from CHANNEL CALIBRATION.
(b) The IRM and SRM chancels shaltbe determined to overlap for at least 1/2 deccde during ead 3
startup af ter entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be deter-mined to-overlap for atJJeast 1/2 decade during each controlled shutdown, if not performed
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within the previous 7 days.
(c) [DELETEDJ (d) This calibration shall consist of the adjustment of-the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONOITION 1 when THERMAL POWERT25% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater tiiar 2% of RATED THERMAL POWER.
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(e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a g
_ -calibrated flow signal.
( 6 Calibrate trip unit at least once per 92 days.The LPRMs shall be calibrated at least once per 1000 MW i)L n
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(h) Verify measured drive flow to be l?ss than c.: equal to estGlith::d drive flow at the existing flow con-g tros valve position.
(i) This calibration sha1! consist of verifying de u 11 secossimulated thermal power _ time constant.
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. T., ' (j ) Not applicable when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.
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M) Not applicable when DRYWELL INTEGRITY is not required.
s Applicable with any L',ntrol rod witi, drawn. Not applicable to control rods Temoved per Specifica-
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tion 3.9.10.1 or 3.9.10.2.
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3/4'. 3 INSTRUMENTATION BASES
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3/4.3.1 kEACTOR PROTEC' T16N' $YSTEM INSTRUMENTATION
.The reactor protection system automatically initiates i reactor scram to:
Preserve the integrity of the fuel cladding.
a.
b.-
Preserve the integrity of the reactor coolant system.
- w Minimize the. energy which must be absorbed following a loss-of-coolant c.
accident, and d.
Prevent inadvertent criticality.
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l-This specification provides the limiting conditions for operation necess.ary s
to preserve the ability-of the system to perform its intended function even i
during periods when instrunhnt channels may be out of_ service because of main-3 tenance.
When necessary,' one channel may be made inoperable for brief. intervals to conduct required surveillance.
The reactor protection syste.n is made up-of two independent trip systems.
There are usually fouf channels to monitor each, parameter with two channeh.in-each trip system.. The outputs of the channels in,a triy system are combined in a logic so that either channel.will= trip that tri:nsystem.' The tripping ofsbo $,'
trip systems will produce a: reactor scram.
Specified surveillance intervals and surveillance and maintenance outage times havh tiaen determined in i, n accordance with NEDC-30851P, ' Technical Specification Improvement Analys:is for BWR Reactor Protection System,"'as approved by the NRC and-' documented in the SER (letter T, A. Pickens from A. Thadani dated July 15,1987).
The bases for 3
the trip settings of the RPS are discussed in the bases fqr Specification 2.2.1.
1 The measurement of response. tit,t at th:1specifi?dfregoduciesprovides assurance.that the protective functions associated with each channel are com-pleted 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 demonstrated by any series of sequential, overlapping or total channel test measurement, prov[ded such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.-
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3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION e
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This specification ensures' the effectivenew"otIthe instrumentation used to mitigate the consequences of accidents by prer.cribing the OPERABILITY' trip setpoints and response times for isolation of the reactor systems.
When neces-r sary, one channel may be inoperable for brief intervals to conduct fequiredt surveillance.
Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have p substantial effect on safety.
Negative barometric pressure fluctuations are accounted for in the trip setpoints and allowable values specified for drywell pressure-high.
1 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.
i GRAND GULF-UNIT 1 B 3/4 3-1 Amendment No. 67 j
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m INSTRUMENTATION j
L BASES.-
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ISOLATIONACTUATIONINSTRUMENTATQL(cw.inued)
Except:for the MSIVs, the safety
- S sis does not address individual sensor: response times or the respona '. w s of the logic systems to which the la,_
a sensors are connected.- For D.C. operated valves,'a 3 second delay is assumed I p before the valve starts to move.
For A.C. operated valves, it is assumed that i g the A.C. power supply is lost and is restored by startup of the emergency diesel lt generators.
In this event, a_ time of 10 seconds is assumed before the valve b
r starts to move.
In addition to the pipe break,;the failure of the D.C. operated valve is assumed; thus the signal delay (censor response):is concurrent with the 1
10 second diesel startup.
The safety analysis considers an allowable inventory ~
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loss in each case which in turn. determines the valve speed in conjunction with '
' t the 10 second delay.
It follows that. checking the valve speeds and the s
i 10 second time for emergency power establishment will. establish the response 1
i time for the isolation functions.
However, to enhance overall' system relia-bility and to monitor' instrument channel response time trends, the isolation actuation instrsmentation 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 fnitiate 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 Setooint 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 NE00-32408 dated March 1987.
The results of the analysis show that the j
Grand Gulf ATWS-RPT design provides adequate protection _for these events in which the normal scram paths fail, The-ATWS-RPT provides fully redundant trip of the recirculation pump i
motors so that the pumps coast down to zero speed.
This trip function reduces core flow creating steam voidg'sure excursions.in the core, thereby decreasing and limiting any power or pres The Grand Gulf ATWS-RPT design provides compliance with the requirements of the NRC ATWS Rule 10CFR50.62.
GRAND GULF-UNIT 1 8 3/4 3-2 Amendment Fo. 67 1
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