Text

Amends 84 & 83 to Licenses DPR-80 & DPR-82,respectively, Incorporating Both Eagle 21 Reactor Protection Sys Upgrade & RTD Bypass Manifold Elimination
	ML20057F466
Person / Time
Site:	Diablo Canyon
Issue date:	10/07/1993
From:	Quay T; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20057F468	List: ML20057F466; ML20057F470;
References
	NUDOCS 9310180043
	Download: ML20057F466 (59)
	v • d • e

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'o UNITED STATES

['3e ~,i NUCLEAR REGULATORY COMMISSION E

WASHINGTON, D. C. 20$55

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PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 84 License No. DPR-80 1.

The Nuclear Regulatory Commission (the Commission) has found that:

f A.

The application for amendment by Pacific Gas & Electric Company (the licensee) dated September 21, 1992, as supplemented February 2, 1993, March 8 and 31, 1993, May 7 and 27, 1993, June 1 and 18, 1993, and August 11 and 27, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I;

i B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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; D.

The issuance of this amendment will not be inimical to the common 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.

3 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. DPR-80 is hereby amended to read as follows:

9310100043 931007 PDR ADOCK 05000275 p

PDR

, (2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 84

, are hereby incorporated in the license.

Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective for Unit.1 after the Eagle 21 reactor protection system upgrade and the resistance temperature detection bypass elimination, to be completed during the IR6 refueling outage that is currently scheduled to begin in February 1994.

FOR THE NUCLEAR REGULATORY COMMISSION d ~~90% of minimum measured~~

~~> 89.7% of minimum measured Tiow++ per loop flow ** per loop j~~

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~~v ti a.~~

~~.M 5~~

a.

h/

~~Hinimum measured flow is 89.000 gpm per loop for Unit I and 90,625 gpm per loop for Unit 2.~~
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.m.-
,e
.r
....mm.m y
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7
TABLL 2.21 (Cont inuedl e
RIACTOR 1 RIP SYST[H INSTRLMNIATION TRIP SETPOINTS TUNCil0NAL_ UNIT 1 RIP _St iP_0 INT ALLOWABLE VALUES n94
~~13. Steam Generator Water 2 7.2% of narrow range instrue nt~~
? 6.8% of narrow range instrument E
level Low-low span cach steam generator span each steam generator g
Coincident with:
M m
a.
RCS Loop aT Equivalent RCS toop al variabic input RCS Loop aT variable input to Power s 50% RTP s 50% RTP e
s 51.5% RTP N
With a time delay (TD) s TD (Note 5) s (1.01)TD (Note 5)
Or b.
RCS Loop aT Equivalent to Power > 50% RTP With no time delay m
~~14. DELETED w~~
~~15. Undervoltage Reactor 1 8050 volts each bus 2 7730 volts each bus l~~
~~l Coolant Pumps~~
~~16. Underfrequency Reactor 2 54.0 Hz each bus 1 53.9 Hz each bus Coolant Punps 37~~
~~17. Turbine Trip 3~~
a.
Low Autostop 011 2 50 psig 9
Pressure 2 45 psig E
b.
Turbine Stop Valve 2 It open
[
Closure 2 It open o
T
~~13. Safety injection input N.A.~~
g from ESF N.A.
g""E
~~19. Reactor Coolant Pump N.A.~~
N.A.
Breaker Position Trip
~~20. Reactor Trip Breakers N.A.~~
N.A.
e g
~~21. Automatic Trip and N.A.~~
N.A.
Interlock Logic
I I
1ABlt 2.2 M Continuedl g;
RCAC10R iP,lP SYSTLH INSTRUMENTATION TRIP SEiPOINTS r
FUNCTIONAL UNIT TRIP SETPOINI ALLOWABLE VALUES n
I
\\
g
~~22. Reactor Trip System Interlocks~~
~~a. Intennediate Range Neutron t 1 x 10 " amps t 6 x 10* amps E~~
~~Flux P 6 Q~~
~~b. Low Power Reactor Trips m~~
Block. P 7 e.
m
~~1) P 10 Input 10% of RATED t 7.9t. 512.1% of RATED l~~
~~TilERML POWER THERML POWER~~
~~2) P 13 Input 5 10t RTP Turbine Impulse 5 12.1% RTP Turbine Impulse i~~
~~Pressure Equivalent Pressure Equivalent l~~
~~c. Power Range Neutron Flux P-8~~
< 35% of RATED
< 37.1% of RATED THERML POWER IHERML POWER
~~d. Power Range Neutron Flux. P-9 5 50% of RATED 5 52.1% of RATED TilERHAL POWER THERML POWER~~
~~e. Power Range Neutron Flux. P-10 10% of RATED~~
~~> 7.93. < 12.1% of RATED THERML POWER THERML E0WER~~
~~f. Turbine Impulse Chamber Pressure. P-13~~
$ IO! RTP Turbine impulse 5 12.1% RTP Turbine impulse l
Pressure Equivalent Pressure Equivalent 1
~~23. Seismic Trip 5 0.35 g 5 0.40 g N~~
?a z
M Ma u
.D e
...____.___.___.=..m
TABLE 2.2-1 Sg REACJOR_iRIP SYSitH INSTRUMENTATION TRIP SETPOINTS G
- - -TABLE NOTATIONS h
NOTE 1:
OVLHTEMPERATURE aT s
aT (1 + r,S )
aT, K,- K, ( 1 + r,5 ) [i T *] + K (P - P")
f,(al) e s
g 1 + r,S 1 + r,5 3
v, e.
Where: 1+rS
= lead-lag compensator on incasured aT I + r,S m
r,. r,
= Time constants utilized in the lead lag controller for aT. r, = 0 seconds, r = 0 seconds g
aT,
= Indicated 4T at RATED THERMAL POWER i
K,
= 1.2 K,
= 0.0182/*F 4
1+rS
= The function generated by the lead lag controller for T, dynamic compensation I + r,S r, r,
= Time constants utilized in the lead lag controller for T,. r, = 30 seconds, r, = 4 seconds i
ej T
= Average temperature. 'F E'
9; O*
I
)
s
TABLE 2.2-1 (Continuedl tog REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS b
1ABll NOTAi!ONS k
NOTE 1: (continued)
<9 T*
= Nominal T, at RATED 1111RML POWER Gc R
= 0.000831/psig 3
d u
P
= Pressurizer pressure psig o.
to P'
= 2235 psig (Nominal RCS operating pressure)
S
= laplace transform operator, s
chamtw,(rs; with gains to be selected based on measured instrument response during plant startup te (i) for q, q, between 192 and +9%. f (al) = 0 (where q, and as are percent RATED TifERML POWER in the top and i
tottom halves of the core respectively, and q, + q, is total THERML POWER in percent of RATED THERML POWER).
(ii) for each percent that the magnitude of (q - q,) exceeds -19t. the aT Trip Setpoint shall be automatically reduced by 2.75% of its value at RATED TlikRML POWER.
(iii) for each percent that the magnitude of (q - q ) exceeds +9t the aT Trip Setpoint shall be automatically reduced by 1.76% of its value at RATED THkRML POWER.
y NOTE 2:
The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 1.0% aT span.
3 E
5
TABLE 2.21 (C_ontinued).
5g RtACTOR 1 RIP SYSTIM INSTRUMENTATION TRIP SETPOINTS Ei IN!LE NOTATIONS n
Eg NOTE 3:
Overpower aT z
ai (1 + r,S) s aT, K K,(
r5
) T - ( [T - T*] - f,(al) 3 p
1 + r,S 1+r5 sn 3
w Where: 1 + r,5
= lead-lag compensator on trasured aT N
1+r5 3
- Time constants utilized in the lead-lag controller for 4T. r, = 0 seconds. r = 0 seconds
~~r., r3 g~~
aT,
= Indicated aT at PATED TilERMAL POWER K,
= 1.072 K,
= 0.0174/*F for increasing average temperature, and 0 for decreasing average temperature r5
= The function generated by the rate-lag controller for T,,, dynamic compensation 3
1+r53 r3
- Time constants utilized in the rate lag controller for T,,,, r = 10 secs.
3 K.
- 0.0014/'F for T > T". and 0 for T s T*
T
= Average temperature. *F
[
T*
= Indicated T,,, at RATED TiiERMAL POWER O
P S
= laplace transform operator, s
f,(al)
= 0 for all al au f?
IABil 2.2 l_(Continuedl f
RfACTOR TRIP SYSTIM INSTRUMENTATION TRIP SETPOINTS b
_1A__B_LE NOTATIO_NS._
n h
N0tt 4:
The thannel's maximum Trip 5ctpoint shall not exceed its computed trip Setpoint by more than 1.0% at span 9
7 NOTE 5:
Steam Generator Water level low low Trip time Delay l
E 3
TD
= [B1(P)' + B2(P)' + B3(P) + B4][0.99]
Where: P
= RCS Loop aT Equivalent to Power (tRTP). P s 50% RTP m
TD
= Time delay for Steam Generator Water level low Low Reactor Trip (in seconds).
B1 = -0.0072 B2 = +0.0181 m
g B3 = -31.72 B4 = +468.8 N
3 Y
a 5
E %i "e
i m
O e
- ~.
,e v
2.2 L7MITING SAFETV SYSTEM SETTfNGS BASES i
2.2.1 REACTORTRIPSYSTEMINSTRUiENTATIONSETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.21 are the nominal values at which the Reactor trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. The various reactor trip circuits automati-cally open the reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level. In addition to redundant channels and trains. the design approach provides a Reactor Trip System which monitors numerous sys-tem variables, therefore, providing protection system functional diversity. The set-point for a reactor trip system or interlock function is considered to be adjusted con-sistent with the nominal value when the *as left' setpoint is within the band allowed for calibration accuracy. There is a band allowed for calibration accuracy only for those setpoints which use analog hardware. For example, the Power Range, Neutron Flux High setpoint is properly adjusted when it is set at 1091 i 0.3% (0.25% of 1201 power span). The setpoints which use digital hardware are set at the nominal value in the
system, The Reactor Trip System initiates a Turbine trip signal whenever Reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.
To accommodate the instrument drift that may occur between operational tests and 1
the accuracy to which setpoints can be measured and calibrated. Allowable Values for the Reactor Trip Setpoints have been specifif in Table 2.2-1.
Operation with a trip set less conservative than its Trip Setpoint, but within its specified Allowable Value, is acceptable.
The methodology to derive the Trip Setpoints is based upon combining all of the untertainties in the channels. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.
Penual Reactor Trio The Reactor Trip System includes manual Reactor trip capability.
Po.er Rence. Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting.
The Low Setpoint trip provides protection during subtritical and low power operations to mitigate the consequences of a power excursion beginning from low power and the High i
Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.
J i
DIABLO CANYON - UNITS 1 & 2 B 2-3 Amendment Nos. 81 &
I
u LIMITING SAFETY SYSTEM SETTINGS BASES Power Range, Neutron Flux (Continued)
The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically re stated below the P-10 setpoint.
Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.
Specifically, this trip complements the Power Range Neutron Flux High and Low
. trips to ensure that the criteria are met for rod ejection from mid power.
The Power Range Negative Rate Trip provides protection for control rod drop-L accidents.
At high power, a rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist.
The Power Range Negative Rate Trip will prevent this from occurring by tripping the reactor.
No credit is taken for operation of the Power Range Negative Rate Trip for those control rod drop accidents for which the DNBRs will be greater than or equal to the DNBR limits.
Intermediate and Source Range, Neutron Flux The Intermediata and Source Range Neutron Flux trips provide core protec-tion durihg reactor STARTUP to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition.
These trips provide redundant protection to the Low Setpoint trip of the Power Range Neutron Flux channels.
The Source Range channels will initiate a Reactor trip i
at about 10+5 counts per second unless manually blocked when P-6 becomes active.
The Intermediate Range channels will initiate a Reactor trip at a current level l
equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.
No credit was taken for operation of the trips asso-ciated with either the Intermediate or Source Range channels in the accident t
analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Trip System.
Overpower AT The Overpower AT trip provides assurance of fuel integrity e.g., no f
fuel pellet cracking or melting, under all possible overpower co,nditions, limits the required range for Overtemperature AT protection, and provides a backup to the High Neutron Flux trip.
The Setpoint is automatically varied i
t i
DIABLO CANYON - UNITS I & 2 B 2-4 Amendment Nos.37and 36 Y Effective at end of Unit 1 Cycle 3 RAY 101389
LIMITING SAFETY SYSTEM SETTINGS BASES Overoower eT (Continued) with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and (2) rate of change at temperature for dynamic compensation for delays associated with flui( transport from the core to the loop temperature detec-tors (RTDs), and thermowell and 1TD iesponse time delays. The Overpower 4T trip pro-vides protection to mitigate the consequences of various size steam breaks as reported in WCAP 9226.
~~Reactor Core Response to Excessive Secondary Steam Releases."~~
Delta T RTP value as,. as used in the Overtemperature and Overpower 4T trips, represents the 1002 measured by the plant for each loop. This normalizes each loop's 4T trips to the actual operating conditions existing at the time of reasurement, thus forcing the trip to reflect the equivalent full power conditions as assumed in the accident anal-yses. These differences in RCS loop 4T can be due to several factors, e.g., measured RCS locp flows greater than minimum measured flow, and slightly asymmetric power distri-butions between quadrants. While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific 4T values. Accurate determination of the loop specific 6T value should be made when performing Incore/Excore quarterly recalibration and under steady state conditions (i.e., power distributions not affected by xenon or other transient conditions).
Pressurizer Pressure In each of the pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pres-sure range in which reactor operation is permitted. The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a less of rea:n r coolant pressure.
On dec. ing power. the Low Setpoint trip is automatically blocked by P 7 (a power level of a;; cximately 10% of RATED THERKAL PDL'ER with turbine impulse chamber pressure at approximately 101 of full power equivalent); and on increasing power, automatically reinstated by P-7.
The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.
Pressurizer Water Level The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power, the Pressurizer High Water Level trip is automatically blocked by P 7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full p3ser equivalent); and on increasing power, automatically reinstated by P-7.
4 t
4e DIABLO CANYON - UNITS 1 & 2 B25 Amendment Nos. 64 & 83
i i
L1HITING SAFETY SYSTEM SETTINGS I
BASES 4
Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.
i 4
On increasing power above P 7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10% of full power equiva-lent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow.
Above P-8 (a power level of approximately 35% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. Conversely on decreasing power between P 8 and P 7 an automatic reactor trip will occur on loss of flow in more than one loop and below P 7 the trip function is automatically blocked.
]
Overtemperature 6T The Overtemperature 4T trip provides core protection to prevent DNB for all combi-nations tJ pressure, power. coolant temperature. and axial power distribution, provided (1) that the transient is slow with respect to delays associated with fluid transport from the core to the loop temperature detectors (RTDs), and thermowell and RTD response time delays, and (2) pressure is within the range between the Pressurizer High and Low l
Pressure trips. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for pi detectors. (2) pressurizer pressure, ping delays from the core to the loop temperature and (3) axial power distribution. With normal i
exial power distribution. this Reactor trip limit is always below the core Safety Limit-as shown in Figure 2.1-1.
If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is i
automatically reduced according to the notations in Table 2.21.
Delta T RTP value as,. as used in the Overtemperature and Overpower 4T trips. represents the 100%
measured by the plant for each loop. This normalizes each loop's AT trips to the actual operating conditions existing at the time of measurement, thus forcing the trip to reflect the equivalent full power conditions as assumed in the accident anal-yses. These differences in RCS loop 4T can be due to several factors e.g., measured RCS loop flows greater than minimum measured flow, and slightly asymmetric power distri-butions between quadrants. While RCS loop flows are not expected to change with cycle life. radial power redistribution between quadrants may occur. resulting in small changes in loop specific 4T values. Accurate determination of the loop specific aT value should be made when performing Incore/Excore quarterly recalibration and under steady state conditions (i.e.. power distributions not affected by xenon or other transient conditions).
t I
i 1
t
{
4 DIABLO CANYON - UNITS 1 & 2 B26 kmtsit fbs. M & 83 t
o LIMITING SAFETY SYSTEM SETTINGS
]
BASES i
Steam Generator Water Level
]
The Steam Generator Water Level Low Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater or a feedwater system pipe break inside or outside of containment.
i This function also provides input to the steam generator level control system. IEEE 279 requirements are satisfied by 2/3 logic for protection function actuation. thus allow-ing for a single failure of a channel and still performing the protection function.
Control / protection interaction is addressed by the use of the Median Signal Selector which prevents a single failure of a channel providing input to the control system requiring protection function action. That is, a single failure of a channel providing input to the control system does not result in the control system initiating a condition requiring protection function action. The Median Signal Selector performs this by ngi selecting the channels indicating the highest or lowest steam generator levels as input to the control system.
The Trip Time Delay (TTD) creates additional operational margin when the plant needs it most, during early escalation to power. by allowing the operator time to recover level when the primary side load is sufficiently small to allow such action.
The TTD is based on continuous monitoring of primary side power through the use of RCS loop e.T.
The magnitude of the delays decreases with increasing primary side power level. up to 50% RTP. Above 50% RTP there are no time delays for the Low Low Level trips.
In the event of failure of a Steam Generator Water Level channel, the channel is placed in the trip condition as input to the Solid State Protection System and does not affect the TTD setpoint calculations for the remaining OPERABLE channels. Failure of the RCS loop 4T channel input to the TTD does not affect the TTD calculation for a pro-tection set. This results in the requirement that the operator adjust the threshold power level for zero seconds time delay from 501 RTP to Of RTP. through the Man Machine interface.
Undervoltece end Underfrecuency Reactor Coolant Pumo Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips provide core protection against DNB as a result of complete loss of forced coolant flow. The speci-fied Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Set-point is reached. Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Peactor trips from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the Reat-ter 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 i
Underfrequency Trip Setpoint is reached shall not exceed 0.3 seconds. On decreasing power, the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automa-tically blocked by P 7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, reinstated automatically by P 7.
L DIABLO CANYON UNITS 1 & 2 B 2-7
/cendment Nos. M & M.12 & hl, 84 & 83
LIMITING SUETY $YSTEM SETTINGS I-BASES Turbine Trio A Turb!n* trip initiates a Reactor trip. On decreasing power, the Turbine trip is automatically blocked by P-S (a power level of approximately 50% of RATED THERMAL POWER); and on increasing power, reinstated automatically by P-9 j
Safety Infection Input from ESF k
i instrueentation, the ESF automatic actuation logic channels will Reactor trip upon any signal which initiates a Safety Injection. The ESF-instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3.
Rea:ter Coolant Pump Breaker Position Trip The Reactor Coolant Pump Breaker Position trip is an anticipatory trip l
which provides score protection against DNB.
The Open/Close Position trip as-sures a reactor trip signal is generated before the Low Flow Trip 5etpoint is reached.
No credit was taken in the safety analyses for operation of this j
trip.
The functional capability at the open/close position settings is re-quired to enhance the overall reliability of the Reactor Trip System.
t Above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10% of full power equivalent) an auto-eatic reactor trip will occcur if more than ane reactor coolant pump breaker is i
opened.
Below P-7 the trip function is a ' mtically blocked.
j i
Rea:ter Trip Syste-Interlocks i
The Reactor Trip System Interlocks perfore the following functions:
{
P-6 On increasing power, P-6 allows the manual block of the Source Range i
trip and de-energizing of the high voltage to the detectors. On de-creasing power, Source Range Level trips are automatically reacti-wated and high voltage restored.
P-7 On increasing power, P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, more than one reactor i
i coolant pump breaker open, reactor coolant pump bus undervoltage and i
underfrequency, pressurizer low pressure and pressuriter high level.
l On decreasing power, the above listed trips are automatically blocked.
~
't DIABLD CANYON - UNITS 1 & 2 B 2-8 Artendment Nos. 30 and 29 S.
1 L
1 TABll 3.3-1 (Continued) cs g
REACTOR TRIP SYSTEM INSTRlHENTATION r-O HINIMUM p
TOTAL NO.
CilAf LS CitANNELS APPLICABLE
~~5 FUNCil0NAL UNIT Of CllANNELS 10 :xlP OPERABjl MODES ACTION 8~~
c-12.
Reactor Coolant Flow low 5
c4 a.
Single Loop 3/ loop 2/ loop in 2/ loop in 1
6 (Above P 8) one loop each loop c
e m
b.
Two Loops 3/ loop 2/ loop in 2/ loop in 1
6 (Above P 7 and below P-8) two loops each loop 13.
Steem Generator Water Level low Low a.
Steam Generator Water 3/S.G.
2/S.G.
2/S.G.
1, 2 6
Level tow-Low in one in each y
S.G.
S.G.
g",
b.
RCS Loop ai 4 (1/ loop) 2 3
1, 2 27 14.
DELETED 15.
Undervoltage-Reactor Coolant 2/ bus 1/ bus 1/ bus 1
28 l
Pumps both busses I
g 16.
Underfrequency-Reactor Coolant 3/ bus 2 on same bus 2/ bus 1
28 i
g.
Punps n,
Po 17.
Turbine Trip a.
Low Autostop 011 Pressure 3
2 2
1 7
ED b.
Turbine Stop Valve Closure 4
4 4
1 7
=e h
g 1 Allt i 1. 3-1 (Cont inucil)
RIACIOR IRIP %Y5itM INSTRilMINIAll0N Q
MINIMilM 10iAl NO.
CIIANNils CHANNils APPLICABLE h
FlMCTIONAL UNIT Of CilANNil$
10 TRIP OPE RAlltI.-
MODES ACTION e
18.
Safety Injection Input c5 from ESF 2
1 2
1, 2 26 a
l 19.
Reactor Coolant Pump Breaker Position Trip above P-7 1/ breaker 2
1/ breaker 1
9 20.
Reactor Trip Breakers 2
1 2
1, 2 10, 12 2
1 2
3*, 4 *, $*
11 f
21.
Automatic Trip and Interlock 2
1 2
1, 2 26 l
Logic 2
1 2
3*, 4*, 5*
11 22.
Reactor Trip System Interlocks
{
a.
Intermediate Range Neutron Flux, P-6 2
1 2
2##
8 b.
Low Power Reactor Trips Block, P-7 P-10 Input 4
2 3
1 M
P-13 Input 2
1 2
1 M
c.
Power Range Neutron Flux, P-8 4
2 3
1 M
[
d.
Power Range Neutron Flux, P-9 4
2 3
1 M
=
&g e.
Power Range Neutron Flux, P-10 4
2 3
1, 2 M
[
f.
Tveine Impulse Chas6er g
Pressure, P-13 (Input to P-7) 2 1
2 1
M
.p 23.
Seismic Trip 3 direc-2/3 loca-2/3 loca-1, 2 13 s.-
tions (x,y,z) tions one tions all in 3 locations' direction directions e
I
..,. - -. ~. _.. _ _
. _ _ -. -,. - - _ _. ~.
TABLE 3.3-1 (Continued)
TABLE NOTATIONS "When the Reactor Trip System breakers are in the closed position and the Control Rod Drive System is capable of rod withdrawal.
~~The provisions of Specification 3.0.4 are not applicable.~~
~~Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.~~
~~Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.~~
ACTION STATEMENTS ACTION 1 -
With the number of channels OPERABLE one less than 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 at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
ACTICN 2 - With the number of OPERABLE channels one less than the Total Number of Cnannels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
The inoperable channel is placed in the tripped condition a.
within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.
The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1, and Either, THERMAL POWER is restricted to less than or equal to 75%
c.
of RATED THERMAL POWER and the Power Range Neutron Flux Trip 5etpcint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUADRANT POWER TILT RATIO is monitored per Specification 4.2.4.2 when THERMAL POWER is greater than or equal to 5D% of RATED THERMAL POWER I
l 1
l i
DI ASLO CANYON - UNITS 1 & 2 3/4 3-5 Amendment Nos. 61 and jj0 MAY 2 31931 i
TABLE 3.3 1 (Continued)
ACTION STATEMENTS (Continued)
/
ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
Below the P 6 (Intermediate Range Neutron Flux Interlock) Setpoint, a.
restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P 6 Setpoint. and b.
Above the P 6 Setpoint, but below 101 of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 101 of RATED THERMAL POWER.
ACTION 4 - With the numbcr of channels OPERABLE one less than the Minimum Channels OPERABLE requirement suspend all operations involving positive reactivity changes.
ACTION 5 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement, verify compliance vith the SHUTDOWN MARGIN require-ments of Specification 3.1.1.1 or 3.1.1.c. as applicable, within I hour and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.
A:T]DN 6 - With the number of OPERABLE channels one less than the Total Number of Channels. STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
The inoperable channel is placed in the tripped condition within a.
6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and b.
The Minimum Channels OPERABLE requirement is met: however, the inoper-able channel or one additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1.
A: TION 7 - With the number of OPERABLE channels less than the Total Number of Channels.
STARTUP and/or PDWER OPERATION may proceed provided the inoperable channel (s) is (are) placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
A: TION B - With less than the Minimum Number of Channels OPERABLE. within I hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.
DIABLD CANYON - UNITS 1 & 2 3/4 3-6 Amendment Nos. 4 and'T, 80 8 83
TABLE 3.3 1 (Continued)
ACTION STATEMENTS (Continued)
ACTION 9 -
With less than the Minimum Number of Channels OPERABLE. operation may continue provided the inoperable channel is placed in the tripped condition within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
ACTION 10 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement be in at least HOT STANDBY within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s: however, one channel 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. provided the other channel is OPERABLE.
ACTION 11 - With the number of OPERABLE channels one less than the Minimum Channels OPEPABLE 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 open the Reactor trip breakers within the next hour.
ACTION 12 - With one of the diverse trip features (Undervoltage or shunt trip attach-rent) inoperable, restore it to OPEPABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the breaker inoperable and apply ACTION 10. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the tire required for performing maintenance to restore the breaker to OPERABLE status.
A: TION 13 With the number of OPEPABLE channels one less than the Total Number of Channels. STARTUP and/or POWER OPEPATION may proceed provided the following conditions are satisfied:
The Minimum Channels OPERABLE requirement is met. and a.
The inoperable channel is placed in the tripped conditions within 6 b.
hours: however. the inoperable channel may se bypassed for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for surveillance testing per Specification 4.3.1.1 or for performing maintenance.
A:TIUN 25 - With the number of OPERABLE channels one less than the Minimum Channels 0;EPABLE requirement, restore the inoperable Channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT STANOBY within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s:
homever. one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.1.1 provided the other channel is OPERABLE.
A:T:CN 27 uth the number of OPERABLE channels one less than the Total Number of Channels. STARTUP and/or POWER OPEPATION may proceed provided that within 6 hurs, for the affected protection set, the Trip Time Delay threshold power level for zero seconds time delay is adjusted to 02 RTP.
ACTION 2E - With the number of OPERABLE channels one less than the Total Number of Channels. STARTUP and/or POWER OPERAT]DN may proceed provided the following conditions are satisfied; The inoperable channel is placed in the trip condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , a.
and b.
The Minimum Channels OPERABLE requirement is met: however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing of other channels per Specification 4.3.1.1.
i DIABLO CANYON - UNITS 1 & 2 3/4 3 7 AmendmentNos.67andh0 El. & B3 '
l
\\
TABLE 3.3 2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME 1.
Manual Reactor Trip N.A.
2.
Power Range. Neutron Flux s 0.5 second"'
3.
Power Range. Neutron Flux.
High Positive Rate N.A.
4.
Power Renge. Neutron Flux.
High Negative Rate s 0.5 second'"
5.
Intermediate Range. Neutron Flux N.A.
6.
Source Range. Neutron Flux 5 0.5 second"'
7.
Overtemperature 4T s 7 seconds"'
B.
Overpower eT s 7 seconds ")
l 9.
Pressurizer Pressure-Low s 2 seconds 10.
Pressurizer Pressure High s 2 seconds II.
Pressurizer Water Level-High N.A.
12.
Reactor Coolant Flow-Low a.
Single Loop (Above P 6) s I second b.
Two Loops (Above P 7 and below P 8) s I second 13.
Steam Generator Water Level Low Low a.
Steam Generator Water Level Low Low s 2 seconds"'
b.
RCS Loop 4T Equivalent Power N.A.
14.
DELETED 15.
Undervoltage Reactor Coolant Pumps s 1.2 seconds 16.
Underfrequency Reactor Coolant Pumps s 0.6 second Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.
Does not include Trip Time Delays. Response times include the transmitters.
Eagle-21 Process Protection cabinets. Solid State Protection System cabinets and actuation devices only. This reflects the response times necessary for THERKAL POWER in excess of 50% RTP.
tenendnent Nos. M & M. M & h, DIABLD CANYDN - UNITS 1 & 2 3/4 3-8 84 & 83 c
i TAftf 3.3-2 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION' RESPONSE TIMESi
\\
F N:'10 sit UNIT J
RESPONSE TIME
7 Tur

.ine Trip 6
a.
Lc= Fluid Oil Pressure N.A.
t.
Turbine Stop Valve N. A.
IE.
Safety Inje: tion Input from ESF N.A.
19.
Rea: tor Coolant Pump Breaker Position Trip N.A.
20.
Rea:ter Trip Break.ers W.A.
i 22.
Aute stic Trip and Interlock Logic N.A.
a II Rea:t:e Tri; Syste-Irtericcks l
N.A.
23.
Se's-4: Trip k.A.
l I
i b
1 i
e l
i l
l I
l DI ABLO CANYON - UNITS 1 & 2 3/4 3-9 Q-I
IAntt 4.3 1 5
le RfAC10R 1 RIP SYSitM INSTRUMENTAil0N SURVElllANCE RIQt11REMENTS G
TRIP g
ACTUATING MODES FOR l
.c CllANNEL Of VICE WHICH 5
CllANNEL CilANNEL OPERAi!ONAL OPERATIONAL ACTUATION SURVEILLANCE 7 FUNCTIONAL _ UNIT CHECK CAllBRATION TEST TEST LOGIC TEST IS RE0VIRED 1.
Manual Reactor Trip N.A.
N.A.
N.A.
R(14)
N.A.
1, 2, 3*, 4*, 5*
G1
[
2.
Power Range. Neutron Flux a.
High Setpoint 5
0(2. 4).
O N.A.
N.A.
1, 2 y
M(3. 4).
0(4. 6),
R(4. S) b.
Low Setpoint 5
R(4)
S/U(1)
N.A.
N.A.
Ifff,2 1
3.
Power Range, Neutron Flux, N.A.
R(4) 0 N.A.
N.A.
~~1. 2 High Positive Rate M~~
4.
Power Range, Neutron Flux.
N.A.
R(4) 0 N.A.
N.A.
1, 2
[
liigh Negative Rate O
S.
Intermediate Range, S
R(4. S)
S/U(1)
N.A.
N.A.
Ifff,2 Neutron Flux 6.
Source Range, Neutron Flux 5
R(4, 5)
S/U(1).0(8)
N.A.
N.A.
2ff. 3. 4. S 7.
Overtemperature aT S
R 0
N.A.
N.A.
1, 2 l
8.
Overpower.T S
R 0
N.A.
H.A.
1, 2 R
9.
Pressurizer Pressure Low S
R 0
N.A.
N.A.
1
@R
~~10. Pressurizer Pressure-High S~~
R 0
N.A.
N.A.
1. 2 8*
~~11. Pressurizer Water level-High 5~~
R 0
N.A.
N.A.
1
~~12. Reactor Coolant Flow low S~~
R 0
N.A.
N.A.
1 92O o
o,,
i G y e
TABLE 4.3-1 REACTOR TRIP SY511H INSTRUMENTATION SURVLill ANCE REQUIREMENTS 5
9 1 RIP 4
ACTUATING H0 DES FOR l
3 CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CliECK CAllBRAT10N 1EST TEST LOGIC TEST IS REQUIRED c5
~~13. Steam Generator Water Level-d low low w~~
e a.
Steam Generator S
R Q
N.A.
N.A.
~~1. 2 m~~
Water level Low low b.
RCS Loop aT N.A.
R Q
N.A.
N.A.
~~1. 2~~
~~14. DELETED~~
~~15. Undervoltage Reactor Coolant N.A.~~
R N.A.
O N.A.
I Puws wE
~~16. Underfrequency Reactor N.A.~~
R N.A.
Q N.A.
1 wg Coolant Pumps
~~17. Turbine Trip a.~~
Low Fluid Oil Pressure N.A.
N.A.
N.A.
S/U(1. 9)
N.A.
I b.
Turbine Stop Valve N.A.
N.A.
N.A.
S/U(1. 9)
N.A.
1 Closure N
6;
~~18. Safety Injection Input from N.A.~~
N.A.
N.A.
R N.A
~~1. 2 cf ESF 3~~
~~19. Reactor Coolant Pu g Breaker N.A.~~
N.A.
N.A.
R N.A.
1 g
Position Trip ID
~~20. Reactor Trip System Interlocks a.~~
Intermediate Range N.A.
R(4)
R N.A.
N.A.
2#
Neutron Flux. P-6 8$
b.
Low Power Reactor Trips Block. P-7 N.A.
R(4)
R N.A.
N.A.
1 W
c.
Power Range Neutron Flux. P 8 N.A.
R(4)
R N.A.
N.A.
I
% C' m
M Tant t 4.31 g
REACIOR_lR IP_SYS II H_I N51RtM N tai _ ION _StlRVL 11 l ANCE_RIQUIREMEN TS E
c' TRIP E
ACillATING HODES FOR G
CllANNEl DiVICE MilCll CllANNIL CilANNTl.
OPIRAiIONAL OPI RATIONAL ACTUATION SURVEILLANCE f FUNCil0NALUNIT CIIICK CAllBRAll0N 1051 1[Si LOGIC TEST
_IS REQUIRED
~~20. Reactor Trip System Interlocks e~~
C (Continued)
U d.
Power Range Neutron m
Flux P 9 N.A.
R(4)
R N.A.
N.A.
I e.
Low Setpoint Power Range N
Neutron Flux, P 10 N.A.
R(4)
R N.A.
N.A.
1, 2 f.
Turbine Im Pressure. pulse Chamber P 13 N.A.
R R
N.A.
N.A.
1
~~21. Reactor Trip Breaker N.A.~~
N.A.
N.A.
M(7. 10)
N.A.
1. 2. 3 *.4 *. 5*
R= 22. Automatic Trip and N.A.
M.A.
M.A.
N.A.
M(7) 1.2.3*.4*.5*
y Interlock logic N
~~23. Seismic Trip N.A.~~
R N.A.
R R
1, 2
~~24. Reactor Trip Bypass N.A N.A.~~
M.A.
H(7.15).R(16)
M.A.
1. 2. 3*. 4 *. 5*
Breaker
,7!?
ira 5&
Q=
O tid *
~~C a~~
r.
S e-
TABLE 4.3 1 (Continued)
TABLE NOTATIONS
?
When the Reactor Trip System breakers are closed and the Control Rod Drive System is capable of rod withdrawal.
f.
Below P 6 (Intermediate Range Neutron Flux Interlock) Setpoint.
Below P 10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
(1) -
If not performed in previous 31 days.
i (2) -
Heat balance only above 15% of RATED THERMAL POWER. During startup in HDDE I above 15% of RATED THERMAL PDWER. the required heat balance shall be performed prior to exceeding 30% of RATED THERKAL POWER. or within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , whichever occurs first. Adjust channel if absolute difference greater than 22. The provisions of Specification 4.0.4 are not applicable for entry into HDDE 2 or 1.
(3) -
Compare incore to excore axial flux difference above 15% of RATED THERMAL P0a'ER at least once per 31 Effective Full Power days. Recalibrate if the absolute difference is greater than or equal to 3t. The provisions of Specification 4.0.4 are not applicable for entry into HDDE 2 or 1.
(c) heutron detectors may be excluded from CHANNEL CALIBRATION.
(E)
Eetector plateau curves shall be obtained and evaluated for the source range 1
re.trcn flux channels. For the Intermediate Range and Power Range Neutron Flus channels a test shall be performed that shows allowed variances of detec-ter voltage do not effect detector operation. For the Intermediate Range and f t.er Renge Neutron Flux Channels the provisions of Specification 4.0.4 are ret appliceble for entry into HDDE 2 or 1.
(f) -
Intere - Excore Calibration, above 75% of RATED THERKAL POWER at least once per 92 Effective Full Power days. The provisions of Specification 4.0.4 are r.ct applicable for entry into HDDE 2 or 1.
(7) -
Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
(E) -
C;arterly Surveillance in HDDES 3*. 4* and S* shall also include verification that permissives P 6 and P 10 are in their required state for existing plant ccr.citions by cbservation of the permissive annunciator window.
(9)
Setpcnnt verification is not applicable.
(20) -
Tne TRIP ACTUATING DEVICE OPERATIONAL TEST shall separately verify the 0FERASILITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.
(II) -
DELETED (12) -
Deleted (13) -
Deleted l
DIABLD CANYDN UNITS 1 & 2 3/4 3-13 AmendmentNos.siandEs, 6 & 83 1
i INSTRUMENTATION 3/4 3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION i
i LIMITING CONDITION FOR OPERATION
3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3 3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3 4 and with RESPONSE TIMES as shown in Table 3.3 5.
t APPLICABILITY: As shown in Table 3.3 3.
5 f
ACTION:
I I
With an ESFAS Instrumentation Channel or Interlock Trip Setpoint less conser-a.
vative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3 4, adjust the Setpoint consistent with the Trip Setpoint value.
b.
With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative i
i than the value shown in the Allowable Value column of Table 3.3 4. declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3 3 until the channel is restored to OPERABLE status with its Trip i
Setpoint adjusted consistent with the Trip Setpoint value.
i l
SURVE1LLANCE REQUIREMENTS e
i 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance of the Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements specified in Table 4.3 2.
F 4.3.2.2 The EN31NEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function j
demonstrated to be within the limit at least once per 18 months.
Each test shall include at least one train such that both trains are tested at least once per 36 months.
i and one channel per function such that all channels are tested at least once per N times 18 mor.ths where N is the total number of redundant channels in a specific ESFAS function 3
as shown in the
~~Total No. of Channels~~
~~column of Table 3.3 3.~~
4 J
I i
i i
i i
i DIABLO CANYON - UNITS 1 & 2 3/4 3 14 Amendment Nos. f4 & E i
i
TABtt 3.3 3 et
[NGINi[RIO SAFETY IEAIURES ACTUAil0N SYSitM INSTRUMENTATION MINIMUM 3
10TAL NO.
CllANNELS CHANNLLS APPLICABLE g
LUNCTIONALUNIT Of CllANNELS TO TRIP OPERABLE MODES ACil0N z
1.
Safety injection (Reactor g
Trip. Feedwater Isoiot b.
O Start Diesel Generators.
d Containment Fan Cooler Units.
and Component Cooling Water) o.
ro a.
Manual Initiation 2
1 2
1,2.3.4 19 b.
Automatic Actuation 2
1 2
1.2.3.4 14 logic and Actuation Relays c.
Containment 3
2 2
1.2.3.4 20 Pressure-liigh w2 w
d.
Pressurizer 4
2 3
~~1. 2. 3#~~
20 g
Pressure. Low e.
DELETED l
f.
Steam Line Pressure-Low 3/ steam line 2/ steam line 2/ steam line
~~1. 2. 3#~~
20 in any steam line 3I 3
?a Es a
=
_TARtt 3.3-3 (Continued)
R,
,g ENGINCIRCO SAFETY FEATURES ACTUAll0N SYSTEM INSTRUM n
)'
6 MINimIM TOTAL NO.
CHANN[l5 CilANNELS APPL!CA8tE FUNCil0NAL UNIT OF CllANNELS 10 1 RIP OPERABLE MODES
_A_C. T ION 5
2.
Containment Spray 3
m d
a.
Manual 2
2 with 2 2
1,2,3,4 19 g
coincident switches I
N b.
Automatic Actuation 2
1 2
1,2,3,4 14 tegic and Actuation Relays
\\
c.
Containment Pressure-4 2
3 1,2,3 17 High-High t'
3.
Containment Isolation s.
Phase "A" Isolation m
1)
Manual 2
1 2
1,2,3,4 19 2)
Automatic Actuation 2 1
2 1,2,3,4 14 i
logic and Actuation Relays 3)
Safety Injection See Item 1. above for all Safety injection initiating functions and requirements.
b.
Phase "B" Isolation 1)
Manual 2
2 with 2 2
1,2,3,4 19 i
coincident switches f
't h
e
.__.. _...___.-_....,___-._ _ _.__ ~.._-_. -._,..____._..._-. _.._. -._,- _ -_ _-.~ _..__ _..__ ~...-___. _...-...,-...-_..~._,__.._ _._..~ ~--._.._.-.-.
IAnti 3.3-antinued).
ENGINEER [D SAFETY FEATURES ACTUATION SYSTIM INSTRUMENTATION e
[W MINIMUM o
TOTAL NO.
CHANNELS CHANNELS APPLICABLE p
MNCTIONALUNIT OF CilANNils TO TRIP OPERABLE MODES ACTION 3.
Containment Isolation (Continued) h 2)
Automatic Actua-2 1
2 1,2,3,4 14 5
tion Logic and d
Actuation Relays M
3)
Containment 4
2 3
1,2,3 17 Pressure-High-High to c.
Containment Ventilation Isolation 1)
Automatic Actua-2 1
2 1,2,3,4 18 tion Logic and o}
Actuation Relays y
2)
Plant Vent Noble 2
1 2
1,2,3,4 18 t-j GasActivity-Higg,)
(RM-14A and 140) 3)
Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.
4)
Containment 2
1 2
1,2,3,4 la Ventilation Ex-haust Radiation-k High (RM-44A and sh 44D)(b)
Ah
$$A 4.
Steam Line Isolation a.
Manual 1 manual 1 manual 1 manual 1, 2, 3, 4 24 switch / steam switch / steam switch /
-.$g line line operating steam line if' (a)The requirements for Plant Vent Noble Gas Activity-fligh (RM-14A and 148) are not appilcable w**
following installation of RM-44A and 440.
(b)The requirements for Containment Ventilation Exhaust Radiation-liigh (RM-44A and 448) are applicable following installation of RM-44A and 440.
TABLE 3.3 3 (Continued)
E y,
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION G
n HINIMUM c
10TAL NO.
CilANNEL5 CilANN[L5 APPLICABLE g
FUNCil0NAL_ UNIT OF CllANNELS
-10 1 RIP OPERABLE H0 DES-ACTION 4.
Steam Line isolation (Continued)
Ey b.
Automatic Actuation 2
1 2
~~1. 2. 3 22 ta logic and Actuation Relays a.~~
N c.
Containment Pressure-4 2
3
~~1. 2. 3 17 High High d.~~
~~Steam Line Pressure-Low 3/ steam line 2/ steam line 2/ steam line~~
~~1. 2. 3#~~
20 in any steam line w
e.
Negative Steam Line 3/ steam line 2/ steam line 2/ steam line 3ff 20 A
Pressure Ratedligh in any steam w
line w*
5.
Turbine Trip &
Feedwater Isolation a.
Automatic Actuation 2
1 2
~~1. 2 25 Logic and Actuation Relays b.~~
Steam Generator 3/stm. gen.
2/stm. gen.
2/stm. gen.
~~1. 2 20 p~~
Water Level-in any operat-in each operat-g High High ing stm. gen.
ing stm. gen.
kn 5r N
Es B"R
,-ar..
-<n
,,ven--,.e---w.w.-
---,-,,,s.--w
--- - +.
-..-.--w w
,-r, e,.

~~w-r-~~

.nn-----
- ~ -
v--

~~+---w-~~

TABLE _3.3 3 (Continued)
ENGINilRtu SAFETY FEATUPES ACTUATION SYST[H INSTRUMENTATION 6
MINIMUM h
10iAL NO.
CilANNELS CilANNELS APPLICABLE g
FUNCTIONAL _ UNIT OF CilANNELS TO TRIP OPERABLE H0 DES ACTION z
6.
Auxiliary Feedwater C5 Manual Initiation 1 manual 1 manual 1 manual
~~1. 2. 3 24 a.~~
d switch / pump switch / pump switch / pump u
e b.
Automatic Actuation 2
1 2
1,2.3 22 N
Logic and Actuation Relays c.
Stm. Gen. Water level-Low low 1)
Start Motor-Driven Pumps w
a w
a.
Steam Generator 3/S.G.
2/S.G.
2/S.G.
~~1. 2. 3 20 0~~
Water Level-in one in each Low-Low S.G.
S.G.
b.
RCS loop ai 4 (1/ loop) 2 3
1, 2. 3 29 2)
Start Turbine-Driven Pump a.
Steam Generator 3/S.G.
2/S.G.
2/S.G.
1, 2. 3 20 g
Water Level-in one in each g
low Low S.G.
S.G.
b.
RCS loop aT 4 (1/ loop) 2 3
1,2,3 29 z
8 d.
Undervoltage RCP Bus 2/ bus 1/ bus on 1/ bus 1
35 Start Turbine-both busses 22Q Driven Pump
'Jn&
e.
Safety Injection Start See Item 1. above for all Safety Injection initiating functions and 8g Motor Driven Pumps requirements.
.~. -
TABLE __3.3-3 (C_ontinued) es INGINEERED SATETY TEAllHES ACTUATION SYSTLH INSTRUMENTATION G
HINIMUM "p
TOTAL NO.
CllANNELS CilANNELS APPLICABLE FUNCTIONAL UNIT OF CilANNLLS 10 TRIP OPERABLE HODES ACTION 7.
Loss of Power a
s (4.16 kV Emergency Bus Undervoltage) v a.
First level 1, 2. 3. 4 a.
m
~~1) Diesel Start 1/ Bus 1/ Bus 1/ Bus 16~~
~~2) Initiation of Load Shed 2/ Bus 2/ Bus 2/ Bus 16 b.~~
~~Second Level 1.2.3.4~~
~~1) Undervoltage Relays 2/ Bus 2/ Bus 2/ Bus 16~~
~~2) Timers to Start wg Diesel 1/ Bus 1/ Bus 1/ Bus 16~~
~~3) Timers to Shed load 1/ Bus 1/ Bus 1/ Bus 16 8.~~
Engineered Safety Features Actuation System Interlocks a.
Pressurizer Pressure. P-11 3 2
2
~~1. 2. 3 21 b.~~
DELETED f
l c.
Reactor Trip. P 4 2
2 2
~~1. 2, 3 23 e~~
8 f
.I
.a r ----
-.-...--~,-m w.
,--y n-
,.--.--,+m.-.
r y
y
.ww w
v v--
e--
e g.,
TABLE 3.3 3 (Continued)
TABLE NOTATIONS
~~Trip function may be blocked inihis HDDE below the P 11 (Pressurizer Pressure Interlock) Setpoint.~~
~~Trip function automatically blocked above P 11 (Pressurizer Pressure Interlock)~~
Setpoint and may be manually blocked below P 11 when Safety Injection on Steam Line Pressure Low is not blocked.
ACTIONSTATEMENTS ACTION 14 With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months s: however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specifica-i tion 4.3.2.1. provided the other channel is OPERABLE.
ACTION 15 - Deleted ACTION 16 With the number of OPERABLE Channels one less than the Total Number of Channels, declare the affected Emergency Diesel Generator (s) inoperable and comply with the ACTION statements of Specification 3.8.1.1: however, one channel 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 4.3.2.1.
ACTION 17 - With the number of OPERABLE channels one less than the Total Number of Channels. operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met.
One additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1.
ACT]DN 18 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply and exhaust valves (RCV 11.
~~12. FCV 660. 661. 662, 663, 664) are maintained closed.~~
1 1
F j
DIABLO CANYON - UNITS 1 & 2 3/4 3-21 AmendmentNos.ffandh%
M&B
v
=
g I
TABLE 3.3 3 (Continued)
)
ACTION STATEMENTS (Continued)
ACTION 19 - With the number of OPERABLE channels one less than the Minimum Channels 4
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 at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in i
COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .
ACTION 20 With the number of OPERABLE channels one less than the Total Number of Channelt STARTUP and/or POWER OPERATION may proceed provided the following t
conditiens are satisfied:
i The inoperable channel is placed in the tripped condition within a.
6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and t
b.
The Minimum Channels OPERABLE requirement is met: however, the inoper-t able channel or one additional channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specification 4.3.2.1.
ACT]ON 21 - With less than the Minimum Number of Channels OPERABLE. within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months deter-mine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.
ACTION 22 With the number of OPERABLE Channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in
[
at least HDT SHUTDOWN within the following 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s: however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance testing per Specifica-tion 4.3.2.1 provided the other channel is OPERABLE.
ACTION 23 With the number of OPERABLE channels one less than the Total Number of Cha.
restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 
.a at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HDT SHUTDOWN or be 1
within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
ACTION 24 - With the number of OPERABLE channels one less than the Total Number of Channels, 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 declare the associated pump or valve inoperable and take the ACTION required by Specification 3.7.1.5 or 3.7.1.2 as applicable.
ACTION 25 With the number of OPERABLE channels one less than the Minimum Channels i
OPERABLE requirement, restore the inoperable channel to OPERABLE status l
within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HDT STANDBY within the next 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s:
however, one channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance I
testing per Specification 4.3.2.1 provided the other channel is OPERABLE.
l 1
ACT]DN 29 - With the number of OPERABLE channels one less than the Total Number of i
Channels. STARTUP and/cr POWER OPERATION may proceed provided that within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , for the affected protection set. the Trip Time Delay threshold
{
power level for zero seconds time is adjusted to Ot RTP.
ACT]DN 35 - With the number of OPERABLE channels one less than the Total Number of j
Channels. STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
I i
The inoperable channel is placed in the trip condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , j
a.
and s
b.
The Minimum Channels OPERABLE requirement is met: however, the inoperable c.hannel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance i
testing of other channels per Specification 4.3.2.1.
DIABLO CANYOW - UNITS 1 & 2 3/4 3 22 AmendmentNos.$andh, 8: & 83 i
i
. - =.,.-~
J.ABLE 3.3_4 o
g ENGINEERED SAflTY FEATifRES ACTUATION SYSTEM INSTRUMENTAll0N TRIP SETPOINTS FUNCTIONAL UNIT TRIP SfiPOINT All0WABLE VALUES k
1.
Safety Injection (Reactor Trip. Feedwater E
isolation. Start Diesel Generators.
Containment Fan Cooler Units, and g
Component Cooling Water) h a.
Manual Initiation N.A.
N.A m
b.
Automatic Actuatioti logic N.A.
N.A and Actuation Relays m
5 3 psig s 3.3 psig l
c.
Containment Pressure Hi.1h d.
Pressurizer Pressure l.m e 1850 psig e 1844.4 psig l
e.
DELETED l
5 f.
Steam Line Pressure Low e 600 psig (Note 1) e S94.6 psig (Note 1) l Y
[3 C
3
?
a r
EU ee m
/
1ARLE 3.3 4 (Continued) f ENGINEERED SAFETY IEAillRES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS r-FUNCTIONAL UNIT TRIP SETPOINT All0WABLE VALUES n
h 2.
Containment Spray 8
a.
Manual Initiation N.A.
N.A C5 b.
Automatic Actuation Logic and N.A.
N.A d
Actuation Relays o-c.
Containment Pressure liigh High s 22 psig s 22.3 psig l
3.
Containment Isolation a.
Phase *A* Isolation 1)
Hanual N.A.
N.A w
2)
Automatic Actuation logic N.A.
N.A.
A and Actuation Relays w
A 3)
Safety injection See item 1. above for all Safety Injection Trip Setpoints and Allowable Values.
b.
Phase *B* Isolation 1)
Manual N.A.
N.A 2)
Automatic Actuation Logic N.A.
N.A and Actuation Relays 3)
Containment Pressure High liigh s 22 psig s 22.3 psig N.
e.
+
4
.,.. ~,
.s-.--.,.,
. ~
....-.4
TABLE 3.3 4 (Continued)_
h ENGINEERED SATETY FEATURES ACTUATION SYSTEM INSTR))MENTAT10N TRIP SETPOINTS r-
[
FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES b
3.
Containment Isolation (Continued) 2 c.
Containment Ventilation Isolation E
i 0
1)
Automatic Actuation logic N.A.
N.A.
d and Actuation Relays 2)
Plant Vent Noble Gas Per the 00CP Activity'-High (RM-14A y
and 148) "
3)
Safety injection See Item 1. above for all Safety injection Trip Setpoints and Allowable Values.
4)
Containment Ventilation Per Specification 3.3.3.10 Exhaust Radiation High t)
(RM 44A and 44B)'"
4.
Y 4.
Steam Line Isolation a.
Manual N.A.
N.A.
b.
Automatic Actuation Logic N.A.
N.A.
and Actuation Relays l
c.
Containment Pressure High-High s 22 psig s 22.3 psig l
d.
Steam Line Pressure Low e 600 psig (Note 1)
= 594.6 psig (Note 1) i N
-.m BEI (a)The requirements for Plant Vent Noble Gas Activity High (RM 14A and 148) are not applicable following installation of g;g RM-44A and 448.
(b)The requirements for Containment Ventilation Exhaust Radiation High (RM 44A and 448) are applicable following z
si.g l
installation of RM 44A and 448.
C 1ABLE 3.3 4 (Continuedl er ENGINEERED SAFETY TEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES h
e.
Negative Steam Pressure Rate liigh s 100 psi / set s-105.4 psi /sec g
5.
Turbine Trip and Feedwater Isolation Eq a.
Automatic Actuation logic N.A.
N.A.
and Actuation Relays x
e b.
Steam Generator Water level-s 75% of narrow range s 75.5% of narrow range l
m High-High instrument span each steam instrument span each steam generator.
generator.
6.
Auxiliary Feedwater s
a.
Manual N.A.
N.A.
b.
Automatic Actuation logic N.A.
N.A.
y and Actuation Relays y
c.
Steam Generator e 7.2% of narrow range e 6.8% of narrow range l
g Water level Low tow instrument span each instrument span each steam generator.
steam generator.
Coincident with:
~~1) RCS loop aT Equivalent to RCS loop aT variable input RCS loop 4T variable input g~~
Power s 50% RTP s 50% RTP s 51.5% RTP With a time delay (TD)
TD (Note 2) s (1.01)TD (Note 2) 3 Or z
3
~~2) RCS loop aT Equivalent to Power > 50% RTP With no time delay em~~
.h.D d.
Undervoltage RCP e 8050 volts e 7730 volts WM e.
Safety injection See item 1. above for all Safety injection Trip Setpoints and Allowable Values.
B g.
G
.w
.m TABLE 3.3-4 (Cnntinued)
[NGIN[LRED SAT [lY II AIORI.5 AC10All0N SY51If1N51RtMNTAT10N TRIP SETP0lNTS 5
F_UNCil0NAL UNIT TRIP $l! POINT ALLOWABLE VALUES 7.
Loss of Power r-
[
(4.16 kV Emergency Bus y
Undervoltage) f a.
First level e
1)
Diesel Start e 0 volts with a e 0 volts with a s 0.8 second time delay 0.8 second time delay 5
e and and o--
t 25P3 volts with a
> 2583 volts with a 10 second time delay s 10 second time delay 5
N 2)
Initiation of load Shed One relay One relay e 0 volts with a e 0 volts with a 5 4 second time delay 5 4 second time delay and and e 2583 volts with a
> 2583 volts with a 5
s 25 second time delay s 25 second time delay with one relay with one relay w
e 2870 volts, instantaneous e 2870 volts, instantaneous 3::
w b.
Second level 1)
Diesel Start e 3600 volts with a e 3600 volts with a s 10 second time delay s 10 second time delay 2)
Initiatim of Load Shed a 3600 volts with a e 3600 volts with a s 20 second time delay s 20 second time delay l
l R
8.
Engineered Safety features Actuation 3
System Interlocks B
a.
Pressurizer Pressure. P 11 s 1915 psig s 1920.6 psig b.
DELETED
]
c.
Reactor Trip, P 4 N.A.
N.A.
i
'g NOTE 1:
Time constants utilized in the lead lag controller for Steam Pressure - Low are r, = 50 seconds and r, = 5 seconds.
NOTE 2:
Steam Generator Water Level Low Low Trip Time Delay D
TD = [B1(P)* + B2(P)' + B3(P) + B4][0.99]
t Where: P = RCS loop aT Equivalent to Power (2RTP). P s 50% RTP h
TD = Time Delay for Steam Generator Water Level low Low Reactor Trip (in seconds) e r,,
Generators affected B1 = -0.0072 B2 = +0.8181 B3 = -31.72 B4 = +468.8 i
~ ~. - -.,. - - - -
n
- ~,,,,
+--+--,,,,-~-----,,-------~~~----+,--+--,--,ne-----n---~--.
, - - --e
-,--.-,m--,;---+n-r:-,-- -.,....
--.~,.+.-,.n.-- -,
~-r------ - - --, -.
l i
TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TINES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS
/
1.
Manual Initiation a.
Safety Injection (ECCS)
N.A.
1)
Feedwater Isolation N.A.
2)
Reactor Trip N.A.
3)
Phase "A" Isolation N.A.
4)
Containment Ventilation Isolation N.A.
5)
Auxiliary Feedwater N.A.
6)
Component Cooling Water N.A.
7)
Containment Fan Cooler Units N.A.
8)
Auxiliary Saltwater Pumps N.A.
b.
Phase "B" Isolation 1)
Containment Spray (Coincident with SI Signal)
N.A.
2)
Containment Ventilation Isolation N.A.
c.
Phase "A" Isolation 1)
Containment Venti ation Isolation N.A.
d.
Steam Line Isolation N.A.
2.
Containment Pressure-High a.
Safety Injection (ECCS) s 27'"/25
1)
Reactor Trip s2 2)
Feedwater Isolation s 63 l
3)
Phase "A" Isolation s 18"'/28'8' 4)
Containment Ventilation Isolation N.A.
5)
Auxiliary Feedwater s 60'88 6)
Component Cooling Water s 38"'/48'8' 7)
Containment Fan Cooler Units s 40'8 8)
Auxiliary Saltwater Pumps s 48"'/58'8' 3.
Pressurizer Pressure-Low a.
Safety Injection (ECCS) s 2 7'"/2 5/3 5"'
1)
Reactor Trip s2 2)
Feedwater Isolation s 63 l
3)
Phase "A" Isolation s 18"8 4)
Containment Ventilation Isolation N.A.
t 5)
Auxiliary Feedwater s 60
6)
Component Cooling Water s 48'8' s 40'8,/38"8 i
7)
Containment Fan Cooler Units 8)
Auxiliary Saltwater Pumps s 58'8'/48"'
DIABLO CANYON - UNITS 1 & 2 3/4 3-28 Amendment Nos. W & M, M &, M, 77 & 7 IDE 1 5. iE3 1
i 1
^*
TASLE 3.3-5 (Continued)
ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS i
4.
Negative Steam Line Pressure Rate High a.
Steam Line Isolation s8 5.
DELETED 6.
Steam Line Pressure-Low a.
Safety Injection (ECCS) s 25'"/35"
1)
Reactor Trip s2 2)
Feedwater Isolation s 63 3)
Phase *A* Isolation s 18("/28"'
4)
Containment Ventilation Isolation N.A.
5)
Auxiliary Feedwater s 60'"
6)
Component Cooling Water s 38!"/48""
7)
Containment Fan Cooler Units s 40"'
8)
Auxiliary Saltwater Pumps s 4BJ"/58"'
b.
Steam Line Isolation s8 l
k DIABLO C3.NYON - UNITS 1 & 2 3/4 3 29 Amendment Nos. iHE and 667 MSW
TABLE 3.3 5 (Continued)
EN31NEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SEC0roS 7.
Containment Pressure-High High a.
Containment Spray s 48.5"'
b.
Phase ~B" Isolation N.A.
c.
Steam Line Isolation s7 8.
Steam Generator Water Level-High High a.
Turbine Trip s 2.5 b.
Feedwater Isolation s 66 l
t 9.
Steam Generator Water Level Low Low a.
Motor-Driven Auxiliary Feedwater Pumps s 60""8)
J b.
Turbine Driven Auxiliary I
i Feedwater Pump s 60
i
~~10. RCP Bus Undervoltage l~~
~~Turbine Driven Auxiliary Feed~~
~~ster Pump 5 60 t~~
II. Plant Vent Noble Gas Activity-High"'
Containment Ventilation Isolation s 11 3?. Containment Ventilation Exhaust Radiation-High' Centainment Ventilation Isolation s 11 (a)The reau1rements for Plant Vent Noble Gas Activity High are not applicable follow 1ng installation of RM 44A and 44B.
(b)The requirements for Containment Ventilation Exhaust Radiation-High are applicable following installation of RM 44A and 448.
i i
DIABLO CANYON - UNITS 1 & 2 3/4 3-30 Amendment Nos. M & 69, fB & M, 84 & 83
I TABLE 3.3 5 (Continued) j TABLE NOTATIONS (1) Diesel generator starting delay not included because offsite power available.
(2) Notation deleted.
i (3) Diesel generator starting and loading delays included.
r (4) Diesel generator starting delay not included because offsite power is available.
Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pg. s (where applicable). Sequen-tial transfer of charging pg suction from the VCT to the RWST (RWST valves open.
then VCT valves close) is included.
(5) Diesel generator starting and sequence loading delays included. Offsite power is not available. Response time limit includes opening of valves to establish SI path and attainrent of discharge pressure for centrifugal charging pumps. Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves open -then VCT valves close) is included.
(6) The maximum response time of 48.5 seconds is the time from when the containment pressure exceeds the High-High Setpoint until.the spray pump is started and the discharge valve travels to the fully open position assuming off site power is not available. The time of 48.5 seconds includes the 28 second maximum delay related to ESF loading sequence. Spray riser piping fill time is not included. The 80-second maximum spray delay time does not include the tine from LOCA start to *P" signal.
i (7) Diesei generator starting and sequence loading delays included. Sequential trans-fer of charging pump suction from the VCT to the RWST (RWST valves open, then VCT
^
4 t
valves close) is not included. Response time limit includes coening of valves to establish 51 flow path and attainment of discharge pressure fc centrifugal l
charging pumps. SI. and RHR pumps (where applicable).
L (B) Does not include Trip Tine Delays. Response times include the transmitters. Eagle-21 Process Protection cabinets. Solid State Protection System cebinets and actua-i i
tien devices only. This reflects the response tines necessary for THERMAL POWER in excess of 50 RTP.
i a
1
?
)
i i
l.
.1 DIABLO CANYON - UNITS 1 & 2 3/4 3 31 Amendment Nos. M & 69. M & k uas
-.,m
-c e
e
.-4u-+,
e.
~~--4~~
6
_ TABLE 4.3-2 5
ENGINEERED SArETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENIS o
n 1 RIP E
ACTUATING l
CHANNEL DEVICE H0 DES FOR
~
CHANNEL OPERA-OPERA-HASTER SLAVE WillCH CilANNEL Call-IIONAL TIONAL ACTUATION RELAY RELAY SURVEILEANCE FUNCTIONAL UNIT CHECK _
BRAT 10N TEST 1EST LOGIC TEST TEST TEST IS REQUIRED 1.
Safety injection. (Reactor Trip Feedwater Isolation.
Start Diesel Generators.
Containment Fan Cooler Units, and Conponent Cooling Water) a.
Manual Initiation N.A.
N.A.
N.A.
R N.A.
N.A.
N.A.
~~1. 2. 3. 4 b.~~
Automatic Actuation N.A.
N.A.
N.A.
N.A.
H(1)
H(1)
Q(3) 1.2.3.4 Logic and Actuation Relays E
c.
Containment Pressure-S R
Q N.A.
N.A N.A.
N.A.
~~1. 2. 3, 4 Y~~
High M
d.
Pressurizer Pressure-Low S
R Q
N.A N.A.
N.A.
N.A.
~~1. 2. 3 e.~~
DELETED f.
Steam Line S
R 0
N.A.
N.A.
N.A.
N.A.
~~1. 2. 3~~
]
Pressure Low 2.
Containment Spray L'
3 a.
Manual Initiation N.A.
N.A.
N.A.
R N.A.
N.A.
N.A.
~~1. 2, 3, 4 ft b.~~
Automatic Actuation N.A.
N.A.
N.A.
N.A.
M(1)
M(1)
Q
~~1. 2. 3. 4 logic and Actuation Relays v.~~
c.
Containment Pressure-5 R
Q N.A.
N.A.
N.A.
N.A.
~~1. 2. 3 High-High Ba e~~
.o_.
TABLL 4.3 2 (_ Continued) h ENGINi[ RED SATETY TEATURES ACTUATION SYSilM INSTRUMENTATION o
Si1RVllllANCE REQUIREHLNTS n
h TRIP o
ACTUATING l
CitANNEL DEVICE MODES FOR c
CilANNEL OPERA-OPERA-MASTER SLAVE WHICH 5
CHANNEL CALI-T10NAL TIONAL ACTUATION RELAY RELAY SURVEILLANCE Gt FUNCTIONAL UNIT CHECK BRt. TION TEST TEST LOGIC TEST TEST TEST IS REQUIRED _
~
3.
Containment Isolation m
a.
Phase "A' Isolation
~~1) Manual N.A.~~
N.A.
N.A.
R N.A.
N.A.
N.A.
~~1. 2. 3. 4~~
~~2) Automatic Actuation N.A.~~
N.A.
N.A.
N.A.
M(1)
M(1)
Q(3) 1.2,3.4 Logic and Actuation Relays
~~3) Safety Injection See Item 1. above for all Safety injection Surveillance Requirements, i s, b.~~
~~Phase "B" Isolation~~
~~1) Manual N.A.~~
N.A.
N.A.
R N.A.
N.A.
H.A.
1, 2. 3. 4 w2
~~2) Automatic Actuation N.A.~~
N.A.
N.A.
N.A.
M(1)
M(1)
Q
~~1. 2. 3. 4 logic and Actuation w~~
~~g, Relays~~
~~3) Containment 5~~
R Q
N.A.
N.A.
N.A.
N.A.
1, 2. 3 w
Pressure-High-High c.
Containment Ventilation Isolation
~~1) Aetomatic Actuation N.A.~~
N.A.
N.A.
N.A.
M(1)
M(1)
Q 1,2.3,4 Logic and Actuation Relays
~~2) Plant Vent Noble Gas S~~
R M(2)
N.A.
N.A.
N.A.
N.A.
~~1. 2. 3. 4 Activity'High (RM 14A and 148) "~~
%R
~~3) Safety injection See Item 1. above for all Safety injection Surveillance Requirements.~~
=3
~~4) Containment Ventilation Exhaust Radiation High g$,~~
(RM 44A and 448)'"
S R
M(2)
N.A.
N.A.
N.A.
N.A.
1, 2. 3. 4
??S (a) The requirements for Plant Vent Noble Gas Activity-High (RM 14A and 14B) are not appilcable following installation of RM.
44A and 448.
o" (b) The requirements for Containment Ventilation Exhaust Radiation High (RM 44A and 248) are appilcable following installation
/@
of RM 44A and 448.
m
TABLE 4.3 ?_(Continuedl 5
ENGINEER [0 SAFETY IEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE Rl_QUIREMLNIS a
n TRIP G
i ACTUATING l
CHANNEL DEVICE H0 DES FOR CHANNEL OPERA-OPERA-HASTER SLAVE WHICH CHANNEL Call-110NAL Il0NAL ACTUATION RELAY RELAY SURVEILLANCE E
FUNCTIONAL UNIT CHECK BRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED
[<
4.
Steam Line Isolation a.
Manual N.A.
N.A.
N.A.
R N.A.
H.A.
N.A.
1, 2. 3 b.
Automatic Actuation Logic N.A.
N.A.
N.A.
N.A.
H(1)
M(1)
Q 1,2,3 and Actuation helays c.
Containment Pressure-S R
0 N.A.
N.A.
N.A.
N.A.
1, 2, 3 High High d.
Steam Line Pressure Low 5
R Q
N.A.
N.A.
N.A.
N.A.
1, 2, 3 e.
Negative Steam Line S
R Q
N.A.
N.A.
N.A.
N.A.
3(3)
Pressure Rate High w
w*
5.
Turbine Trip and Feedwater Isolation a.
Automatic Actuation N.A.
N.A.
N.A.
N.A.
M(1)
M(1) 0 1, 2 Logic and Actuation Relays b.
Steam Generator Water S
R Q
N.A.
N.A.
N.A.
H.A.
1, 2 Level-High High 6.
Auxiliary Feedwater y
a.
Manual N.A.
N.A.
N.A.
R N.A.
N.A.
N.A.
1, 2, 3 b.
Automatic Actuation N.A.
N.A.
N.A.
N.A.
M(1)
M(1)
Q 1, 2, 3 logic and Actuation Relays z
Y c.
Steam Generator Water gg Level Low Low e o,
~~1) Steam Generator S~~
R Q
N.A.
N.A.
N.A.
N.A.
1, 2, 3 g&
Water Level-Low-low k
~~2) RCS Loop aT N.A.~~
R Q
N.A.
N.A.
N.A.
N.A.
~~1. 2. 3 4~~
.~.
+
o VABLE 4.3 2 (Continuedl o
5
[NGINEERED SArETY FEATURES _ ACTUATION _SYSTIH_ INSTRUMENTATION g
SURVllllANCE RLQUIRLHENI5 S
TRIP 22 ACTUATING l.
3 CHANNEL DEVICE HODES FOR CHANNTL OPERA-OPERA-HASTER SLAVE milch c
CHANNEL Call-TIONAL TIONAL ACTUATION RELAY RELAY SURVEILLANCE F_UNCTIONAL UNIT CHECK BRATION
_ TEST TEST LOGIC TEST TEST TEST IS REQUIRED __
vs x
6.
Auxiliary Feedwater (Continued) o.
m d.
Undervoltage RCP N.A.
R N.A.
R N.A.
N.A.
N.A.
1 e.
Safety Injection See Item 1. above for all Safety injection Surveillance Requirements.
7.
Loss of Power a.
4.16 kV Emergency Bus N.A.
R N.A.
R N.A.
N.A.
N.A.
1, 2, 3, 4 Level 1 w%
w b.
4.16 kV Emergency Bus N.A.
R N.A.
R N.A.
N.A.
N.A.
1, 2. 3, 4 g
level 2 8.
Engineered Safety Feature Actuation System Interlocks a.
Pressurizer Pressure.
N.A.
R Q
N.A.
N.A.
H.A.
N.A.
1, 2, 3 P-11 b.
Deleted a
k l
c.
Reactor Trip, P-4 N.A.
N.A.
N.A.
R N.A.
N.A.
N.A.
1, 2, 3 h
TABLE NOTATIONS k
(1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
(2) For the Plant Vent Activity High monitor only, a CHANNEL FUNCTIONAL TEST shall be perfomed at least once every 31 days.
W$
(3) Trip function automatically blocked above P-11 (Pressurizer Pressure Interlock) setpoint and is automatically blocked below P 11 when Safety Injection on Steam Line Pressure Low is not blocked.
~~o. o, 8 E~~
$T>r
INSTRUMENTATION 3 /4. 3. 3 MON 2 TOR 2NG INSTRUMENTATION RADIAT]ON HONITORING FOR PLANT OPERATIONS LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels for plant operations shown in Table 3.3-6 shall be OPERABLE with their Alarm / Trip Setpoints within i
the specified limits.
~
APPLICABILITY: As shown in Table 3.3-6.
ACTION:
With a radiation monitoring channel Alarm / Trip Setpoint for plant a.
operations exceeding the value shown in Table 3.3-6, adjust the Setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel i
inoperable.
With one or more radiation monitoring channels for plant operations b.
I inoperable, take the ACTION shown in Table 3.3-6.
The provisions of Specification 3.0.3 are not applicable.
c.
l 1
SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrue-entation channel for plant operations shall be cee.onstrated OPERABLE by the perforeance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST for the MODES and at the frequencies shown in Table 4.3-3.
n i
DIABLD CANYON - UNITS 1 & 2 3/4 3-36 Amendment Nos. 55 and 54 t
JUN 111990
'n.
O.
ILTI
JP
3 /4. 3 INSTRUMENTATION BASES 3/a.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Recetor Trip System and Engineered Safety Features Actuation System instrumentation and interlocks ensure that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combina-tion thereof reaches its Setpoint, (2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out of service for testing or main-tenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and Engineered Safety Features instrumentation, and (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP 10271.
Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instru-mentation System.* and supplements to that report. Surveillance intervals and out of-service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System.
The Process Protection System is designed to permit any one channel to be tested and maintained at power in a bypassed mode. If a channel has been bypassed for any purpose, the bypass is continuously indicated in the control room as required by applicable codes and standards. As an alternative to testing in the bypass mode, testing in the trip mode is also possible and permitted.
The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various acticents, events, and transients. Once the required logic combination is corpleted.
the system sends actuation signals to those engineered safety features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss of coolant accident:
(1) safety injection pumps start and automatic valves position. (2) Reactor trip. (3) feedwater isolation. (4) startup of the emergency diesel generators. (5) containment spray purrps start and automatic valves position. (6) containment isolation. (7) steam line isola-tion. (B) Turbine trip, (9) auxiliary feedwater pumps start and automatic valve posi-tion (10) containment fan cooler units start, and (11) component cooling water pumps l
start and automatic valves position.
The Engineered Safety Feature Actuation System Instrumentation Trip Setpoints speci-fied in Table 3.3 4 are the nominal values at which the trips are set for each func-tional unit. If the functional unit is based on analog hardware, the setpoint is con-sidered to be adjusted consistent with the nominal value when the *as left* setpoint is within the band allowed for calibration accuracy. For all setpoints in digital i
i hardware, the setpoints are set at the nominal values.
A DIABLO CANYON UNITS 1 & 2 B 3/4 3 1 Amendment Nos. -E4 and %-
84 & 83
INSTRUMEN'TATION BASES REACTOR PROTECTION S'YSTEM ar.d ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
}NSTRUMENTATION (Continued)
To accommodate the instrunent drift that may occur between oprational tests and the accuracy to which setpoints can be measured and calibrated, Allowable Values for the setpoints have been specified in Table 3.3 4.
Operation with setpoints less conserva-tive than the Trip Setpoint, but within the Allowable Value, is acceptable.
The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channel. Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.
ESF response times specified in Table 3.3 5. which include sequential operation of the RWST and VCT valves (Table Notations 4 and 5), are based on values assuced in the non LOCA safety analyses. These analyses take credit for injection of borated water from the RWST. Injection of borated water is assumed not to occur until the VCT charg-ing pump suction isolation valves are closed following opening of the RWST charging pump suction isolation valves.
When the sequential operation of the RWST and VCT valves is nct included in the response times (Table Notation 7), the values specified are based on the LOCA analyses. The LOCA analyses takes credit for injection flow regardless of the source. Verification of the response times specified in Table 3.3 5 will assure that the assumptions used for the LOCA and non LOCA analyses with respect to the operation of the VCT and RWST valves are valid.
i DIASLO CANYON - UNITS 1 & 2 B 3/4 3 la AmendmentNos.iEEandkkr M&W
1NSTRUMENTAT10N BASES REACTOR FROTECTION SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
+
The Engineered Safety Features Actuation System interlocks perform the following functions:
P4 Reactor tripped Actuates Turbine trip. closes main feedwater valves on T below Setpoint, prevents the opening of the main feedwater valves which we,r,e, closed by a Safety Injection or High Steam Generator Water Level signal, allows Safety Injection block so that components can be reset or tripped.
Reactor not tripped prevents manual block of Safety Injection.
l P 11 On increasing pressurizer pressure. P 11 automatically reinstates Safety Injection actuation on low pressurizer pressure and low steam line pressure and blocks steam line isolation on steam line pressure negative rate high.
If Safety injection on low steam line pressure is manually enabled.- P 11 will autcratically block steam line isolation on steam line pressure negative rate high.
On decreasing pressurizer pressure, P 11 permits the manual block of safety injection on low pressurizer pressure and low steam line pres-sure and automatically enables steam line isolation on steam line pressure l
r1e;stive rate high.
3/4.3.3 MONITORIN3 1NSTRUMENTATION t
3/43.3i CDIATION MONITOR]NG FOR PLANT ODERATIONS The 0?ERASILITY of the radiation monitoring channels ensures that: (1) the ratisticn levels are continually measured in the areas served by the individual channels and (2) the alarm or automatic action is initiated when the radiation level trip se:pcint is exceeded.
3 /4.3 ?. ? v:VABLE ItCORE DETECTDRS Tne 0;ERASILITY of the movable incore detectors with the specified minimum te ;1e ent of ecuipment ensures that the measurements obtained from use of this system at:; ately re; resent the spatial neutron flux distribution of the core. The OPERABILITY i
s cf this system is demonstrated by irradiating each detector used and determining the atteptability of its voltage curve.
For the pur core flux caps, pose of measuring Fe(Z) or F",, a full incore flux map is used.
Quarter-as defined in WCAP 8648 June 1976, may be used in recalibration of the Extore Neutron Flux Detection System, and full incere flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range channel is inoperable.
i i
I' DIASLO CANYDN - UNITS 1&2 B 3/4 3 2 Amendment Nos. 84 &'h3
..

ML20057F466

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