ML19347F017
| ML19347F017 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/16/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19347F016 | List: |
| References | |
| NUDOCS 8105150092 | |
| Download: ML19347F017 (61) | |
Text
.
[,p $n rety, s
o, UNITED STATES g y) j NUCLEAR REGULATORY COMMISSION gy
...E WASHINGTON, D. C. 20555 V
\\.94/
THE CONNECTICUT LIGHT AND POWER COMPANY, THE HARTFORD ELECTRIC LIGHT COMPANY, WESTERN MASSACHUSETTS ELECTRIC COMPANY, AND NORTHEAST NUCLEAR ENEP.GY COMPANY DOCKET NO. 50-245 MILLSTONE NUCLEAR POWER STATION. UNIT NO. 1 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 76 License No. DPR-21 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Connecticut Light and Power Company, The Hartford Electric Light Company, Western Massachusetts Electric Company, and Northeast Nuclear Energy Company (the licensees) dated September 9,1980 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter 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.
- 81051500 4
. 2.
Accordingly, the license is amended by chances to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Provisional Operating License No. DPR-21 is hereby amended to read as follows:
3.B Technical ' Soeci fica tions The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 76, are hereby incorporated in the license. Northeast Nuclear Energy Company shall operate the facility in accordance with the Technical Specifications.
~
3.
This license amendment is effective as of the date laf its issuance.
FOR THE NUCLEAR REGULATORY COMMISSI0'N i
~
e
. Crutchf eld, C ef Operating Reactors Branen #5 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: April 16,1981 8
e i
1
V ATTACHMENT TO LICENSE AMENDMENT NO. 76 PROVISIONAL OPERATING LICENSE NO. DPR-21 DOCKET N0. 50-245
_P,eplace.the following pages of Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the area of change.
Also replace Appendix B Technical Specification page 5.5-1 erroneously omitted from Amendment 75.
Pages*
I-3 B3/4 1-2 2-3 B3/4 1-8 2-4 B3/4 2-2 2-5 B3/4 3-1 2-6 B3/4 4-1 B3/4 5-1 B3/4 5-2 B2-7 B3/4 5-3 B2-8 B3/4 5-4 3/4'2-2 B3/4 5-6 3/4 2-4 B3/4 6-*
3/4 2-8 B3/4 7-3 3/4 3-2 B3/4 7-8 3/4 3-3 B3/4 7-10 3/4 3-4, B3/4 9-1 3/4 4-2 5-1 3/4 5-1 3/4 5-2 3/4 5-3 3/4 5-4 3/4 5-5 3/4 5-6 3/4 5-7 3/4 5-8 3/4 5-9 3/4 6-10 3/4 7-14 3/4 7-15 3/4 7~17 3/4 9-2 3/4 9-3 3/4 11-1 3/4 12-4 3/4 13-1 The folicwing' overleaf pages are also included:
11, 3/
2-1, 3/2 2-3, i
3.'4 ?-7, 3/2 3-1, 3/2 1-1, ar.d 3/4 5-10.
K.
Operating Operating means that a system or component is perfonning its intended function in. its required manner.
L.
Operating Cycle Interval between the end of one refueling outage and the end of the next subsequent refueling outage.
M.
Fraction of Limiting Power Density The ratio of the lincar heat generation rate (LHGR) existing at a given location to the design LHGR for that bundle type.
Design LHGR's are 13.4 KW/ft for 8x8, 8x8R and P8x8R bundles.
Maximum Fraction of Limiting Power Density The Me.imum Fraction of Limiting Power Density (MFLPD) is the highest value existing in the core of the Fraction of Limiting Power Density (FLPD).
N.
Primary Containment Integrity Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied.
1.
All manual containment isolation valves on lines connecting to the reactor coolant system or containment which are not required to be open during accident conditions are closed.
2.
At least one door in the airlock is closed and sealed.
3.
All automatic containment isolation valves are operable or are deactivated in the isolation position.
4.
All blind flanges and manways are closed.
O.
Protective Instrumentation Definitions 1.
Instrument Channel An instrument channel means an arrangement of a sensor and auxiliary equipment required to generate and transmit to a trip system a single trip signal related to the plant parameter monitored by the instrument channel.
Amendment No. J$, 76 1-3 3
2.
Trip System A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective action. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip systen function.
Initiation of protective function may require the tripping of a single trip system or the coincident tripping of two trip systems.
3 Protective Action
~
Aa action initiated by.the protection syster ethen a limit is exceeded. A protective action can be at a channel or systen level.
4.
Protective function A system protective actien which results from the protective action of the channels monitoring a particular plant condition.
P.
Rated Neutron Flux Rated neutron flux is the neutron flux that corresponds to a steady-state thermal power level of 2011 megawatts.
Q.
Rated Thennal Power Rated thermal power means a steady state thermal power level of 2011 megawatts.
R.
(Left intentionally blank)
S.
(Left intentionally blank) 1-4 r---
-r
SAFETY LIMITS LIMITING SAFETY SYSTEM SETTIf;GS Whentheprocesscomputchisoutof 2.
where:
service, this safety limit shall be
~
assumed to be exceededrif the neutron S = Setting in percent of rated flux exceeds the scram setting thernal power (2011 MWt) established by Specification 2.1.2A and a control rod scram does not W = Total recirculation flow occur.
in percent f design (29.7 x 10- Ibm /hr)
D.
Whenver the reactor is in the cold shutdown condition with irradiated fuel in the reactor The trip setting shall not vessel, the water level shall not be less exceed 90 percent of rated than that corresponding to 12 inches above power during generator the top of the active fuel when it is seated load rejections from an in the core. This level shall be continuously initial enerator power monitored.
greater than 307 MWe.
The APRM scram setdown shall be 90% of rated within 30 seconds after initiation of full load rejection.
b.
In the event of operation with a maximum fraction of limiting power density (HfLPD) greater than the fraction of rated power (FRP), the sstting shall be modified as follows:
FRP S < (0.65 W + 55% [MFLPD]
- wheie, FRP = fraction of rated thermal power (2011 MWt)
Amendment No. 16, g6, 76 2-3
SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS MFLPD = maximum fraction of simiting power density where the limiting power density is 13.4 KW/f t -
for 8x8, 8x8R and P3x8R fuel.
The ratio of.FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual opera-ting value will be used.
c.
During power ascensions with power levels less than or equal to 90%. APkM Flux Scram Trip Setting adjustments may be made as described below, provided that the change in scram setting adjustment is less than 10% and, a notice of the adjusunent is posted on the reactor control panel.
The APRM meter indication is adjusted by:
APRM=(0fRP)P where:
APRM = APRM Meter Indication P
= % Core Thennal Pcwer Amendment No. gg, 76 24
SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS For no combination of loop recircula-tion flow rate and core therwal ponxr shall the APRM flux scram trip.
setting be allowed to exceed 120% of rated thermal power.
2.
APRM Reduced Flux Trip Setting-(Refuel or Startup/Ilot Standby Mode _[
When the mode switch is in the refuel or Start Up/Ilot Standby position, the APRM scram shall be setdown to less.
than or equal to 15% of rated thermal power. The IRM scram trip setting shall not exceed 120/125 of full scale.
B.
1.
APRM Rod Block Trip Setting a.
The APRM rod block trip setting shall be as shown in Figure 2.1.2 and shall be:
(Run Mode)
SRB $ 0.65 W + 42%
where:
SRB = Rod block setting in percent of rated thermal power (2011 MWt).
W
= Total recirculation flow in percent of desig'n
~
6 (29.7 x 10 lbm/hr).
b.
In the event of operation with a maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
Amendment No. J6, JJ, $$, 76 2-5
SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS FRP RB <- (0.65 W + 42%) [1FLTD-]
S where:
FRP = fraction of rated thermal power (2011 MWL)
MFLPD = maximum fraction of limiting power density where the limiting power density is 13.4 KW/f t for 8x8, 8x8R and P8x8R fuel.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value it less than the design value of 1.0, in which case the actual operating value will be used.
c.
During power ascensions with power levels less than or equal to 90%,
APRM Rod Block Trip Setting adjust-nents may be made as described i
below, provided that the change in scram setting adjustment is less than 10% and a notice of the adjust-ment is posted on the reactor con-trol panel:
The APRM meter indication is adjusted by:
APRM = APRM Meter Indication P
= % Core Thermal Power Amendment No. J$, $$, 76 2-6
'O Because the boiling transition correlation is based on a large quantity of full scale data there is a very high confidence that operation of a fuel assembly at the condition of MCPR = 1.07 would not produce boiling transition.
Ilowever, if boiling transition were to occur, clad perforation would not be expected. Cladding temperatures would increase to approximately 1100*F which is below the perforation temperature of the cladding noterial.
This has been verified by tests in the General Electric Test Reactor (GETR) where fuel similar in design to Millstone operated above the critical heat flux for a significant period of time (30 minutes) without clad perforation. Thus, although it is not required to establish the safety limit, additional margin exists between the safety limit and the actual occurrence of loss of cladding integrity. The limit of applicability of the boiling transition correlation is 1400 psia during normal power operation. Ilowever, the reactor pressure is limited as per Specification 2.2.1.
The scram trip setting must be adjusted to ensure that the LilGR transient peak is not increased for any combination of maximum fraction of limiting power density and reactor core thennal power. The scram setting is adjusted in accordance with the fonnula in Specification 2.1.2.A.1,-when the maximum fraction of limiting power density is greater than the fraction of rated power.
If the APRM scram setting should require a change due to an abnonnal peaking condition, it will be done by increasing the APP.M gain and thus reducing the slope and intercept point of the flow referenced scram curve by the reciprocal of the APRM gain change.
At pressures below 800 psia, the core evaluation pressure drop (0 power, 0 flow) is greater than 4.56 psi.
At low power and all flows this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure dr p at low power and all flows will always be greater than 4.56 psi. Analyses show that with a flow of 28 x 10 lbs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus the bundle flow with a 4.56 psi driving head will be greater than 28 x 103 lbs/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern. Full scale A1LAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors this corresponds to a core thennal power of more than 50%.
Thus, a core thennal power limit of 25% for reactor pressures below 800 psia or core flow less than 10% is conservative.
Plant safety analyses have shown that the scrams caused by exceeding any safety setting will assure that the Safety Limit of Specification 2.1.lA or 2.1.18 will not be exceeded.
Scram times are checked periodically to assure the insertion times are adequate. The thennal-power transient resulting whea a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turhine stop valves) does not necessarily cause fuel damage. Ilowever, for this specification a Safety Limit violation i
will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The con-cept of not approaching a Safety Limit provided scram signals are operable.is supported by the extensive l
plant safety analysis.
Amendment No.11 J$, $J, 66, 76 B2-2
{
setting was selected to provide adequate margin from the thennal-hydraulic safety limit and allow operating margin to minimize the frequency of unnecessary scrams.
' Analyses of the limiting transients show that no scram adjustment is required to assure MCPR > l.07 when the transient is initiated from MCPR's specified in Section 3.ll.C.
In order to assure adequate core margin during full load rejections in the event of failure of the select rod insert, it is necessary to reduce the APRM scram trip setting to 90% of rated power following a full load rejection incident. This is necessary because, in the event of failure of the select rod insert to function, the cold feedwater would slowly increase the reactor power level to the scram trip setpoint. A trip setpoint of 90% of rated has been established to provide substantial margin during such an occurrence. The trip setdown is delayed to prevent scram during the initial portion of the transient.
The specified maximum setdown delay of 30 seconds is conservative because the cold feedwater transient does not produce significant 5 creases in reactor power before approximately 60 seconds following the load rejection. Reference Amendment 16 Response to Questions A-12, A-14, A-15, and D-3.
For operation in the refuel or startup/ hot standby modes while the reactor is at low pressure, the APRM reduced flux trip scram setting of 1 15% of rated power provides adequate thennal margin between the maximum power and the safety limit, 25% of rated power.
The margin is adequate to accomnodate anticipated maneuvers associated with power plant startup.
Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder thar that already in the
~
system, temperature coefficients are small and control rod patterns are constrained to be unifonn by operating procedures backed up by the rod worth minimizer. llorth of individual rods is very low in a uniform rod pattern.
Thus, of all possible sources of rmctivity input, uniform control rod withdrawal is the most probable cause of significant power rise.
In an assumed unifonn rod withdrawal approach to the scr.:m level, the APRM system would be mere than adequate to assure a scram before the power could exceed the safety limit. The ApRM reduced trip scram remains active until the mode switch is placed in the run position.
This switch occurs when the reactor pressure is greater than 880 psig.
The IRM trip at 1 120/125 of full scale remains as a backup feature.
Amendment No. J$, 34, $J, gg, 76 B2-6
The analysis to support operation at various powcr and flow relationships has considered operation uith cithcr one or two recirculation pumps.
During steady-state operation with one recirculation pump opera-ting the cqualizar line shall be open. Analysts of trcusients from this opercting condition tre less severe than the same transients from the two pump operation.
B.
APRM Control Rod Block Trips Reactor power level may be varied by moving control rods or by varying the recirculation flow rate.
The APRM system provides a control rod block to prevent rod withdrawal beyond a given point at constant recirculation flow rate and thus to protect against a condition of MCPR < l.07.
This rod block setpoint, which is automatically 'taried with recirculation flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The specified flow variable setpoint provides substantial margin from fuel damage, assuming steady-state operation at the setpoint, over the entire recirculation flow range. The margin to the Safety Limit increases as the flow decreases for the specified trip setting versus flow relationship. Therefore, the worst case MCPR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the in-core LPRM system.
When the maximum fraction of limiting power density exceeds the fraction of rated thermal reactor power, the rod block setting is adjusted in accordance with the formula in Specification 2.1.2.8.
If the APRM rod block setting should require a change due to an abnonnal peaking condition, it will be done by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced rod block curve by the reciprocal of the APRM gain change.
The APRM rod block setpoint is reduced to 612% of rated thennal power with the mode switch in refuel or Startup/ilot Standby position.
C.
Reactor Low Water Level Scram The reactor low water level scram is set at a point which will assure that the water level used in the bases for the safety limit is maintained.
D.
Reactor Low Low Water Level ECCS Initiation Trip Point The emergency core cooling subsystems are designed to provide sufficient cooling to the core to dissipate the decay heat associated with the loss-of-coolant accident and to limit fuel clad temperature to well below the clad melting temperature to assure that core geometry remain; intact and to limit any clad metal-water reaction to less than 1%.
To accomplish this function, the capacity of each emergency core cooling system component was established based an the reactor low low wster level. To lower the set-point of the low water level scram would require an increase in the capacity of each of the ECCS com-ponents. Thus, the reactor vessel low water level scram was set low enough to permit margin for operation, yet will not be set lower because of ECCS capacity requirements.
Amendment No. J$, H, $J, gg, 76 B2-7
o The design of' the ECCS components to meet the above criteria was dependent on three previously set parameters:
the maximum break size, the low water level scram setpoint and the ECCS initiation setpoint.
To raise the ECCS initiation setpoint would be in a safe direction, but it would reduce l To lower the setpoint f or initiation of the ECCS would prevent the ECCS components from meeting their
<iesign criteria.
the nurgin established to prevent actuation of the ECCS during nonnal operation or during normally expected transients.
E.
Turbine Stop Valve Sc' ram The turbine stop valve scram like the load rejection scram anticipates the pressure, neutron flux and heat flux increase caused by the rapid closure of the turbine stop valves and failure of the bypass. With a scram setting _10% of valve closure the resultant increase in sur face heat flux is limited such that MCPR remains above 1.07 even during the worst case transient that assumes the turbine bypass is closed.
This scram is bypassed when turbine steam flow is < 45% of rated, as measured by the turbine first stage pressure.
F.
Turbine Control Valve fast Closure lhe turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and sub-sequent failure of the bypass; i.e., it prevents MCPR from becoming less than 1.06 for this transient.
For the load rejection from 1002 power, the hcat flux increases to only 106.5% of its rated power value which results in only a small decrease in MCPR. This trip is bypassed below a generator output of 307 MWe because, below this power level, the MCPR is greater than 1.01 throughout the transient without the scram.
In order to accommodate the full load rejection capal;ility, this scram trip must be bypassed because it would be actuated and would scram the reactor during load rejections.
This trip is automatically bypassed for a maximum of 260 millisec following initiatien of load rejection. Af ter 260 millisec, the trip is bypassed providing the bypass valves have opened.
If the bypass valves have not opened after 260 millisec, the bypass is removed and the trip is returned to the active condition.
This hypass does not adversely affect plant safety because the primary system pressure is within limits during the worst transient even if this trip fails.
lhere are many other trip functions which protect the system during such transients. Reference Response D-3 of Amendment 16.
Amendment No.J, Ig,4f, 76 B2-8
f LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.2 PROTECIIVE INSTRUMENTATION 4.2 PROTECTIVE INSTRUMENTATION Applicability:
Applicability:
Applies to the plant instrumentation which performs Applies to the surveillance requirements of the a protective function.
instrumentation that performs a protective function.
Objective:
Objec_tive:
To assure the operability of protective lo specify the type and frequency of surveillance instrumentation.
to be applied to protective irstrinnentation.
Specification:
Specification:
A.
Primary Containment Isolation functions The instrumentation to be funct ionally tested and calibrated as indicated in Table 4.2.1.
Wir'n primary containment integrity is required, the limiting conditions of operation for the instrumentation that initi.tes primary contain-ment isolation are given in Table 3.2.1.
B.
Emergency Core Cooling Subsystems Actuation The 1imitin9 conditions for operation for the instrumentation which initiates the emergency core cooling subsystems are given in Table 3.2.2 except as noted in Specification 3.5.F.6.
C.
Control Rod Block Actuation 1.
The limiting conditions of operation for the-instrumentation that initiates control rod block are given in Table 3.2.3.
3/d 2-1
I Allt t: 3.2.1 INSlRUMtitl A110N lilAT IN111 AlfS PRIMARY CONI All!Mflil IS0t All0N fuNCll0NS Minimum Number of l0perable Instrument
' Channels Per Trip System (1)
Instrumentt Ir_ip_ level Setting Action Q) 2 Reactor 1ow Water
_127 inches above top of active fuel A
2 Reactor low Low Water 79 (+4-0) inches above top of active fuel A
2 (4) liigh Drywell Pressare 1 2 psig A
2 (2) (S) liigh flow Main Steamline 120% of rated steam flow B
' 2 of 4 in each of
'lligh Temperature Main 2 subchannels Steamline Tunnel 1 200*f B
2 liigh Radiation Main Steamline Tunnel 1 7 times normal rated power background B
2 Low Pressure Main Steamlines
> 860 psig B
2 iligh Flow Isolation 164 inches > trip setting (water differential C
Condenser Line on steam liiie) > 150 inches.
44 inches > triji setting (water dif ferential on water side) 35 inches.
(1) Whenever primary containment integrity is required, there shall be two operable or tripped trip systems for each function, except for low pressure main steamline which only need be available in the Run position.
(2) Per each steamline.
(3) Action:
If the first column cannot be met for one of the trip systems, that trip system shall be tripped.
If the first colunn cannot be met for both trip systems, the appropriate actions listed below shall be taken:
A.
Initiate an orderly shutdown and have reactor 'in cold shutdawn condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Initiate an orderly load reduction and have reactor in Hot Standby within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
C.
Close isolation valves in isolation condenser system.
(4) May be bypassed when necessary by closing the manual instrument isolation valve for PS-1621, A through D, l
during purging for containment inerting or deinerting.
(S) Minimum number of operable instrument channels per trip system requirement does not have to be met for a steamline if both containment isolation valves in the line are closed.
AmendmentNo.f,76 3/4 2-2
Amendment tio.pf, 7G TABLE 3.2.2
~
INSTRllHENTATION TilAf INITIATES AND CONTROLS Tile [ HERG [NCY CORE COOLING SYSitHS S.
HI::Imum Number of Op:rable Inst. Channels Trip function Trip Level Setting Remarks Par Trip System (1) 2 Reactor Low Low Water 79 (+4-0) inches above 1 - In conjunction with low reactor
.t Level top of active fuel pressure initiates core spray and LPCI.
1.
fi.g 2 - In conjunction with high dry well g,
pressure,120 sec. time delay, and LP core cooling pump interlock initiates auto blowdown.
1
{
i.
3 - Initiates TWCl aid Isolation Condenser.
4 - Initiates starting of diesel generator and gas turbine generator.
2 liigh Drywell Pressure
< 2 Psig 1 - Initiates core spray, LPCI, and FWCI, and SBG15.
2 - In conjunction with low low water level.
120 sec. time delay, and LP core cooling pump interlock initiates auto blowdown.
3 - Initiates starting of diesel and gas turbine generator.
I Reactor Low Pressure 300 Psig < P < 350 Psig 1 - Permissive for opening core spray and Permissive LPCI adelssion valves.
2 - In conjunction with low low reactor water level initiates core spray and
..,;e {
LPCI.
g
'l.l __
I liigh iteactor Pressure
< 1005 Psl9 1 - In conjunction with IS second time l
delay, initia tes Isola tion Condenser.
l Timer, Isolation
< I S s ece nd s 1 - In conjunction with higit reactor Condenser pre;sure, initiates isolation Conlenser.
3/4 2-3
C TABLE 3.2.2 (Continued)
~
If4STRUMENTATION TilAT IfilIIATLS' At4D C0f4 Tit 0LS Tilf LNElttaliCY CORE C00LitlG SYSILMS Minimum fiumber of Opcrable Inst. Channels Trip function Trip Level Setting Raurks Per Trip System (1) 1 Timer Auto Blowdown 1 120 seconds 1 - In conjnaction with low low reactor water level and high drywell pressure ar.d LP core cooling pump interlock.
1 - Prevent inadvertent operation of l
~
~
> 2{j Core ileight i.e.,
_[<[nchgs5 Psig C "EdI"*""' SPCdY-Interlock 4,
2 APR LP Core Cooling 90 < P < 110 Psig 1 - Defer " d actuation pending confirma-Pump Interlock
~
tion of LP core cooling systen operation.
2 sets of 2 Power Available on Normal for 120 Volt 1 - Pennissive for auto-close of emergency (Total)
Emergency Buses undervoltage relays power source on buses.
(Monitor Buses 5 and 6) 2 - Permissive to start core spray and LPCI punps.
2 sets of 7 Loss of Normal Power Normal for 120 Volt 1 - Initiates start of emergency power (Total)
Undervoltage Relays sources.
(Monitor buses 1,2,3 and 4) 2 - Strip loads from buses.
3 - Permissive for emergency power sources to close on buses.
1)
If the first column cannot be met for one of the trip systems, that syston may be tripped.
If the first colunn cannot be met for both trip systems, innediately initiate an orderly shutdown to cold conditions.
3/4 2-4 Anendment do. 76
TABLE.4.2.1 (continued) glNIMUM TEST AND CALIBRATION FREQUENCY FOR CORE COOLING INSTRUMENTATION R00 BLOCKS AND ISOLATIONS instrument Channel Instrument Functional Test (2)
Calibration (2)-
Instrument Check (2)
~
Isolation Condenser Isolation 1.
Steam Line Nigh flow (1)
Once/3 Months (1) 2.
Condensate Line liigh Flow (1)
Once/3 Honths (1)
R actor Building Ventilation and Standby Gas Treatment System Initiation-1.
Ventilation Exhaust Duct and (1)
Once/3 Months Once/Qay Refueling Floor Radiation lioni tors' Air Ejector Off-Gas Isolation 1.
Radiation Monitors (1) (3)
Once/3 Months (4)
Once/ Day Notes:
1)
Initially once per month until exposure hours (M as defined on Figure 4.1.1) is 2.0 X 105, thereafter according ta i figure 4.1.1 with an interval not less than one month nor more than three months. Millstone will use data compiled by Coninonwealth Edison on the Dresden 2 Unit in addition to Hillstone Unit I data.
2)
Functional test calibrations and instrument checks are not required when these instruments are not required to be l
operable or are tripped.
)
3)
This instrumentation is excepted from the functional test definition. The functional test will consist of injecting a simulated electrical signal into the measurement channel. See Note 4.
4)
These instrument channels will be calibrated using simulated electrical signals once every three months.
In addition.
calibration including the sensors will be perfonned during each refueling ou'tage.
1 i
S)
The individual power available on emergency bus relays will be functionally tested at the frequency specified by (1) t above.
A full f unctional test including the actuation of the permissives will be perfonned every refueling outage.
i 6)
This instrumentation is excepted from the functional test definition. The functional test will consist of injecting j
a simulated electrical signal into the measurement channel. Functional tests shall be performed before each startup with a required frequency not to exceed once per week. Calibrations including the sensors will be performed during each refueling outage.
Instrument checks shall be perfonned at least once per day during those periods when the instruments are required'to be operable.
l Anendment No. 34 - Correction - December 4,1978
LIMITING CONDITION FOR OPERATION SilRVEllLANCE REQUIREMEN1 2.
The minimum number of operable instrument channels specified in Table 3.2.3 for the Rod Block Monitor may be reduced by one for maintenance and/or testing for periods not in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 30-day period.
D.
Air Ejector Of f-Gas System 1.
Except as specified in 3.2.D.2 below, both air ejector off-gas system radiation monitors shall be ope'rable during reactor power opera-tion. The trip settings for the nonitors shall be set at a value not to exceed the equivalent of the instantaneous stack release limit specified in Specification 3.8.
The time delay setting for closure of the steam jet-air ejector off-gas isolation valve shall not exceed 15 minutes.
2.
frun and af ter the date that one of the two air ejector off-gas system radiation nonitors is nude or found to be inoperable, reactor power operation is permissible only during the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, provided the inoperable monitor is tripped, unless such system is sooner made operable.
E.
Reactor Building Ventilation Isolation and Standby Gas Treatment System Initiation l
1.
Except as specified in 3.2.E.2 below,four radiation monitors shall be operable at all
- times, i
Amendment No. 76 3/42-8 4
Q LIMITING CONDITION FOR OPERATION SilRVEILLANCE REQlllREMENT 1.3 REACTIVIIY CONTROL 4.3 REACTIVITY CONTROL Appl icability:
Applicabi1ity:
Applies to the operational status of the control rod Applies to the surveillance requirements of the system.
control rod system.
Objec t ive:
Objective:
To assure the ability of the control rod system to To verify the ability of the control rod system control reactivity.
to control reactivity.
Specification:
Specification:
A.
Reactivity Limitations A.
Reactivity Limitations 1.
Reactivity Margin - Core Loading 1.
Reactivity Margin - Core Loading i
The core loading shall be limited to that Suf ficient control rods shall be with-which can be :nade subcritical in the most drawn following a refueling outage when core alterations were perfonned to reactive condition during the operating cycle with the strongest operable control demonstrate with a margin of 0.33% AK rod in its full-out position and all other that the core can be made subcritical operable rods fully inserted, at any time in the subsequent fuel cycle with the strongest operable 2.
Reactivity Margin - Stuck Control Rods control rod fully withdrawn and all other operable rods fully inserted.
Control rod drives which cannot be moved with control rod drive pressure shall be 2.
Reactivity Margin - Stuck Control Rods considered inoperable. The control rod directional control valves for inoperable Each partially or fully withdrawn control rods shall be disarmed electrically-operable control rod shall be exercised and the rods shall be in such positions one notch at least once each week.
This that Specification 3.3.A.1 is met.
In no test shali be perfonned at least once per case shall the number of non-fully inserted 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the event power operation is rods disanned be ornater than eight during continuing with three or more inoperable mendment flo. J, 22 february 26, 1976 3/43-1
LIMITING CONDITION FOR OPERATION SilHVEILI ANCE REQUIREMENT power operation, if a partially or f ully cantrol rods or in the event power opera-withdrawn control rod drive cannot be Lion is continuing with one fully or moved with drive or scram pressure the partially withdrawn rod which cannot be reactor shall be brought to a shutdown moved and for which control rod drive condition within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> unless investiga-mechanism damage has not been ruled out.
tion :1emonstrates that the cause of the The surveillance need not be completed failu.e is not due to a failed control rod within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number of inoperable drive mechanism collect housing.
rods has been reduced to less than three dnd if it has been demonstrated that B.
Control Rod Withdrawal control rod drive mechanism collet housing failure is not the cause of an inenovable 1.
Each control rod shall be coupled to its control rod.
drive or completely inserted and the control rod directional control valves B.
Control Rod Withdrawal _
disarmed electrically. However, for purposes of removal of a control rod drive, 1.
The coupling integrity shall be verified as many as one drive in each quadrant may for each withdrawn control rod as follows:
be uncoupled from its control rod so long as the reactor is not in the shutdown or a.
when the rod is fully withdrawn the refuel condition and Specification 3.3.A.1 first time subsequent to each is met.
refueling outage or af ter maintenance, observe that the drive does not go 2.
The control rod drive housing support to the overtravel position; and system shall be in place during power operation and when the reactor coolant h.
when the rod is withdrawn the first system is pressurized above atmospheric time subsequent to each refueling pressure with fuel in the reactor vessel, outage or af ter maintenance, observe unless all control rods are fully inserted discernible response of.the nuclear and Specification 3.3.A.) is met.
instrumentation, however, for initial l
roos when response is not discernible, subsequent exercising of these rods after the reactor is critical shall be performed to verify instrumenta-tion response.
2.
The control rod drive housing support system shall be inspected after reassembly dnd the results of the inspection shall be recorded.
Amendment No. #," Y, 76
LIMITING CONDITION FOR OPERATION.
SURVEILLANCE REQUIREHENT 3.3.8 Control Rod Withdrawal 4.3.8 Control Rod Withdrawal 3.
Whenever the reactor is in the startup 3.(a) To consider the rod worth minfulzer or run mode below 201 rated thennal operable, the following steps wust power. no cnntrol rods shall be moved be performed:
unless the rod worth minimizer is operable or a second independent (1) The control rod withdrawal operator or engineer veriffes that sequence for the rod worth the operator at the reactor console minimizer conputer shall be is following the control rod pro-verified as correct.
gram. The second operator may be (ii)
The rod work minimizer used as a substitute for an inoper-diagnostic test shall able rod worth minimizer during be sucess fully completed.
a startup only if the rod worth minimizer falls af ter withdrawal of at least twelve control rods.
(iii) Proper annunciation of the select error of at least ona 4.
Control rods shall not be withdrawn out-of-sequence control rod in for startup or refueling unless at each fully inserted group shall least two source range channels be vertfled.
have an observed count rate equal to or greater than three counts *
(iv) The rod block function of ti.e roJ worth mininitzer shall be vertiled per second, by attempting to withdraw an out-of-sequence control rod L.:-
yond the block point.
(b)
If the rod worth minfulzer is inap.rsble while the reactor is in tlie startup or run mode belou 10% rated thems) po.J.-r.
and a second independent operator or engineer is being used, he shall verify that all rod positions are correct prios to conmencing withdrawal of each sta group.
I AmentLient No. 77. M1, Jf, 76 3/4 3-3
O
',URVfitLANCE RIMllRfMINT LIMillNG CONulll0N FOR OPIRA110N 5.
During operation with limiting control 4.
Prior to control rod withdrawal for rod pattersis, as dete sistisied by the startup or during refueling, verify reactor engineer, either:
that at least two source range channels have an oteserved count rate of at a.
Both RilM channels shall be operable; least three counts per second.
or 5.
When a limiting control rod pattern b.
Control rod withdrawal shall be exists, an instrument functional test blocked; or of the HilM shall be perfonned prior to withdrawal of the designated c.
.lhe operating power level shall be rod (s) and daily thereafter.
limited so that the HCPR will remain above 1.06 assuming a single C.
Scram insertion Times error that results in complete withdrawal of any single operable During cach operating cycle, each operable control rod, control rod shall be subjected to scram time tests frons the fully withdrawn position.
C.
Scrain Insertion T imes If testing is not accomplished during reactor power operation, the measured 1.
The average scram insertiori time, based scram insertion times shall be extrapolated l
on the de-energization of the scram pilot to the reactor power operation condition valve solenoids as tinie zero, of all utilizing previously detennined correlationis.
operable control rods in the reactor power operation condition shall be no greater than:
- 7. Inserted from Average Scram fully Withdrawn leisertioni limes (sec) 5 0.375 20 0.900 50 2.000 90 3.500 Amendment No. JS, A/, 76 3/4 3-4 t
LIMITING CONDITION FOR OPERATION SilRVEILLANCE REQlllREt1ENT 3.4 STANDBY LIQUID CONTROL SYSTEli 4.4 SlAflDilY LIQlllD CON 1p0L SY,S,Il f)
Applicab il i ty:
Applicability; Applies to the operating status of the standby Applies to the periodic testing requirements for the liquid control system.
standhy liquid control system.
Objective:
0,tgj ec t iv_e :
To assure the availability of the standby liquid lo verity the operability of the standby liquid control system.
control system.
Specification:
Spec _i f_ica tion :
A.
Normal Operation A.
flormal 0.peration During periods when fuel is in the reactor, the The operability of the standby liquid control liquid poison system shall be operable except system shall be verified by performance of the when the reactor is in the cold shutdown following tests:
condition and all control rods are fully inserted and Specification 3.3. A is met and 1.
At least once per month, except as noted in Specification 3.4.8 below.
Demineralized water shall be recycled to the test tank. pump minimum flow rate of 32 ypm shall be verified against a system head corresponding to a reactor vessel pressure of 1225 psig.
2.
At least once during each operating cyde a.
Manually initiate one.of the two standby liquid control systems and pump demineralized water into the reactor vessel at or near operatinn pressures.
This test checks explosion of the charge associated with the tested system, proper operatinn of the valves and pump capacity.
Ihe 3/4 4-1
LIMITING CONDITION FOR OPERATION SilRVEILLANCE REQUIREMENT o
=
replacenent charges to be ir. stalled will be sel. cted from the same batch e
as those tested. Both systems shall be tested and inspected, including each explosion valve in the course of two opera ting cycles.
'b.
Manually initiate each system, except the explosion valve and pump solution in the recirculation path back to the stos_ge tank.*
c.
Test that the setting of the system pressure relief valves is between 1350 and 1450 psig.
B.
Operation with inoperable Components from and after the date that a redundant component is nude or found to be inoperable.
Specification 3.4.A shall be considered fulfilled, provided that:
1.
The component is returned to an operable condition within 7 days or 2.
A written report shall be suNnitted to the Atomic Energy Coninission when the maintenance to restore the component to an operable condition will last longer than 7 days.
Per errata sheet dtd 10-7-70*
3/4 4-2 Aaendment No. 76 4
LIMITlHG CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.5 CORE AND CONTAINMENT C00!";G SYSTEMS 4.5 ' CORE AND CONTAINMENT COOLING SYSTEH,5 A plicability:
Applicability:
J Applies to the operational status of the emergency Applies to periodic testing of the emergency cooltag subsystems.
cooling subsystens.
Objective:
Objective:
To assure adequate cooling capability for heat To verify the operabill'ty of the core and contain-removal in the event of a loss of coolant accident ment cooling subsystems.
or isolation from the nonnal reactor heat sink.
Specification:
- Specifica tion:
A.
Surveillance of the Core Spray and LPCI Sub.
A.
Core Spray and LPCI Subsystems systems shall be perfonned as follows:
1.
Except as specified in 3.5.A.2, 3.5.A.3, 1.
Core Spray Subsystem Testing:
3.5.F.6, 3.5.F.7 and 3.5.F.8, both core spray subsystems shall be operable Item frequency _
whenever irradiated fuel is in the reactor vessel.
a.
Simulated Automatic Each Refueling Actuation Test Outage b.
Pump and Valve Per Surveillance Operabilii.y Requirement 4.13 Amendment No. Jg. 76 J/4 5-1
LlHITING CONDITION FOR OPERATION SURVEILLANEE REQUIREMENT 2.
From and after the date that one of the core spray subsystems is made or found to be inoperable for any reason, reactor operation is permissible only during the c.
succeeding fifteen days unless such sub-Op instrumentation systen is sooner made operable, provided check Once/ day that dulrbg.such fif teen days all active calibrate Once/3 months canponents of the other core spray sub-test Oncef3 months system and the LPCI subsysten and both emergency power sources required for operation of such conponents if no external source of power were available shall be operable.
3.
From and af ter the date that both core spray subsystems are made or found to be inoperable. for any reason, reactor opera-tion is permissible only during the succeeding seven days unless at least one of such subsystems is sooner nude operable, provided that during such seven days all active components of the LPCI subsystem and both emergency power sources required for operation of such canponents if no external source of power were available shall be operable.
4.
Except as specified in 3.5.A.5, 3.5.B.3,4,5, 3.5.F.6, 3.5.T.7 and 3.5.F.8, the LPCI 2.
LPCI Subsystem Testing shall be as specified subsystem shall be operable whenever in 4.5 A.I.a, b and c except that irradiated fuei is in the reactor vessel.
three LPCI pumps shall deliver at least 15,000 gpm against a system head correspor4-ing to a reactor vesset pressure of 1 14.7 psia.
Amendment No. g, 76 3/4 5-2
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 5.
from and after the date that one of the LPCI pumps is made or found to be inoper-able for any reason, reactor operation is pennissible only during the succeeding 30 days unless such pump is sooner made operable, provided that during such thirty days the remaining active components of the LPCI and containment cooling sub:.ystem and all active components of both core spray subsystems and both energency power sources required for operation of such components if no external source of power 3
During each five-year period, an air were available shall be operable.
test shall be performed on the drywell spray headers and nozzles.
6.
A maximum of one drywell spray loop may B.
Surveillance of the Cor.talmnent Cooling Sub-be inoperable for 30 days when reactor water terperature is greater than 212*F.
systems shall be performed as follows:
7.
If the requirenents of 3.5.A cannot be met, 1.
Emergenc: Service Water Subsystem Testing:
an orderly shutdown of the reactor shall be initiated and the reactor shall be in item frequency the cold shutdown condition within 24 a.
Dump & Valve Per Surveillance hours.
Operability Requirenent 4.13 B.
Containment Cooling Subsystems 1.
Except as specified in 3.5.B.2, 3.5.B.3, 3.5.f.6, 3.5.F.7 and 3.5.F.8, both containment cooling subsystems shall be' operable whenever irradiated fuel is in the reactor vessel.
l Amendment No. 23, 2$. A6, 76
LIMITING CONDITION FOR OPERATION SilRVEILLANCE REQUIREMENT 2.
Fran and af ter the date that one of the emergency service water (ESW) subsysten pumps is made or found'to be inoperable '
for any reason, reactor operation is permissible only during the succeeding thirty days unless pump is sooner made operable, provided that during such thirty days all other active canponents of the containment coolinn system are operable.
3.
f rom and af ter the date that one active canponent in each containment cooling subsystem or a LPCI and ESW in one containment cooling subsystem is nude or sound te be inoperable for any reason, reactor operation is pennissible only during the succeeding 7 days provided the renaining active components in eich containment cooling subsystem, both core spray subsystems and both emergency power sources for operation of such camponents if no external source of powcr were available shall be operable.
4.
From and af ter the date that one LPCI and one ESW pump in each containment cooling subsystem is made or found to be inoperable for any reason, reactor opera-tion is pennissible only during the succeeding four days provided the remain-ing active components of the containment
- ooling subsystems, both core spray sub-systems and both emergency power sources for operation of such components if no external source of power were available, shall be operable.
3/4 6-4 Amendment No. 76 e
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 5.
From and af ter the date that one contain-ment cooling subsystem is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding four days provided that ail active components of the other contain-ment cooling subsystem, both core spray subsystems and both emergency power sources for operation of such components if no external source of power were available, shall be operable.
C.
Surveillance of FWCI Subsystem shall be 6.
If the requirements of 3.5.8 cannot be performed as follows:
met, an orderly shutdown shall be initiated and the reactor shall be in a 1.
Item Frequency cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
a.
Pump and valve Per Surveillance C.
FWCI Subsystem operability Requirement 4.13 1.
Except' as specified in 3.5.C.3 below, the FWCl subsystem shall be operable whenever the reactor pressure is greater than 90 psig and irradiated fuel is in the reactor vessel.
b.
Simulated Autamatic Every refueling Actuation Test outage 2.
There shall be a minimum of 225,000 gallons of water in the condensate storage tank 2.
Once a week th'e quantity of water in the for operation of the FWCI.
condensate storage tank shall be logged.
Amendment th). 73,19. pg, 76 3/4 5-5
LIMIT!!iG CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.
From and af ter the date that the TWCl subsy' stem is made or found to be inoperable for any reason, reactor operation is permissible only during the succeeding seven days unless such subsysten is sooner made operable, provided th*t during such seven days all active com-ponents of the Automatic Pressure Relief Subsystem. the core spray sut, systems. LPCI subsystem, and isolation c.ondenser system are operable.
4.
If the requirenents of 3.5.C cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
D. Surveillance of the Autonic Pressure Relief D. Automatic Pressure Relief (APR) Sybsystems
- 1. During each operating cycle the follba-1.
Except as specified in 3.5.0.2 and 3 below, ing shall be performed:
the APR subsysten shall be operable when-ever the re* actor pressure is greater than
- a. A simulated automatic initiation of 90 psig and irradiated fuel is in ;he the system throughout its operating reactor vessel, sequence but excludes actual valve opening, and
- b. With the reactor at low pressure. Each relief valve shall be manually opened until valve operability has been veri-fled by torus water level instrumenta-tion, or by an audible discharoe detected by an individual located outside the torus in the vicinity of each relief line.
Amendment tio. 79, AJ, $J, 76 3/4 5-6
SURVEILLANCE REQUIRENENT LlHITil4G C0H0li!0N f0R OPERATION 2.
When it is detenalned that one safety /
2.
From and af ter the date that one of the relief valve of the automatic pressura three relief / safety valves of. the auto-rallef subsystem is inoperable the matic pressure relief subsystem is made or actuation logic of the remaining AFR found to be Inoperable when the reactor is valves and FWCl subsysten shtl1 be pressurized above 90 psig with Irradiated demonstrated to be operable luecdlatel fuel in the reacter vessel, reactor opera-and daily thereafter.
~
tion is pennissible only durino the succeeding seven days unless repairs are E. Surveillance of the Isolation Cor.Jenser made and provided that during such time System shall be performed as follows:
the remaining automatic pressure reller valves, fWCl subsysten and gas turbine 1.
Isolation Condenser Systna Testis,1:
generatur e,re operable.
- a. The shell side water level and 3.
If the requironents of 3.5.0 cannot be temperature shall be checked met. ar, orderly reactor shutdown shall daily.
be initiated and the reactor shall be In a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
E. Isolation Condenser System 1.
Whenever the reactor pressure is greater than 90 psig and irradiated fuel is in the reactor vessel, the isolation con-denser shall be operable except as specified in 3.5.E.2 and the shell side water level shall be greater than 66 inches.
3/4 5-7 Aaendrent No. 16,;57,76 4
I fililifill t.flNiilil0N f 0ll OPtttAll0N TJ?.Mll t AHf.0 RfQtilR[HfMT 7.
From anel af ter the date t hat t.he isola-le.
Sleeulated automat *: ectuation anal tion Condenser is made or found to be Inoperable. for any reason, power functional systo4 testing shall be operation shall be restricted to a performed during each refueling maximum of 40% of full power, f.e.,
outage or whenever major repairs (804 H ( ) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> untti are cannpleted on the systca.
such thhll the Isolation Condenser c.
The system heat removal capability is returned to service.
shall be determined once every five years, d,
Calibrate vent Ifne radiation 3
If the requirteents ai 1.5.E cannot be monitars quarterly, met, an orderly Stdow shall be infilated amt the reactor shall be in a cold e.
Motor operated valves shall be shutdown conditic9 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
tested per surveillance requirement 4.13.
F.
Miniann rnre and Containment Cooling Syster.
Availability 1.
Tucept as specified in 3.5.F.2. 3.5.r.3, F.
Survalliance of Core and Containment Cooling 3.5.I'.7 and 3.5.f.R below, both emer-Jystu 4
gency power sources shall be operable whenever irradiated fuel is in the 1.
The surveillance requireme?its for normal reactor.
operation are in Section 4 3 l
A:nenJuent No. /#, J$, $), 76
~
LIMITING CONDITION FOR OPERATION StlRVEILLANCE REQUIREMENT 2.
From and after the date that the diesel generator is made or found to be inoper-able for any reason, continued reactor operation is permissible only during the succeeding seven days provided that the gas turbine generator, FUC1, Automatic Pressure-Relief Subsystem, all components of the low pressure core cooling and the contaimnent cooling subsys+ ems shall be operable.
3.
From and after the date that the gas turbine generator is made or found to be inoperable for any reason, continued reactor operation is permissible only during the succeeding four days provided that the diesel generatnr, all components of the APR subsystem, all components of the low pressure core cooling and contain-ment cooling subsystems shall be operable.
4.
If the requirements of 3.5.F.1 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
5.
Any combination of inoperable canponents in the cort and containment cooling systems shali not defeat the capability of the remaining operable cuponents to fulfill the core and containment cooling functions.
Auendment flo. 29, 76 3/4 5-9
4 T
N EM E
k l
itQ E
R E
CNA L
L I
EVR l
iS 0
+
1 5
4
/
3 n
'r w
n ho a
eo ct r
m i i ic d
l t t ha r
h d
e e
s ma l
it we odit e
l l e
v i
ec e
n l
di e r
tl di i
ip h
i e
ti h
k r
nn nel e conw f
wm tl r t
e sr e
t eaoib ne aco f s o
ed st nwyt ht ss m
h ca aoi e
ceii t c nu ynoosc t nyh wl rdt re b
t n
ifd sa l
e aat l
onae hg da el o
l) nl
)l ri swt pgg et nyes nl dnr yol ieoe 1
- 7. s o ri onsi h i l ii oe cf mhl rciott e
,c d
i opw av eu ea aol i
os F. edoi i
i ii t
s e
N i
s vt cl er e
nr a
rt o p
cnl d
ee 0
5 u
bi s l udea l a.
o r
serdeeb e
i l
sf or p
s f rt rov jd a ieern ehd e a
1 i
r rog 1
3hsr b sr er s
etd ren eu
,ae nr i
i A
t a
id s
r m
n ruo eh o rs uci i i i
l ee H
n d
y ie as/l r
l so or pt c ossi pp rss a i
l i rl eak r oroii dh o
i cemet eo dei P
i orinro l t ornt ntl e
an ret i d ree r
n O
d co of ec t ai aio bl e
h pt sane s pt b es csw uanrihd wr eep t
s yli b u
s H
ii e n
l f eoocn l
t l sl o owyso r w yl i
0 f
h eeoa ricco aen l si bos o yl oosl 1
il t rt ni d
t nc vno i
i ehs l
ycbl ce us t
eeie ooc wvwi t l i
)
na a
nw O
eunsyt n hdl eg nd t
horedh eeo 1
pfi ssae at ngvn e
f krgg ti pres hl is 6
l s
ebl t i
i oe ii i
reo ronn rot sui tbt m 4
l dsruL o di ct rw b
otii ebc sl r ace n
seips p
an sdo l
t wcps h
eespe hret lu at d
ran l
et n apu t)j repw t ejs 1
N arwgee r
wael uoe o ei o noruo ipny 3
O ti oondhl i 1 il o
h o f nm Nrsh e1 cpsp Wois i
C pdtl i it e ed
,tf eac l v s
re st n p
G cral ooss es uod e
)
)
)
)
7 N
x rel orae h ehi nh a
b c
d l
Ei r a c ph v Wvst at
(
(
(
l
(
3 i
H 7
I L
6 7
tneman e
nA 4
1 j
L IMITING CONDIlltd fcR OPERAliON 91RVEILL ANCE REQUIREMENT G.
Jet Pungy
- t..
Jet Puenps 1.
Wheneser the re.sctor is in t he startup/tiot 1.
Whenever there is a recirculation fl a Standby or hun aiodes, all jet pumps shall with the reactor ie the startup/ hot be intact and all operating let pumps shall standby or run modes, act pump inte-be operable.
If it is determined that a grity and operability 4 hall be checked jet pump is inoperable, an orderly shutdown daily by verifying that the following shall be initiated and the reactor shall be two conditions do not occur in a cold shutdown condition within 24 simultaneously:
hours.
a.
The recirculation pump flow 2.
Flow indication from each of the twenty jet differs by more than 101 from the pumps shall be veritied prior to initiation established speed-flow of reactor startup froa a cold shutdown characteris tics.
condition.
b.
The indicated total core flow is 3.
the irdi ated cos e iIow is the sum of the inore than 101 greater than the flow indicati6n irrau each of the twenty jet core flow value derived from pumps.
If flow indication failure occurs established power-core flow for two or sucre jet puinps, inanediate currec-relationships.
tive action shall be taken.
If ilow inoica-tion can not be obtained for at least nine-2.
Additionally, when operating with one teen jet pu:ups, an orderly shutdown shall recirculation pump with the equalizer be initiated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and the reactor valves closed, the diffuser to lower shall be in a cold shutdown condition within plenum differential pressure shall be 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
checked daily, and the differential pressure af any jet pump in the idle 11.
Recirculation Puup flow Hismatch loop shall not vary by more than 101 from established patterns.
1.
Wheneve r t,oth recirculation pump'. are in opera t ion, p m.p speeds sna il be ma intained 3.
The baseline data required to evaluate within 10% of each other when power level the conditions in Specifications is greater than 807. and within l*ft or each 4.6.G.1 and 4.6.G.2 will be acquired other when power level is less t hari !i01.
each operating cycle.
11.
Recirculation Pump flow Mismatch l
1.
Recirculation pump speed shall be checked daily for mismatch.
.4aendinent No. 4, 76 3/4 6-10
LIMITING CONDITION FOR OPERATION SHtVEILLANCE REQUIRLMEOT d.
the fuel cask or irradiated fuel is capauility shall be denonstrated at not being noved within the reactor three or more points within the building.
containment prior to fuel movement and may be dononstrated up to 10 days D.
Primary Contai Snent Isolation Valves prior to fuel movement. Secondary containnent capability need net be
~
1.
During reactor power operating conditions, demonstrated more than once per all isolation valves listed in Table 3.7.1 operating cycle unless damage or and all instrument line flow check valves nodifications to the secondary shall be operable except as specified in containment have violated the integrity 3.7,0.2.-
of the pressure retaining boundary of that structure.
D.
Primary Containment Isolation Valves 1.
The primary contaisunent isolation valves surveillance shall be perfonned as follows:
a.
At least once pe operating cycle l
the operable isoe_.. ion valves that are power operated and automatically initiated shall be tested for simulated automatic initiatior, and closure times, b.
At least once per operating cycle ti.e instrument line flow check valves shall be tested for proper operation, c.
At least once per quarter:
1)
All normally open power-operated isolation valves (except for the main steam-line power-operated isolation valves) shall be fully closed and reopened.
Amendment No. //, 76 3/4 7-14
o-TAnaF 3.7.1 PRINAllY COttIAlffffNT 150I All,0N Isolation 741ve (Valve Ntsdier of Power Group identifica tion Hunber)
Operated Valves
~
Opera ting Initiatirej Inboara Outboard Time (Sec) Position S igna_I__,
1 min Steam 1.ine Isolation (MS-1A, 2A,18, 28 IC, 2C, 4
4 3<T1 S 0
GC ID, 20) -
1 Main Steam Line Drain (MS-5) 1 35 C
SC 1
35 C
SC 1
liain Steau Line Drain (MS-6) 1 Recirculation Loop Sample Line (SM-1, 2) 1 1
S C
SC f
1 Isolation Condenser Vent to Main Steam Line(IC-6, 7) 2 5
0 CC 2
20 0
CC 2
Dryall Floor Drain (SS-3, 4)
I 2
20 0
GC 2
Dryuc11 Equiprent Drain (SS-13,14) 1 10 C
SC 2
Drytell Vent (AC-7) 1 15 C
SC 2
Drp cli Vent Relief (AC-9) 1 10 C
SC 2
Drpell and Suppression Chamber Vent from Reactor Eallding (AC-8) 2 Drywell Vent to Standby Gas Treatment System (AC-10) 1 10 C
SC 1
10 C
SC l
2 Suppression Chamber Vent (AC-11) 1 15 C
SC 2
Suppression Chamber Vent Relief (AC-12) 1 10 C
SC 2
Suppression Chamber Supply (AC-6) i 1
10 C
SC 2
Drp cIl Supply (AC-5) 1 10 C
SC 2
Drytell and Suppression Chamber Supply (AC-4) 3 Clean-rp Dcmineralizer System (CU-2) 1 18 0
GC 3
Cleanup Dcmineralizer Systesa (CU-3,
,28) 2 18 0
GC l
3 Shutdevn Cooling System (50-1) 1 48 C
SC 4
43 C
SC 3
Stutdenn Cooling System (SD-2A, 28, 4A, 48) 1 43 C
SC 3
Shutdem Cooling System (50-5) 3 Reactor Ifead Cooling Line (115-4) 1 45 C
SC 4
Isolation Condenser Stea:n Supply (IC-1) 1 24 0
GC 4
Isolation Condenser Steam Supply (IC-2) 1 24 0
CC 4
Isolation Condenser Condensate Return (10-3) 1 19 C
SC 4
Isolation Condenser Condensate Return (IC-4) 1 19 0
CC iccenter Check Valves (fW-9A,10A, 98,108) 2 2
na O
Process Centrol Rod llydraulic Return Check Yalves (301-95, 98) 1
-1 MA 0
reocess Reactor Head Cooling Check Valves (115-5) 1 HA C
Process Stan%y Liquid Control Check Valves (SL-7, 8) 1 1
na C
Process 3
Clearnp Dtuineralizer System (CU-5) 1 la C
SC l
3/4 7-15 taendentno.J3[N,76
l n
)
a e
l h
m ia e
n mi oi nn t
oi at nit e
it rg nil ut aaan it a
l t,
ml pi ihe es o
n dscb nae esye eoa ri v
hibp t pel i
eaei t
o s
l wmnt
,d d e iena eer lhih o
o pp( e ht s t
s i
r t ait e
,eve
'rrsu nrcn v
ot es oere l el v vo mpeu al al tc
,l l
oxq vb va aliac r - ee a
v eev er s
nrr rt y
peei oeee d
phl anf wb i i
eroi ec s
t ot b i r cpl anoa h
T d
l i r
t f L e n
l N
oav oean o
ee E
nha ssnpy h
l M
%od ti s i i ool f
t 5 s n sl e
ni i
4 W7i a a-sr n1 ti a l
l emeu a
s d
ll l avs 7
an 0
el o r
ead E
tt al e3l e
)
R
?
Asvc v
gd eet nr E
nl ai o C
eb vc N
h af ae A
d WT oh r t
t i
f V
2 R
U S
7 1
7 1,
/
3 d r d
eo n
i t t o
f ca p
i a ns n
ced ae
,eo
' ere r
t vi p
d gr eat no mhi s,i evic d
el ov ed n vb rae bnoac a ph d lar on t
veeemo onn nww pun i
N noniet noo O
onil hi add iit td ct t I
t nh n
uu T
asocno
- 0. h h A
ecaic ss l
R om e
7 f
yd soy sd p
canie 3ll i
O emi t
ro R
yb ea nec.
O n
nevl od s
rer F
a1 ovl o i
. i l as t oh u N
t 7. t avi a
t o cn h
n av O
an e3 r ee i
I T
v eel h f
i4 i e 2
eepnbt I
ooa ct r l
D N
en ro eaon 6
O h art et piti 7
C tI esp St ch waog iat G
nnoenn f nei o
rw N
pl i i I i I i N
I t
T ne I
M m
2 3
d I
L nem A
w
LIMITItiG C0flDITION FOR OPERATION
' UIWLl'LLANCE REQti!RIHFN1 e
S.
All station and switchyard 24 and 125 volt c.
During theanonthly generato te.t,
~
the diesel fuel oil transfer pump.
batteries and associated battery chargers shall be operated.
are operable.
B.
tihen the mode, switch is in Run, the availability 2.
Gas Turbine Generator or power shall be as specified in 3.9.A. except a.
The gas turbine generator shall be as specified below:
fast started and the output brealers 1.
Froin and a f ter the date that incoming power closed within 48 seconds once a month is available frodi only one 345 kv line, to demonstrate operational readiness.
The test shall continue until the reactor operation is permissible only during the secceeding seven days unless an gas turbine and generator are at additional 345 f:v line is sooner placed in equilibrium temperature at full load service.
output. Lise of this unit to supply power to the system electrical net-2.
f rom the a f ter the date that incaining power work shall constitute an acceptable is not available froin any 345 Lv line, demonstration of operability, reactor operation shall be penaitted pro-b.
During each refueling outage, the vided both emergency power sources are conditions under which the gar turbine-operating and the isolation condenser system l
is operable.
fhe NRC shall be notified, generator is required will be simulated i
within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the precautions to be and a test conducted to verify that taken during this situation and the plans it will start-and be able to accept for restoration af inconing power The emergency loads within 48 seconds.
minimuci fuel supply for the gas turbine during this situation shall be maintained B.
Hatteries O
above 20,000 gallons.
1.
Station Batteries 3.
From and af ter the date that either emer-a.
Every week the specific gravity and gency power source or its associated bus is N
made or found tn be inoperable for any voltage of the pilot cell and tem-Q"J reason, reactor operation is permissil.le perature of adjacent cells and overall according to Specific'ation 3.S.f/4.Sf unless battery voltage shall be measured, such emergency power source and its hus M
are sooner aade operable, provided that b.
Eve,y three months the measurements during such time two otisite lines (345 -
shall be made of voltage of each cell to nearest 0.01 volt, speci fic or 27.6 kv) arc operable.
gravity of each cell and temper. ture of every fifth cell.
Amendment flo. //, jf, 76 3/4 9-2
LIMITING CGNDITION FOR OPELATION SilRVEILLANCE REQUIREMENT 4
from and af ter the date that one of the c.
At every refueling cutage or at two 125 volt or 24 volt battery systens is 18 months intervals, the station a
made or found to be inoperable for any batteries snall be subjected to a l
reason reactor operation is pennissible performance test in accordance with only during the succeeding seven days the procedures described in Section unless such battery systen is sooner made 5.4 in IEEE Standard 450-1972, operablq.
"IEEE Recommended Practice for Ma'intenance, Testing, and Replace:-
C.
Diesel and Gas Turbine fuel ment of Large Stationary Type Power Plant and Substation Lead Storage There shall be a minimum of 20,000 gallons of Batteries".
diesel fuel supply onsite for the diesel and a minimum of 35,000 gallons onsite.'or the gas 2.
Switchyard Batteries turbine, except as permitted in Specification l
3,9.B.2 a.
Every week the specific gravity and voltage of the pilot cell and tenperature of 4djacent cells and overall battery voltage shall be measured.
b.
Every three months the measurements shall be made of voltage of each cell to nearest 0.01 volt, specific gravity of each cell, and tenperature of every fifth cell.
C.
The quantity of gas turbine generator and diesel generator fuel shall be logged weekly and af ter each operation of the unit.
Once a month a sample of the diesel and gas turbine fuel shall be taken from the undar-ground storage tanks and checked for quality.
1 Amendment No. 76 3/4 9-3
_iHITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.11 REACTOR FUEL, ASSEMBLY 4.11 REACTOR FUEL ASSEMBLY Applicability Aplicability The Limiting Conditions for Operation associated with The Surveillance Requirements apply to the para-the fuel rods apply to those parameters which monitor meters which monitor the fuel rod operating r4e fuel rod operating conditions.
conditions.
Objective Objective The Objective of the Limiting Conditions for Opera-The Objectivt of Surveillance Requirements is tion is to assure the performance of the fuel rods.
to specif" the type and frequency of surveillance to be applied to the fuel rods.
Specifications Specifications A.
Average Planar Linear lleat Generation Rate (APLilGR)
A.
Average Planar Linear lleat Genention Rate (APLilGR) 1.
During power operation, the APLilGR for each type of fuel as a function of average The APLHGR for each type of fuel as a planar exposure shall not exceed the limit-function of average planar exposure shall ing value shown in Figure 3.11.1.
be detennined daily during reactor operation at > 25% rated thermal power.
2.
If at any time during operation it is detennined by nonnal surveillance that the limiting value for APLilGR specified in Section 3.11.A.1 is teing exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the APLilGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
4 3/4 11-1 Amendment No. A, AA, fe, 34, 49, 67, 73, 76-
LIMIT it;r, C0;O! TI9 3 f 0it OpfrATlord
$URUfill AfiCE RCfPflREMEtiTS (h) Verifying the diesel starts from ambient conditions on the auto-start signal and operates for >20 minutes while loaded with the fire pump.
h.
The fire pump diesel starting 12-volt batteries and charger shall be demonstrated OPERABLE:
1 At least once per 7 days by veri-fying that:
(a) The electrolyte level of each battery cell is above the l
plates, and (b) The individua.1 overall battery voltages are > 12 volts.
2.
At least once per 92 days by vert-fying that the specific gravity is appropriate for continued service of the batteries.
l 3.
At least once per 18 months by verifying that:
+.
(a ) The batteries, cell plates and battery racks show no visual indication of physical danage or abnormal deteriora-y tion, and (b) The terminal connections are l
clean, tight, free of corrosion and coated with anti-corrosion Amendment No. f')', 76 material.
i 3/4 12-4
LIMITING C0flDITION FOR OPERATION SURVEILLANCE REQUIREMEllT 3.13 INSERVICE INSPECTION
-f 4.13 INSERVICE INSPECTION-
,Applicabi1ity Applicability Applies to ASME Boiler and Pressure Vessel Code Applies to the periodic inservice inspection Section 15.1 Class I, 2, and 3 equivale it components.
'and. testing of ASME Boiler and Pressure Vessel Code Section 111 Class 1, 2, and 3 equivalent Obiective components.
To assure the structural integrity of the applica-Oh.iective ble components defined above.
To verify the structural integrity of the appli-Speci fica tion cable components defined above.
The structural integrity of ASME Code Class 1, 2, and Speci fication 3 equivalent components shall be maintained at an acceptable level in aCCordance With 10CFR50.55a(g).
A.
Inspections Inservice inspection of. ASME Boiler and Pres-sure Vessel Code Section III Class 1 Class 2, and Class 3 equivalent components shall be performed in accordance'with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10CfR5G.
Section 50.55a(g) with' the exemptions and al-ternate inspections that have been approved by the NRC pursuant to 10CfR50, Section 50.55 a(9)(6)(i). These exemptions and alternate inspections are included in.the Inservice Inspection Program.
B.
Testing 1.
Surveillance with operable components Inservice testing of ASME !! oiler and Pressure Vessel Code Section 111 Class 1, Class 2, anil Amendment No. (//, 76 3/4 13-1
+
a
ApRH's #4, #5 and #6 are arranged similarly in the other protection trip system. Each protection trip system has one mere APRH than is necessary to meet the minimum. number required per channel.
This allows the bypassing of one APRH per protection trip system for maintenance, testing or calibration. Additional IRM chaenels have also been provided to allow for bypassing of one such channel.
The bases for the scram settings for the IRM, APRM, high reactor pressure, reactor low water, generator load rejection, and turbine stop volve closure are discussed in Section 2 of these specifications.
Instrumentation (pressure switches) in the drywell is provided to detect a loss of coolant accident and initiate l
the emergency core cooling equipment. This instrumentation is a backup to the water level instrumenatation which is discussed in Specification 3.2.
A scram is provided at the same setting as the emergency core cooling system (ECCS) initiation to minimize the energy which must be acconunodated during a loss of coolan't accident and to prevent the reactor from going crit.ical foiiowing the accident.
The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping. A part of this piping is an instrument volume which accanunodates in excess of 39 gallons of water and is the low point in the piping. No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram.
During normal operation the discharge volume is empty; however, should it fill with water, the water discharged to the piping from th? reactor could not be acconnodated which would result in slow scram times or partial or no control rod i nsert. isn. To preclude this occurrence, level switches have been provided in the instrumented volume which alarm and scran; the reactor when the volume of water reaches 39 gallons. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the scram times or amount of insertion of the control rods.
This function shuts the reactor down while sufficient volu.ne remains to acconnodate the discharged water and precludes the situation in which a scram would be required but not be able to perfonn its function adequately.
Loss of condenser vacuum occurs when the condenser can no longer handle the heat input. Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves which eliminates the heat input to the condenser. Closure of the turbine stop and bypass valves causes a pressure transient, neutron flux rise, and an increase in surface heat flux. To prevent the clad safety limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure. The turbine top valve closure scram function alone is adequate to prevent the clad safety limit from being exceeded in the event of a turbine trip transient with bypass closure.
Ref. Section 4.4.3 FSAR. The condenser low vacuum scram is a back-up to the stop valve closure scram and causes a scram before the stop valves are closed and thus the resulting transient is less severe. Scram occurs at 23" Hg vacuum, stop valve closure occurs at 20" Hg vacuum and bypass closure at 7" Hg vacuum.
Amendment No. 76 8 3/4 1-2 1
l
.e the one performed just prior to shutdown or startup; i.e., the tests that are performed. just prior to use of the instrument. While included in Group (C), the Condenser low Vacuum trip is treated. differently.,
This is because t.he condenser low vacuum tyip sensor can only be tested during shutdown.
The primary function of l
this trip is to protect the turbine and condenser, although it is connected into the reactor protection system; thus testing the sensor at each refueling outage is adequate.
Calibration frequency of the instrument channels is divided into two groups.
These are as follows:
a.
Passive type indicating devices that can be compared with like units on a continuous basis.
b.
Vacuum tube or semiconductor devices and detectors that drif t or lose sensitivity.
Experience with passive type instruments in generating stations and substations indicates that the specified calibrations are adequate.
For those devices which employ amplifiecs, etc., drif t specifications call for drift to be less than 0.4%/ month; i.e., in the period of a month a drift of 0.4% would occur ar.d thus providing l
for adequate margin. For the APRM system drif t of electronic apparatus is not the only consideration in detennining a calibration frequency. Change in power distribution and loss of chamber sensitivity dictate a calibration every seven days. Calibration on this frequency assures plant operation.at or below thennal limits.
B.
The peak heat flux shall be checked once per day to determine if the APRM scram requires adjustment.
This will normally be done by checking the LPRM readings. Only a small number of control rods are moved daily and thus the peaking factors are not expected to change significantly and thus a daily check of the peak heat flux is adequate.
4 9
Amendment No. 76 8 3/4 1-8
The high drywell pressure instrumentation is a back-up to the water level instrumentation and in addition to.
initiating ECCS it causes isolation of Group 2 isolation valves. For the breaks discussed above, this instrumenta-tion will initiate ECCS operation at about the same time as the low low water level instrumentaticn; thus the results given above are applicable here also. Group 2 isolation valves include the drywell vent, purge and sump isolation valves, and reactor building ventilation isolation valves. Group 2 actuation also initiates the SBGTS. liigh drywell pressure activates only these valves because high drywell pressure could occur as the result of non-safety related causes such as not purging the drywell air,during startup. Total system isolatio.. is not desirable for these condi-tions and only the valves in Group 2 are required to close.
The low low water level instrumentation initiates protection for the full spectrum of loss of coolant accidents and causes a trip of all primary system isolation volves.
Venturis are provided in the main steamlines as a means of measuring steam flow and also limiting the loss of mass iaventory from the vessel during a steamline break accident.
In addition to monitoring the steam flow, instrumenta-tion is provided which causes a trip of Group 1 isolation valves. The primary function of the instrementation is to detect a break in the main steamline, thus Group i valves are closed.
For the worst case accident, mai., steamline break outside the'drywell, this trip setting of 120% of rated steam flow in conjunction with the flow limiters and main steamline closure, limit the mass inventory loss such that fuel is not uncovered, fuel tenperatures rem 0in less than 150",'F and release of radioactivity to the environs is well below 10 CFR 100 guideline values. The main steam-line high flow break detection is a one out of two twice logic for each individual steamline, four detectors per line for a total of 16 detectors. When a steamline is isolated by closing both main steam isolation valves the operable instrument channels per trip systen requirenents are not required to be met because ti e protection afforded by the remaining operable logic in the in-service steamlines provides couplete recognition of toe steam flow measurunents required for correct protective action.
Temperature monitoring instrumentation is provided in the main steamline tunnel to detect leaks 'in this area.
Trips are provided on this instrumentation and when exceeded cause closure of Group 1 isolation valves.
Its setting of 200*F is low enough to detect leaks of the order of 5 to 10 gpm; thus, it is capable of covering the entire spectrum of breaks. For large breaks, it is back-up to high steam flow instrumentation discussed above, and for small breaks with the resultant small release of radioactivity, gives isolation hefore the guidelines of 10 CFP.100 are exceeded, liigh radiation monitors in the main steamline tunnel have been provided to detect gross fuel failure. This instru-mentation causes closure of Group 1 valves, the only valves required to close to prevent further release to the environment. With the established setting of seven times normal background, and main steamline isolation valve closure, fission product release is limited so that 10 CFR 100 guideline values are not exceeded for the most rapid failure mechanism postulated (control rod drop accident).
Pressure instrumentation is provided which trips when main steamline pressure at the turbine drops below 880 psig. A trip of this instrumentation results in closure of Group 1 isolation valves.
In the " Refuel," " Shutdown," and "Startup/ilot Standby" mode this trip function is bypassed. This function is provided primarily to provide protection against a pressure regulator malfunction which would cause the control and/or bypass valves to open. With the trip set at 880 psig inventory loss is limited so that fuel is not uncovered and peak clad tenperatures are much less than 1500"F; thus, there is no release of fission products other than those in the reactor water.
Amendnent No. /, 76 B 3/4 2-2
3.3 Bases
A.
Reactivity Limitations The core reactivity-limitation is a restriction to be applied principally to_ the design of new fuel which nuy be loaded in the core nr into a particular refueling pattern.
Satisfaction of the: limitation can only be demonstrated at the time 'of loading and must be such that it will apply to the entire subsequent fuel cycle. The reactivity of the loaded core will be limited so the core can be made g.ubcritical by dt least R + 0.33% AKt in the most reactive condition during the ' operating cycle, with the strongest control rod fully withdrawn and all others fully inserted.
The value of R in %AK is the amount by which the core reactivity, at any time in the operating cycle, is calculated to be greater than at the time of the check, i.e.,
the initial loading. R must be a positive quantity or zero.
The value of.08% AK has been added to the normal value of 0.251 AK to allow for potential maximum settling of the B4C powder in the inverted control rod tul'es still remaining in the core. A core which contains temporary control or other burnable neutron absorbers may have a reactivity characteristic which increases with core lifetime, goes' through a maximum and then decreases thereafter. See Figure 3.3.2 of the FSAR for such a curve.
The vilue of R is the difference between the calculated core reactivity at the beginning of the operating cycle and the calculated value of core reactivity any time later in the cycle where it would be greater l
than at the beginning. for the first fuel cycle, R was calculated to be not greater than 0.10% AK.
A neu value of R must be determined for each fuel cycle.
The 0.33% AK in the expression R + 0.33% AK is provided as a finite, demonstrable, subCriticality margin.
This margin is demonstrated by full withdrawal of the strongest rod and partial withdrawal of an adjacent rod to a position calculated to insert at least R
- 0.33% AK in reactivity. Observation of subcriticality in this conditior assures subcriticality with not only the strongest rod fully withdrawn but at least a R + 0.33% AK margin beyond.this.
2.
Peactivity Margin - Stuck Control Rods Specification 3.3.A.2 requires that a rod be taken out of service if it cannot be moved with drive pressure.
If the rod is full-y inserted and then disarmed electrically,* it is in a safe position of maximum contribu-tion to shutdown reactivity.
If it is disanned elctrically in a non-fully inserted position, that position shall be consistent with the shutdown reactivity limitation stated in Specification 3.3.A.I.
This assures that the core can be shutdown at all times with the remaining control rods assuming the strongest. operable control rod does not insert. An allowable pattern for control rods out of service, which shall meet this specification, will be available to the operator.
- To disarm the drive electrically, four amphenol type plug connectors are removed from the drive insert and withdrawal solenoids rendering the drive immoveable. This procedure is equivalent to valving out the drive and is. preferred because, in this condition, drive water cools and minimizes crud accumulation in the drive.
'See October 22, 1974 Inverted Control Rod Inspection and Analysis Report, and Technical Specifications Change Request.
4 Amendmeet.n./,76 8 3/4 3-1 o
3.4 Bases
A.
The design ob.fective of the liquid control system is to provide the capability of bringing the reactor from full power to a epid, menon-free shutdown asstuning that none of the withdrawn control rods can be inserted.
Io meet this ob.fective, the liquid control system is designed to inject a quantity of boron which produces a concentration of.660 ppu of boron in the reactor core in ins than 125 minutes. The 660 ppu concentration in the reactor core'would bring the reactor from full power to a minimum 2.6% delta K subcritical condition considering the hot and cold reactivity swing, xenon poisonirn, analytical biases and uncertainties, etc.
An additional 25% of beron solution is provided for possible imperfect mixing of the chemical solution in the reactor coolant. A minimum quantity of 2720 net callons of solution having a 13.4% sodium pentaborate concentra-tion is required to meet this shutdown requirement. Actual system volume for this quantity is 2960 gallons.
(240 gallons are contained Delow the pump suction and, therefore, cannot be inserted.)
The time requirement (125 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of the reactor following the xenon poison peak.
for the minimum required pumping rate of 32 gallons per minute, the maximum storage volume of the baron solution is established as 4190 gallons.
Doron concentration, solution temperature (within the tank and connecting piping including check of tank heater and pipe heat tracing system) and volume are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Experience with pump operability indicates that monthly testing is adequate to detect if failures have occurred.
Components of the system are checked periodically as described above anu make a functional test of the entire system on a frequency of less than once during each operating cycle unnecessary. A test of one installed explosive charge is made at least once during each operating cycle to assure that the charges are satisfactory.
The replacement charge will be selected from the same batch as the tested charge. A continual check of the firing circuit continuity is provided by pilot lights in the control room.
The relief valves in the standby liquid control system protect the system piping and positive displacement pianos which are nominally designed for 1500 psi from overpressure. The pressure relief valves discharge back to the standby liquid control solution tank.
B.
Only one of the two standby liquid control pumping circuits is needed for proper operation of the system.
If one pumping circuit is found to be inoperable, there is no innediate threat to shutdown caoabilitv. and reactor operation may continue while repairs are bein,g made.
B 3/4 4-1 Anendment No. Jd, 76
3.5 Bases A.
Core Spray and LPCI This specification assures that adequate emergency cooling capability 'is available.
Based on the loss of coolant analysis included in Section VI FSAR, either of the two core spray subsystems provides sufficient cooling to the core to dissipate the energy associated with the loss of coolant accident and to limit fuel clad temperature (around 2000*F) to well below the clad melting temperature to assure that core geometry remains intact and to limit any clad metal-water reaction to less than 1%. Core spray distribu-tion has been shown, in full scale tests of systems similar in design to that of Millstone Unit 1, to exceed the minimum requirements by at least 25%.
In addition, cooling effectiveness has been demonstrated at less than half the rated flow in sinulated fuel assenblies with heater rods to duplicate the decay heat characteristics of irradiated fuel. The accident analysis is additionally conservative in that no credit is taken for spray coolant entering the reactor before the pressure has fallen to 90 psig.
The LPCI subsystem is designed to provide emergency cooling to the core by flooding in the event of a loss of coolant accident. This system is completely independent of the cora spray subsystem; however, it does function in combination with the core spray system to prevent excessive fuel clad temperature. The LPCI subsystem in combination with the core spray subsystem provides adequate cooling for break areas of approximately 0.2 square feet up to and including 5.8 square feet, the latter being the double-ended recirculation line break without assistance from the high pressure emergency core cooling subsystems.
The-allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate. The method and concept are descr.ibed in Reference (1). Using the results developed in this reference, the repair period is found to be less than 1/2 the test interval. This assumes that the core spray and LPCI subsystems constitute a 1 out of 3 systen, however, the combined effect of the two systems to limit excessive clad temperatures must also be considered. The test interval specified in Specification 4.5 was 3 months. Therefore, an allowable repair period which maintains the basic risk considering single failures should be less than 45 days and this specification is within this period.
For multiple failures, a shorter interval is specified.
Although it is recognized that the infonnation given in reference (1) provides a quantitative method to estinate allowable repair times, the lack of operating dcta to support the analytical approach prevents complete acceptance of this method at this time. Therefore, the times stated in the specific items were established with due regard to judgnent.
i l
(1) APED 5736, Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards, April 1969, i
- 1. M. Jacobs and P. W. liarriott.
I l
Amendment No. 76 l
Should one core spray subsystem become inoperable, the renaining core spray and the entire LPCI-system are available should the need for core cooling arise.
Should the loss of one LPCI pump occur, a nearly full complement of core and containment cooling equipment is available. Three LPCI pumps in conjunction with one core spray subsystem will perform the core cooling function. Because of the availability of the majority of the core cooling equipment, which will be demonstrated to be operable, a 30-day repair period is justified.
If more than one LPCI pump is inoperable, the repair time is set considering the containment cooling function of the LPCI pumps.
B.
Containment Cooling Subsystems The two containment subsystems are provided to remove heat energy from the containment in the event of a loss-valves,oneheatexchanger(40x10gntainmentcoolingsubsystemincludestwoservicewaterpumps., associated of-coolant accident. Each single c BTU /hr), two LPCI pumps and necessary instrumentation, control and power equipment. With two heat exchangers (i.e., both loops) operable, it is possible to degrade system performance to one LPCI and one service water pump operating per loop and still not exceed significantly the equipment design temperatures and not rely canpletely on containment pressure for net positive suction head (NPSil). An interlock to prevent containment spray actuation is included in the design of engineered safetv features to prevent inadvertent pressure reduction below that required for NPSil. Iteference Anendment Nos. 9, 16, 18, 22 and 23.
The heat removal capacity of a single cooling loop is adequate to prevent the torus water temperature from exceeding the equipment temperature capability which is specified to be 203*F in Amendment No. 23.
It also provides sufficient subcooling so that adequate NPSil could be assured without reliance on contaimnent pressure except for short intervals during the postulated accident.
In the event that only one heat removal loup is operable, station operation will be permitted for four days unless necessary repairs are made to make the other loop operable. A four-day period is selected to permit reasonable time for operator action and maintenance operations.
C.
Feedwater Coolant injection The feedwater coolant injection subsystem is provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI or core spray subsystems can protect the core.
The FWCl meets this requirement Amendment No. 76 B 3/4 5-2
. -. - - - -. - ~
without the use of of f-site electrical power. For the pipe breaks for which th2 FWCl is intend:d to function, The repair times for the the core never uncovers and is continuously cooled and thus no clad damage occurs.
1imiting conditions of operatton were set considerIng the use of the fWCl as part of the emergency core coolthy system and isolation cooling systeig.
lhe IWCl utilizes portions of the nonnally operating feedwater system; e.g., condensate, condensate booster Therefore, the reliability of the pumps, valves and motors is constantly being demonstrat.:d.
and feedwater pumps.
Since asi Thus the system has an inherently higher degree of reliability than nonnally non-operating systems.
operating string of pump and valves is progranned for FWCI operation, it is not expected that the nonnally operating portions of the FWCl would be out of operation during nonnal operation.
D.
Automatic Pressure Relief (APR)
They enable The relief valves of the automatic pressure relief subsystem are a back-up to the FTJCI subsystem.
the core spray or LPCI to provide protection against the small pipe break in the event of TWCl failure, by The core spray and/or depressurizing the reactor vessel rapidly enough to actuate the core sprays or LPCI.
LpCI provide sufficient flow of coolant to limit fuel clad temperatures te well below clad melt and to assura that core geometry remains. intact. _
It has been demonstrated that APit testing at low reactor pressure is required during each operating cycle.
the blowdown of the APR to the torus cau es a wave actlpn that is detectable on the torus water level instru-The discharge of a relief line is audible to an individual located outside the torus in the mentation.
vicinity of the line, as experienced at other BWR's.
E.
Isolation Condenser System The isolation condenser is provided for core decay heat The turbine main condenser.is normally available.The isolation condenser has a heat removal capacity sufficient removal following reactor isolation and scram.
Water will be lost fron the reactor to handle the decay heat production at 300 seconds following a scram.
vessel through the relief valves in the 300 seconds following isolation and scram.
This represents a. minor loss relative to the vessel inventory.
0 leend.nent rio. Jf, 7'i, 76
The system may be manually initiated at any time. The system is automatically initiated on high reactor pressure in excess of 1085 psig sustained for 15 seconds. The time delay is provided to prevent unnecessary actuation of the system during turbine trips. Automatic initiation is provided to minimize the coolant loss can be l folloutng isolation from the main condenser. Hake-up water to the shell side of the isolation condenser provided by the condensate transfer punps from the condensate storage tank. The condensate transfer pumps are operable from on-site power. The fire protection system is also available as make-up water. An alternate method of cooling the core upon isolation from the main condenser is by using the relief valves and FWCl subsystem in a feed and bleed manner. The minimum shell side water volume in the isolation condenser is 15.500 gallons.
The function of the isolation Condenser during a small break accident, is to assist the automatic pressure relief system-in depressurtzing the reactor as a backup to.the FWCI system. The two effects of isolation condenser depressurization are:
(1) the minimization of reactor inventory loss which normally occurs during APR blowdown; this reduces the time of core uncovery prior to reflooding; and (2) eariter onset of'10w pressure core spray cooling.
Analysis performed by General Electric.in March 1976. in support of extended operation of Millstone while the isolation condenser was being retubed indicated. that from 401 rated power, over 30 minutes is available to initiate operator action to mitigate the consequences of a loss of all feedwater. This is based upon manual depressurization with APR and coolant supplied by the LPCI and core spray systems. The FWCl m s assuned lost as part of the non-mechanistic assumption of loss of feedwater. The successful mitigation of this postulated event was no uncovering of the fuel. Operators are instructed regarding special procedures to be utt11 red during this mode of plant operation. Thus, reducing pcuer to 40% when the isolation condenser is inoperable provides a limtting condition for operation th3t ir suffieient to preclude the need for any additional limiting conditions for operation on other ECCS systws.
F.
Emergency Cooling Availabt11ty The purpose of Specification F is to assure a minimum of core cooling equipment is available at all times.
If, for example, one core spray were out of service and the emergency power source which powered the opposite core spray were out. of service, only two LPCI pumps would be available. Likewise, if two LpCI pumps were out of service and two emergency service water pumps on the opposite side were also out of service, no contaircent cooling would be available.
It is during refueling outages that major maintenance is performed and during such time that. low pressure core cooling systems may be out of service depending on the activities being perforraed.
Specifictfication F allows renoval of one CRD mechanism or fuel rmoval and replacement while the torus is in s drained condition witt.out compromising core cooling capability. The specification establishes the mininum operable low pressure core cooling system, water inventories, electrical power supplies and other additional requirements that must exist to allow such activities as CRD mechanism maintenance or fuel renoval and replace-nient, to be perfmued in parallel with other major activities. The available core cuoling capability for a potential draining of the reactor vessel while this work is perfonned is based on an estinated drain rate ard the maintained minimum volune of water. 383.000 gallons, in the refueling cavity to be supplied to the reactor Amendment No. 2J K. N. A(, A/, 76
4.5 Bases
l j
The testing interval for the core and containnent cooling systems is based on a quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be fully testable during operation.
For example, the core spray final admission valves do not open until reactor pressure has fallen to less than 350 psig, thus, during operation ~ even if high drywell pressure were simulated, the final valves would not open.
l The systems can be automatically actuated during a refueling outage and this will be done. To increase the availability of the individual components of the core and containment cooling systems, the components which make up the system; i.e., instrumeniation, pumps, valve operators, etc., are tested more frequently.
The instrumentation is functionally tested each month.
I 8 3/4 5-6 lbnendment No. 76
fcr a crack siza uhich giv;s a lcakaga rato ef 2.5 9p% the probability cf r:pid propagatta is Icss th.a 10. A leakage rate of 2.5 gpa is detectable and naasurable.
lhe 25 9pra limit on total leakage to the containneH was established by considering the removal capabilittes cf tha punps.
The capacity of eithes' of the drywell floor drain sump pumps is 50 gpn and the capacity of either of the drywell equipnent drain sump pumps is also 50 gpm.
Removal of 26 gpu from either of these sumps can be accomplished with considerable margin.
The perfornunce of the reactor coolant leak detection system will be evaluated during the 'first year of conmercial operation and the conclusions of this evaluation will be reported to the AEC.
The main steam line tunnel leakage detection system is capable of detecting small leaks.
The system per-formance will be evaluated during the first five years of plant operation and the conclusions of the evalu-ation will be reported to the AEC.
E.
Safety and Relief Valves Present experience with the new safety / relief valves indicates that testing of at least 50% of the safety valves per refueling outage is adequate to detect failures or deterioration.
The tolerance value is speci-fied in Section III of the ASME Boiler and Pressure Vessel Code as +1% of design pressure.
An analysis has been performed which shows that with all safety valves set 1% higher the reactor coolant pressure safety limit of 1375 psig is not exceeded.
The relief / safety valves have two functions:
1.e., power relief or sel f-actuated by high pressure. The solenoid actuated function (automatic pressure relief) in whi<.h external instrumentation signals of cointi-dent high drywell pressure and low-low water level initiate the valves to open.
This function is discussed in Specification 3.5.0.
In addition, the valves can be operated manually.
The safety function is performed by the same relief / safety valve with a pilot valve causing main valve operation.
It is understood that portions of the Acoustic Valve Position Indication cannot be repaired or replaced during operation, therefore, the plant must be shutdown to accomplish such repairs.
The 30-day period to do this allows the operator the flexibility to choose his time for shutdown; meanwhile, because of the redundancy provided by the valve discharge temperature monitor and the continued nonitoring of the renuining valves by both methods, the ability to detect the opening of a safety / relief valve would not be compromised.
The valve operability is not affected by failure of the Acoustic Valve Position Indi-cation System.
v G
- l Because of, the.large volume and thermal capacity of the suppression pool, the volume and temperature nornully change very slowly and monitoring these parameters once per shift is sufficient to establish any tunpera ture trends.- By requiring the suppression pool temperature to be continually monitored ~
and frequently logged during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken.
The ' requirement for an ex ternal visual-examination following any event where potentially high loadings could occer provides assurance tlut no significant danuge was encountered.
Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.
- 3. and 4.
Vacutan Relief The purpose of the vacuum relief valves is to equalize the pressure between the drywell and suppre,-
sion chamberjand suppression chamber and reactor building so that the structural integrity of the I
contaimient is maintained.
The vacuum relief system between the pressure suppression. chamber and reactor building consists of two 100% vacuum relief breakers (2 parallel sets of 2 valves in series).
Opera tion of either syst.m will nuintain the pressure defferential'less than 1 psig; the external design pressure. One valve may be out of service for repairs for a period of seven days.
If repairs cannot be completed within seven days, the reactor coolant system is brought to a condition where vacuum relief is no longer required.
9 e
Anen3 ment no. g g, 76 n ara r_,
The design basis loss of coolant accident was evaluated at the primary contaltunent caximum allowable acGlent leak rate of 1.5%/ day at 43 psig. The anslysis showed that with this leak rate and a standby qas treatment syste.n filter ef ficiency of 901 for halogens, 951 for part iculates, and assumints the fission product release fractions s ated in I10-14844, *.he maximum total whole body passing cloud dose ts about 5 run end the manimum total thyroid dose is about 125 rem at the site bouridary over an esposure duration of 'two hours. The resul' tant doses that would occur for the duration of the accident at the low population distance of 2.3 miles are 4 rem whole body and 16h rem maximum rotal thyrold dose. Ihns, these doses reported are the maximum that would he empected in the unlikely event of a design basis loss of coolant accident.
These do".es are also h.r.ed on the assumption of no holdup in the secondary contairwent resulting in a direct release of fission products fros.: the primary containment through the fil ters and stack to trie environs.
T here fore, the specified primary containment leak rate and filter ef ficiency are conservative and provide margin between expected of f-site doses and 10 CTR 100 guidelines. The fission product source term defined in TID-14844 was also used in the design of factitty engineered safety features including shielding and filter sizing.
The maximum Clowable test leak rate is 1.21/ day at a pressure of 43 psig. This value for the test condition was derived..ua tne maximum allowable accident leak rate of about 1.5%/ day when corrected for the ef fects of containment environment under accident and test conditions.
In the accident case, the containment atmosphoe initially would be coinposed of steam and h'ot air depleted of oxygen whereas under test conditiosis the test atmosphere would be air or nitrogen at ambient conditions. Considering the di f ferences in mixture composition and tempgtures, the appropriate correction factor applied was 0.8 and determined from the guide on contaltunent testing Although the does calculations suggest'.that the accident leak rate could be allowed to increase to about 3.01/ day before the guidelines thyroid dose value given in 10 CFR 100 would be exceeded, establishing the test limit of 1.21/ day provides an adequate margin of safety to assure the health and sa fety of the general public.
It is further considered that the allowable leak rate should not deviate significantly from the containment design value to take advantage of the design leak-tightness capability of the structure over its service lifetime. Additional margin tio maintain tre contaltmient in the "as built" condition is cchieved by establishing the allowable opera-tional leak rate. The operational limit is derived by multiplying the allowoble test leak rate by 0.15 thereby providing a 25% margin to allow for leakage deterioration which snay occur during the period between leak rate testi The primary containment leak rate test frequency is basert on maintai 29 adequate assurance that the leak rate remains within the specification. The leak rate test fr l
testing and surveillance of reactor containment vessels.gency is based on the NRC quide for dtvelopir.g leak rata Allowing the test intervals to be extended up to 8 months permits sone flexibility needed to have the tests coincide with scheduled 0: unscheduled shutdown periods.
{T)~~ TID T0581, Leakage Characteristics of Steel Containment Vessel and the f
.ysis of Leakage Rate Determinations.
~
(2) Technical Safety Guide Reactor Contairmient Leakage Testing and Surveillance Requirements. USAIC, Ulvision of Safety Standards, Revised Dra f t. December 15, 1966.
A.aendment No. 76 B 3/4 7-8
s B.
Standby Gas Treatment Systeiend C.
Secondary Containment Pressure drop across the combined llEPA filters and charcoal adsorbers of less than 7 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. I! eater capability, pressure drop and air distribution should be determined at least once per operating cycle to show system perfornance capability.
The frequancy of tests and sample snalysis are necessary to show that the !! EPA filters and charcoal adsorbers can perform as evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be perfonned in accordance with USAEC Report DP-1082.
Iodine removal efficiency tests shall follow RDT Standard M-16-1T.
The charcoal adsorber efficiency test procedures should allow for the removal of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent thoroughly and obtaining at laast two samples. Each sample should be at least two inches in diameter and a length equal to the thickness of the bed, If tett results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent quallfled according to Table 1 of Regulatory Guide 1.52.
The replacement tray for the adsorber tray removed for the test should meet the same adsorbent quality.
Tests of the llEPA filters with DOP aerosol shall be performed in accordance to ANSI N101.1-1972. Any llEPA filters found defective shall Le replaced with filters qualified pursuant to Requlatory Position C.3.d of Regulatory Guide 1.52.
Although the SCTS design flow rate is 1100 SCFM, the DOP test at reduced flow rate is actually more sensitive because diffusion is the primary mechanism of small particle collection. The lower limit for test flow rate (500 SCFH) is based on test instrument sensitivity.
All elements of the heater should be demonstrated to be functional and operable during the test of heater l
capacity. Ooeration of the heaters will prevent moisture buildup in the filters and adsorber system.
Denonstration of the automatic taltiation capabflity and operability of filter cooling is necessary to assure system performance capability.
If one standby gas treatment system is inoperable, the other systen must be tested daily. This substantiates the availability of the operable system and thus reactor operation may continue.
Amendnent No. 5, 7, //, 76 8 3/4 7-10
\\
3.9 Bases
A.
The objective of the auxiliary electric power availability specification is to assure that adequate power will be available to operate the emergency safequards equipnent. Adequate power can be provided by any one of the following power sources: one 345 kv _line, the 27.6 kv system, the gas turbine-generator and the diesel ger.erator.
This specification assures that at least two offsite and two onsite power sources will be available before the reactor is started up.
In addition to assuring power source operability, all of the associated switch-gear and vital equipment must be operaole as specified to assure that the emergency cooling equipnent can be operated, if required, from the power sources.
B.
fionnally, three 345 kv lines will be available to provide emergency power to the plant when the reactor is operating.
Ilowever, adequate power is available with only one 345 kv line in service. Therefore, reactor operation is permitted for up to seven days with only one 345 kv line in service to acconmodate nect:ssary maintenance, etc.
In the event that all 345 kv lines are cut of service, continued reactor operation is permitted provided both onsite emergency power sources are operating with an adequate fuel supply. Two operational power sources provide an adequate assurance of emergency power ;vailability under these circtanstances.
In addition, the isolation condenser system is required to be ope'.able as a standby heat removal system.
i;ormally both the gas turbine-generator and diesel generator are required to be operable to assure adequate unergency power with no offsite power sources.
However, due to the redundancy and reliability of offsite power, one of the two emergency onsite power sources may be out of service for limited periods of time pro-viding two offsite power sources are available during these periods.
C.
Either of the two station batteries has enough capability to energize the vital buses and power the other emergency equipment.
Due to the high reliability of battery systems, one of the two batteries may be out af service for up to 7 days.
This minimizes the probability of unwarranted shutdowns by providing adequate time for reasonable repairs.
D.
The diesel fuel supply of 20,000 gallons will supply the diesel generator with about five days of full load operation. The gas turbine generator fuel supply of 35,000 gallons is sufficient to operate the unit for at least two and one-half days considering the fuel consumption vs. load and load vs. time requirenents during the postulated accident. Reference Amendment 18. Additional fuel can be supplied to the site within twelve hours.
Amendment flo. 76 B 3/4 9-1
~~
5.0 DESIGN FEATURES S.)
Site I
The Unit 1 reactor building is located on the site at Hillstone Point in Waterford, Connecticut. The nearest
~
site boundary on land is 1620 feet northeast of the reactor building, which is the minimum distance to the boundary of the exclusion area as described in 10 CFR 100.3(a). No part of the site which is closer to the reactor building than 1620 feet shall be sold or leased except to (i) The Connecticut Light and Power Company, The liartford Electric Light Company, Western Hassachusetts f.lectrir Company or the Northeast Nuclear Energy l
Company or their corporate affiliates 'for use in conjunction with nonnal utility operations and (ii) to the tuo leasees under the leases referred to in the following paragraph.
A United State Navy research Laboratory and. a desalination pilot operation of the Maxim Evaporator Division of the Cuno Engineering Corporation may be permitted to operate within the exclusion area under leases which make activities and persons on the leased premises subject to health and safety requirenents of the owners of the site.
5.2 Reactor A.
The core shall consist of 580 fuel assemblies.
N B.
The reactor core shall contain 145 cruci form-shaped control rods. The control material shall be boron carbide powder (B C) compacted to approximately 70% of theoretical density.
4 5.3 Reactor Vessel The reactor vessel shall be as described in Table IV-1 of the FSAR.
The applicable design codes shall be as described in Table IV-1 of the FSAR.
S.4 Containment A.
The principal design parameters and applicable design codes for the primary contalmnent shall be as given in Table V-1 of the FSAR.
B.
The secondary containment shall be as described in Section V-3 of the FSAR and the applicable codes shall be as described in Section XII of the FSAR.
Amendment No. //, 76 S-1
c e
~ ~ '....
s 5.5 Proce'dures 5.5.1 Detailed written procedures, including applicable check lists and ins:ructic s,
shall be prepared and followed for all activities i=volved in carrying cut the environ = ectal tech =1 cal specificatic,s.
?:ocedures shall include sa=pling, data recording and storage, instre:ent ealioration, censure en:s and anal-yses, and actions to be taked v'.e li=1ts are approached or exceeded. Testing frequency of any alar =s shall be included.
These frequancies shall be de e:-
=ined fro = experience with similar instru=en:s in si=1lar envirec=ents and fro =anufacturers' tech =ical =anuals.
?:ocedures shall be prepared for assuring the quality of progra= results, including analytical =casure=ents, which doce=ent the progra= in policy direc:ives, designate a responsible organi:stien or individuals, include purchased services, (e.g., centrac:ual lab c: c:her cen::ac: s ervices),
include audits by licensee personnel, and include sys:e=s to identify and correct deficiencies, investigate ano:21ous or suspec: resul:s, and review and evaluate progra= results and repor:s.
5.5.2
-Plant standard operating procedures shall include provisions, in addi:1cn to the procedures specified in Sectica 5.5.1. to insure that all plant systens and ce=penents are operated in co:pliance with the li=i:ing ec di: ions for opera:1cus established as part of the enviro:: ental technical specificaricas.
5.5.3 All procedures described above and changes :nerato shall be reviewed prior to i=ple=entation as follows:
a.
By PORC/SORC as applicable for precedures rela:ed to Sections 1 and 2 and these parts of Sectic 5 related :c PORC/SORC.
b.
By a qualified individual of :he Inviren= ental ?:ogra=s Branch, other than :he auther, f e: procedures rela:ed to Sections 3 and a anc these par:s cf Sec:ica 5 ne related
~
to PORC/SORC.
c.
Te=porary changes to procedures reviewed'by PORC/50RC should be made in accc: dance vi:h see:ien 5.1.6.1, d.
Te=porary changes to procedures reviewed-by. :he Inviro:: ental Progra:s Bra =ch, which dc not change the inten of the original procedure =ay be =ade provided the change is doce ented and reviewed by a =e=ber of the Inviren=e :a1 ?:esra=s Branch with-in 14 days of i=ple:entation.
5.5.4
? ocedures described above shall. Le reviewed se=1-annually by the Inviren=enta:
Review 3cstd as specified in See:ica 5.3.2.
Appendix B Amendment No. 76
-