ML20150D563
| ML20150D563 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 06/30/1988 |
| From: | SYSTEM ENERGY RESOURCES, INC. |
| To: | |
| Shared Package | |
| ML19292J046 | List: |
| References | |
| NUDOCS 8807140048 | |
| Download: ML20150D563 (6) | |
Text
NL 6 - 68/O'T 3/4.3 INSTRUMENTATION 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 shall be OPERABLE with the REACTOR PROTECTION SYSTEM RESPONSE TIME as shown in Table 3.3.1-2.
APPLICABILITY:
As shown in Table 3.3.1-1.
ACTION:
a.
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system, place the inoperable channel and/or that trip systam in the tripped condition
- within y hour The provisions of Specification 3.0.4 are not applicable, b.
With the number of OPERABLE 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 CHELK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.
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 be demonstrated to be within its limit at least once per 18 months.
E6ch test shall include at least one chan-nel per trip system such that all 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 shallberestoredtoOPERABLEstatuswithin,27J1oursortheACTIONrequiredby Table 3.3.1-1forthatTripFunctionshallbetakgn.
- 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 bott systems have the same number of inoperable channels, place either trip system in the tripped condition.
GkAND GULF-UNIT 1 3/4 3-1 brne.ndrneni No.
8807140048 880630 PDP ADOCK 05000416 P
NLS -88/o 1 TABLE 3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to ours fo.-
required surveillance without placing the trip system in the tripped c.'ndi-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.
(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 1
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
~ ide open (WO) steam flow when operating with rated feedwater temperature of greater than or equal to 420*F, or (2) < 22.5%** of the value of turbine first-stage pressure at WO steam flow 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.
- Allowable setpoint values of turbine first-stage pressure equivalent to
- THERMAL PQWER less than 40% of RATED THERMAL POWER.
l GRAND GULF-UNIT 1 3/4 3-5 Amendment No. 16 ___
j t
D r- -
(P e'o oo
-4 5"
lAalt 4.3.1.1-1 RtACIOR P90lfC110N SY5ifM INSTRtBEENI,All0N SURYfitt AMC[ Rit}IllRilENTS k
CllANNIL OPtRAll0NAL s
CilANMTt I tNGCll0NAl CllANNft CONDilloni 10R 1AllCit E
flNICll0NAL istli ClelCE
._st cal titRAll0N(*
SIfRViltl ANCE Ril{lilRf D U
l.
Intermediate Range fennitors:
a.
Inestron Ilum - tilgli 5/U.S (h) 5/U. w R
2 5
W R
- 3. 4. 5 i
b.
Inoperative NA W
NA
- 2. 3. 4. 5 III 2.
Average Power Range Moniter:
a.
leestren isus - High, 5/U.S.(b) 5/lf. W 5A 2
Setdsess 5
W SA 3, 5 b.
Flow Blased Slaulated I
'I 54. R 'I I
I Thermal Power - Higen
- 5. D yQ W
U WIdI. 5A I
[
c.
Ilectron Flus - Nigh 5
p{
a 1
4 d.
Insperative 11 4 pq M4
- 1. 2. 3. 5
~
3.
Reactor Vessel Steam Dame I i)
Q R
I. 2 Pressure - High 5
4.
III
!. 2 Law. Level 3 5
q R
2 5.
Reacter vessel Water level -
{
Nigh, tevel 8 5
- Q R
1 3
E.
Main Steam time Isolation l
val.e - Cl.s.re NA sq R
I i 1 7
Main Steam Line Radiation -
g33 g
High 5
gQ R
I. 2 O
8.
Drywell Pressure - tilash 5
- Q R(q)
I. 2( E )
1 l
l 1
2":
r-LP TA8LE 4.3.1.1-1 (Continued) o",
t REACTOR PROTECTION SYSTEM INSTRUNENTATION SURVEILLANCE REQUIRDE cm h
o CHANNEL OPERATI3 ital CHANNEL FUNCTIONAL CHANNEL CONDITI0ft5 FOR 1AIICH FUNCTIONAL UNIT
_ClfECK TEST cat 1 BRA 110N SURVEILLANCE REQUIRED h
9.
Scree Olscharge Vcluse Water g
level - High
-4 a.
Transmitter / Trip Unit 5
I93 III XQ R
1, 2, S b.
Float Switch na 1, 2, SII)
- q R
10.
Turbine Stop Valve - Closure I9}
WQ R
1
- 11. Tur:.ine Control Valve Fast closure Valve Trip System 011 Pressure - Low S
I9)
.Fg R
1 12.
Reactor Mode Swltch Shutdown Position MA R
NA
- 13. Manual Scram 1,2,3,4,5 NA
)
Jt'W NA 1,2,3,4,5 l
(a) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(b)
The IRM and Sitt channels shall be determined to 6verlap for at least 1/2 decade during each startup after entering OPERATIONAL CONCITION 2 and the IRN and APRM channels shall be deter-eined to overlap for at least 1/2 decade during each controlled shutdown, if not performed within the previous 7 days.
(c) (DELETED]
(d)
This calibration shall consist of the adjustment of the APRM channel to confors to the power values calculated by a heat balance during OPERATIONAL CONDIT10001 wtta THERMAL POWER > 25% of RATED THERMAL POWER.
THERMAL POWER. Adjust the APRM channel if the absolute difference is greater tEan 2% of RATED (e)
This calibration sha11 consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.
O (f)
{g The LPRMs shall be calibrated at least once per 1000 ledD/T using the TIP system.
(g)
Calibrate. trip unit at least once per,34f8ays.
,. Cl2.
,, if (h)
Verify measured drive flow to be less than or equal to estab11shed drive flow at the existing flow con-
- 7
- trol valve position.
- < +
(t)
This calibration shall consist of verf fying the 61 I second slaulated thermal power time constant.
Fy (j)
Not appitcable when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.
O *,o (k)
Not appilcable when DRYWELL INTEGRITY is not required, yf (1)
Appilcable with any control rod withdrawn.
g tion 3.9.10.1 or 3.9.10.2.
Not appilcable to control rods removed per Specifica-e4,
O
' NLs - M/o 7 3,/_4. 3 INSTRUMENTATION BASES 3/4.3.1. REACTOR PPOTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor sc. ram to:
Preserve the integrity of the fuel cladding.
a.
b.
Preserve t'he 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 serv' ice because of main-When necessary, one channel may be made inoperable for brief intervals tenance.
to conduct required surveillance.
The reactor protection system is made up of tyn independent trip systems.
There are usually four channels to monitor each nemter with two channels in The outputs of the channels to a trip system are combined each trip system.
in a logic so that either channel will trip what trip system. The tripping of both trip systems will produce a *ector scramMIRe bases for the trip 'Tnser'}'
settings of the '!Pi ar1 hst sW in the ba:es for Spkcification 2.2.1.
The measurement of response time at the sp:cified frequencies provides assurance that the protective functions associated with each channel are com-No credit was plated within the time limit assumed in the accident analysis.
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 seasurement, provided such tests demonstrate the total Sensor response time verification may be channel response time as defined.
demonstrated by either (1) inplace, onsite or offsite test seasurements, 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 mitigate the consequences of accidents by prescribing the OPERABILITY trip set-When necessary, points and respese times for isolation of the reactor systems.one cha Some of the trip settings may have tolerances explicitly stated where both the high Negative and lw values are critical and may have a substantial effect on safety.
barometric pressure fluctuations are accounted for in the trip setpoints and allwable 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 For D.C. operated valves, a 3 second delay is assumed sensors are connected.
For A.C. operated valves, it is _ assumed that befort the valve starts to move.
B 3/4 3-1 kmendmenY Mo.
GRAND GULF-UNIT 1
4-
...~iMCS-8s/o 7 Insert to Bases 3/4.3.1 Specified surveillance, intervals and surveillance-and. maintenance outage times have been determined in accordance with NEDC-30851P,'"Technical Specification ~
Improvement Analyses for BWR Reactor Protection System," as approved by the
^
NRC and documented'in the SER (letter T. A. Pickens from A. Thadani dated
~ JJuly 15, 1987).
h e
J16AECM88060201 - 16~