ML19260D114
| ML19260D114 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 01/08/1980 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19260D115 | List: |
| References | |
| NUDOCS 8002070435 | |
| Download: ML19260D114 (17) | |
Text
po nto,
u UNITED STATES y
e g
NUCLEAR REGULATORY COMMISSION 3
E WASHINGTON, D. C. 20555 f
o
'59.....,d BOSTON EDIS0N COMPANY DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION, UNIT NO 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 39 License No. DPR-35 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The applications for amendment by the Boston Edison Company (the licensee) dated March 10, 1977, December 28, 1977, November 13, 1979 and November 21, 1979, comply 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 applications, 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.
2.
Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 3.B of Facility License No. DPR-35 is hwcby amended to read as follows:
3.B Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 39, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
1935 003 8002070
. 3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 2A2>
omas
' ppolito, Chief Operating Reactors Branch #3 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
January 8,1980 1935 004
ATTACHMENT TO LICENSE AMENDMENT NO. 39 FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 Revise Appendix A as follows:
Remove the following pages and insert identically numbered pages:
82 89B 104 111 112 119 152 152B 166 203 205 205B 205C-6 Insert the following new page:
205H 1935 005
CONDI"'
70R OPERATION SURVEIll.ANCE REQUIREMENTS 3.3.B Control Rods 4.3.B Control Rods 2.
The control rod drive housing b.
When the rod is fully support system shall be in withdrawan the first time place during reactor power subsequent to each re-operation and when the reactor fueling outage or after coolant system is pressurized maintenance, observe that above atmospheric pressure with the drive does not go to fuel in the reactor vessel, the overtravel position.
unless all control rods are fully inserted and Specifica-tion 3.3.A.1 is set.
2.
The control rod. rive hous-support system shall be in-3.
a.
No control rods shall be spected after reassbmely moved when the reactor is and the results of the below 20% rated power, inspection recorded.
except to shutdown the reactor, unless the Rod 3.
Prior to control rod with-Worth Minimizer (RWM) is drawal for startup or in-operable. A maximum of two sertion to reduce power rods may be moved below 20%
below 20% the operability design power when the RWM of the Rod Worth Minimizer is inoperable if all other (RWM) sha'.1 be verified by:
rods except those which cannot be moved with control rod drive pressure are fully a.
verifying the correctness inserted.
of the control rod with-drawal sequence input to b.
Control rod patterns and the RWM computer.
the sequence of withdrawal or insertion shall be b.
performing the RWM computer established such that:
diagnostic test
- 1) when the reactor is c.
verifying the annunciation critical and below 20%
of the selection errors of design power the maximum at least one out-of-sequence worth of any insequence control rod in each distinct control rod which is RWM group not electrically dis-armed is less than 0.010 d.
verifying the rod block delta k.
function of an out-of-sequence control rod which
- 2) and when the reactor is withdrawn no note than is above 20%' design three notches.
power the maximum worth of any control rod, including allowance for a single operator error, is less than 0.020 delta k.
Amendment No.
39 82 1935 006
3.3 and 4.3 BASES:
When THERMAL POWER is greater than 20% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the vslocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus requiriag the RWM to be OPERABLE when THERMAL POWER is less than or equal to 20% of RATED THERMAL POWER provides adequate control.
We are therefore requiring as a limiting condition of operation (LCO) that the Rod Worth Minimizer (RRM) be operable when the reactor is critical and below 20% of design power in accordance with Specification 3.3.B.3a so that the maximum in-sequence control rod worth will be limited to 0.010 delta k as given in Specification 3.3.B.3b(1) even assuming a single failure of the RWM or an operator error. The RWM assists and supplements the operator with an effective backup control rod monitoring routine that enforces adherence to pre-established startup, shutdown, and low power level control rod procedures. The RWM computer prevents the operator from establishing control rod patterns that are not con-sistent with prestored RWM sequences by initiating appropriate rod select block, rod withdrawal block, and rod insert block - interlock signals to the reactor manual control systems rod block circuitry.
Reference:
FSAR Section 7.16.4.3.
The RWM sequences stored in the computer memory are based on control rod withdrawal procedures designed to limit the individual control rod worths to levels given in Specification 3.3.B.3.b.
Two exceptions to the requirement for RWM operability are permitted. Control rods may be moved to shutdown the reactor, and up to two control rods esn be moved provided all other rods, except those which cannot be moved with control rod drive pressure, are inserted. The first exception permits the operator to shutdown the reactor 1". the event the RWM should become inoperable while the reactor is critical. In this case, the operator is moving the rods to reduce the reactivity in the core. Outward movement of any control rod is limited to a short adjustment and the general sequence of control rod movement is always toward a safer pattern during shutdown operations. The second exception permits the control rod drives to be moved when the RWM is inoperative provided that all but two rods are fully inserted except for those control rods which cannot be moved with control rod drive pressure.
1935 007 Amendment No.
39 ggB
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE EQUIPME'..'T 3.5.A Core ?prav and LPCI Subsvstems 4.5.A Core Sprav and LPCI Subsystems (cont'd)
(cont'd)
Check Once/ day Calibrate once/3 months Tese Once/3 months 2.
Frem and after the datr that one 2.
When it is determined that one core of the core spray subsystems is spray subsystem is inoperable, made or found to be inoperable the operable core spray subsystem, for any reason, continued reactor the LPCI subsystem and the diesel operation is permissitie during generators shall be demonstrated to the racceeding seven days, pro-be operable immediately. The oper-vided that during such seven days able core spray subsystem shall be all active components of the other demonstrated to be operable daily core spray subsystem and active thereafter.
components of the LPCI subsystem and the diesel generators are op-erable.
3.
The LPCI Subsystems shall be oper-3.
LPCI Subsystem Testing shall be as able whenever irradiated fuel is fdllows:
in the reactor vessel, and prior to reactor startup from a Cold a.
Simulated Automa-Once/ Operating Condition, except as specified tic Actuation Test Cycle in3.5.A.4,3.5.A.5and3.5.F.5.l b.
Pu=p Operability Once/ month c.
Motor Operated Once/ month valve operability d.
Pump Flow Rate Once/3 months Three LPCI punps shall deliver 14,400 gpm against a system head correspondin? to a vessel pressure cf 20 psig.
1935 008 104 Amendment No. 39
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.5.F Minimum Low Pressure Cooling and Diesel Generator Avail-ability (Cont'd) 3.
When irradiated fuel is in the re-actor vessel and the reactor is in the Cold Shutdown Condition, both core spray systems, the LPCI and containment cooling subsystems may be inoperable, provided no work is being done which has the potential
.for draining the reactor vessel.
4.
During a refueling outage, for a period of thirty days, refueling operation may continue provided that one core spray system or the LPCI system is operable or spec-ification 3.5.F.5 is met.
5.
When irradiated fuel is in the reactor vessel and the reactor is in the Refueling Condition with the torus drained, a single control rod drive mechanism =ay be removed, if both of the following conditions are satis-fied:
a) No work on the reactor vessel, in addition to CRD removal, will be performed which has the po-tential for exceeding the maximum leak rate from a single control blade seal if it became unseated.
I b) 1) the core spray systems are operable and aligned with a suction path, from the condensate storage tanks, 11) the conden-sate storage tanks shall contain at least 200,000 gallons of usable water and the refueling cavity and dryer / separator pool shall be flooded to at least elevation 114'-0".
j I
1935 009 (Intentionally left blank)
Amendment No.
39 111
LIMITINC CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.5.H Maintenance of Filled Dis-4.5.H Maintenance of Filled Discharge Pipe charge Pipe Whenever core spray subsystems, LPCI The following surveillance requirements subsystem, HPCI, or RCIC are required shall be adhered to to assure that the to be operable, the discharge piping discharge piping of the core spray sub-from the pump discharge of these sys-systems, LPCI subsystem, HPCI and RCIC tems to the last block valve shall be are filled:
filled.
1.
Every month prior to the testing of the LPCI subsystem and core spray subsystem, the discharge piping of these systems shall be vented from the high point and water flow cb-se rved.
2.
T ' lowing any period where the LPCI subsystem or core spray subsystems have not been required to be oper-able, the discharge piping of the inoperable system shall be ventea from the high point prior to the return of the system to service.
3.
Whenever the HPCI or RCIC system is lined up to take suction frem the torus, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.
4.
The pressure switches which monitor the discharge lines to ensure that they are full shall be functionally tested every month and calibrated every three months.
1935 010 Amendment No-39 112
BASES:
3.5.F Minimum Low pressure Cooling and Diesel Generator Availability The purpose of Specification F is to assure that adequate core cooling equip-ment is available at all times.
If, for example, one core spray were out of service and the diesel which powered the opposite core spray were out of service, only 2 LPCI pu=ps would be available.
It is during refueling outagri that major maintenance is performed and during such time that all low pres-sure core cooling systems may be out of service. This specification provides that thould this occur, no work will be performed on the primary system which could lead to draining the vessel. This work would include work on certain control rod drive components and recirculation system. Specification F allows removal af one CRD =echanism while the corus is in a drained condition without compromising core cooling capability. The available core cooling capability for a potential draining of the reactor vessel while this work is perfor=ed is based on an estimated drain rate of 300 gpm if the control rod blade seal is unseated. Flooding the refuel cavity and dryer / separator pool to elevation 114' 0" corresponds to approximately 350,000 gallons of water and will provide core cooling capability in the event leakage from the control rod drive does occur.
A potential draining of the reactor vessel (via control rod blade leakage) would allow this water to enter into the torus and after approximately 140,000 gallons have accumulated (needed to meet mini =um NPSH requirements for the LPCI and/or core spray pu=ps), the torus would be able to serve as a co==on suction header.
This would allow a closed loop operation of the LPCI system and the core spray system (once re-aligned) to the torus. In addition, the other core spray system is lined up to the condensate storage tanks which can supplement the refuel cavity and dryer / separator pool water to provide core flooding, if required.
Specification 3.9 must also be consulted to determine other requirements for the (.iesel generators.
1935 01i 119 Amendment No.
39
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS
/~alicability:
Applicability:
Applies to the operating status of the primary Applies to the primary and secondary and secondary containment systems.
containment integrity.
Objective:
Objective:
To assure the integrity of the primary and To verify the integrity of the primary secondary containment systems, and secondary containmant.
Soecification:
Specification:
A.
Primary Con tainment A.
At any time that the nuclear system is 1.
The suppression chacher water pressurized above atmospheric pressure level and temperature shall or work is being done which has the be checked once per day, potential to drain the vessel, the pressure suppression pool water volume b.
Whenever there is indication and te=perature shall be maintained of relief valve operation or within the following limits except as testing which adds heat to the specified in 3.7.A.2 and 3.7.A.3.
l suppression pool, the pool temperature shall be con-3 a.
Minimum water volume - 84,000 f t tinually monitored and also observed and logged every 5 b.
Maximum water volume - 94,000 ft minutes until the heat addition is terminated.
c.
Maximum suppression pool temperature during normal continuous power c.
Whenever there is indication operation shall be $ 80 F, except as of relief valve operation with specified in 3.7.A.I.e.
the temperature of the suppression pool reaching 160,F d.
Maximum suppression pool temperature or more and the primary coolant during RCIC, HPCI or ADS operation system pressure greater than shall be 1 93 F, except as specified 200 psig, an external visual in 3.7.A.l.e.
examination of the suppression chamber shall be conducted e.
In order to continue reactor power before resuming power operation.
operation, the suppression chamber pool temperature must be reduced to d.
A visual inspection of the 180 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
suppression chamber interior, including water line regions, f.
If the suppression pool temperature shall be made at each major exceeds the limits of Specification refueling outage.
3.7.A.l.d, RCIC, EPCI or ADS testing shall be terminated and suppression pool cooling shall be initiated.
g.
If the suppression pool temperature during reactor power operation exceeds 110 F, the reactor shall be scranmed.
1935 012 Amendment No.
39 15 2
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
?..
Primary containment integrity shall be 2.
Integrated Leak Rate Testing maintained at all times when the The primary containment reactor is critical c.
or when the reactor water integrity shall be demon-temperature is above 212*F and strated by performing an fuel is in the reactor vessel except Integrated Primary Con-while performing "open vessel" physics tainment Leak Test (IPCLT) tests at power levels not to exceed in accordance with either 5 Mw(t).
Method A or Method B, as follows:
Method A Perform leak rate test prior to init.a1 unit operation at the test pressure 45 psig, P; (45), to obtain measured leak rate Lm (45), or Method B Perform leak rate test prior to initial unit operation at the test pressure of 45 psig, P (45), and 23 psig, Pe (23),
g to cbtain the measured leak rates, L (45) and L, (23),
3 respectively.
3.
The suppression chamber can be drained if the conditions as specified in Sections 3.5.F.3 and 3.5.F.5 of this Technical Specification are adhered to.
152B 1935 013 Amendment No.
39
BASES:
- 3. 7. A & 4. 7. A Primary Containment The integrity of the primary containment and operation of the core standby cooling system in combination limit the off-site doses to values less than those suggested in 10 CFR 100 in the event of a break in the primary system piping.
Thus, containment integrity is specified whenever the potential for violation of the primary reactor system integrity exists.
Concern about such a violation exists whenever the reactor is critical and above atmospheric pressure.
An exception is made to this requirement during initial core load-ing and while the low power test program is being conducted and ready access to the reactor vessel is required.
There will be no pressure on the system this time, thus greatly reducing the chances of a pipe break.
at The reactor may be taken critical during this period; however, restrictive operating procedures will be in effect again to minimize the probability of an accident occurring.
Procedures and the Rod Worth Minimizer would limit control worth such that a rod drop would not result in any fuel damage.
In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shcIl be operational during this time, offer a sufficient barrier to keep off-site doses well below 10 CFR 100 limits.
The pressure suppression pool water provides the heat ains. for the reactor primary system energy release following a postulated rupture of the system.
The pressure suppression chamber water volume must absorb the associated decay and structural sensible heat released during primary system blowdown from 1035 psig.
Since all of the gases in the drywell are purged into the pressure supression chamber air space during a loss-of-coolant accident, the pressure resulting frcm isother=al ce=pression plus the vapor pressure of the liquid must not exceed 62 psig, the suppression chamber maximum pressure.
' volume of the suppression cha=ber (water and air) was obtained by considering The design that the total volu=e of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the suppression cha=ber.
Using the minimum or maximum water volumes given in the specification, containment pressure during the design basis accident is approxi=stely 45 psig which is below the maximum of 62 psig.
Maximum water volume of 94,000 ft results in a downcomer submergency of 4'9" and the minimum volume of 84,000 ft 3 results in a' submergence approx 1=stely 12-inches less.
The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation.
Thus, with respect to downcomer submergency, this specification is adequate, Shi
$~ be necessary to drain the suppression chamber, provision will be made to cr'
.a those requirements as described in Section 3.5.F BASES of this Tche Jpecification.
Experimental data indicates that excessive steam condensing loads can be avoided if the peak temperature of the pressure suppression pool is maintained below 1600F during any period of relief-valve operation with sonic conditions at the discharge exit.
Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high pressure suppression chamber loadings.
Amendment No. 39 166
Exhibit A LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 2.
The SRM shall have a minimum c.
Spiral Reload of 3 cps except as specified in 3 and 4 below.
During spiral reload, SRM operability will be verified by using a portable 3.
Prior to spiral unloading, the external source every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until SRM's shall have an initial the required amount of fuel is loaded count rate of > 3 cps.
During to maintain 3 cps. As an alternative spiral unloading, the count to the above, up to two fuel assemblies rate on the SRM's may drop will be loaded in different cells below 3 cps.
containing control blades around each SRM to obtain the required 3 cps.
' +. During spiral reload, each Until these assemblies have loaded, control cell shall have at the cps requirement is not necessary.
least one assembly with a minimum exposure of 1000 MWD /t.
39 Amendment No.
1935 015 23
3.10 EASES B.
Ccre Monitorig The SRM's are provided to nicritor the core during periods of station shutdown and to guide the operator during refueling operations and station startup.
RNguiring two operable SRM's in or adjacent to any core quadrant where fuel or control rods are being coved assures adequate monitoring of that anadrant during such alterations. The requirement of 3 counts per second pr m des assurance that neutron flux is being monitored and insures that startup is conducted only if the source range flua level is above the minimum assumed in the control rod drop accident.
The limiting conditions for operation of the SRM subsysrem of the Neutron Monitoring System are derived from the Str. tion Nuclear Safety Operacional Analysis (Appendix G) and a functional coalysis of the neutron monitoring system. The specification is based or. the Operational Nuclear Safety Re-quirements in subsection 7.5.10 of the Safety Analysis Report.
A spiral usloading pattern is one by which the fuel in the outemost cells (four fuel bundles surrounding a control blade) is removed first. Unloading continues by removing the remaining outermost fuel cell by cell. The center cell vill be the last removed. Spiral loading is the reverse of unloading. Spiral unloading and reloading vill preclude the creatio ; of flux traps (moderator filled cavities surrounded on all sides by fuel).
During spiral unloading, the SRM's shall have an initial count rate of 2 3 cps with all rods fully inserted. The count rate vill diminiah during fuel removal.
Under the special condition of complete spiral core unloading, it is expected that the count rate of the SRM's vill drop below 3 cps before all of the fuel is unloaded.
Sirice there vill be no reactivity additions, a lower number of counts will not present a ha:ard. When all of the fuel has been removed to the spent fuel storage pool, the SRM's will no longer be required. Requiring the SRM's to be operational prior to fuel removal assures that the SRM's are operable and can be relied on even when the count rate may go below 3 eps.
During spiral reload, SRM operability vill be verified by using a portable exter-nal;sourcs every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until the required amount of fuel is loaded to maintain 3 cps. As an alternative to the above, up to two fuel assemblies will be loaded in different cella containing control blades around each SRM to obtain the required 3 cps.
Until these assemblies have been loaded, the 3 cps requirement is not necessary.
C.
Soent Fuel Pool Water Level To assure that there is adequate water to shield and cool the irradiated fuel assemblies stored in the pool, a minimum pool water level is established.
The minimum water level of 33 feet is established because it would be,a significant change from the normal level (-1 foot) and is well above the level to assure adequate cooling.
4.10 BASES 1935 016 A.
Refueling Interlocks 7
Complete functional testing of all refueling interlocks before any refueling if outage vill provide positive indication that the interlocks operate in the S
situations for which they were designed. By loading each hoist with a weight equal to the fuel assembly, positioning the refueling platform, and with-drawing control rods, the interlocks can be subjected to valid operational tests. Where redundancy is provided in the logic circuitry, tests can be performed to assure that each redundant logic element can independently w
perform its functions.
B.
Core Monitoring Requiring the SRM's to be functionally tested prior to any core alteration assures that the SRM's will be operable at the start of that alteration.
205 The daily response check of the .Ji's ensures their continued operability.
LIMITING CO'Q: "'ICNS FC JPERATION SURVEIL.
.CE FIQUIREMENTS C.
Minimum Critical Power Ratio (MCPR)
C.
Minimum Critical Power Ratio (MCPR)
During power operation MCPR shall be MCPR shall be determined daily
> 1.31 for 8x8 fuel.
If any ti$e during reactor power operation at during operation it is determined
> 25% rated thirr.al power and by normal surveillance that the following any et ange in power limiting value for MCPR is being level or distri:ation that would exceeded, action shall be initiated cause operation with a limiting within 15 minutes to restore operation control rod pattern as described to within the prescribed limits within in the bases for Specification two (2) hours, the reactor shall be 3.3B.5.
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.
For core flows other than rated the MCPR shall be 21.31 for 8x8 fuel times Kg, where Y.c ic ac chewn in Figure 3.11-8.
As an alternative method providing equivalent thermal-hydraulic protec-tion at core flows other than rated, the calculated MCPR may be divided by Kf, where Kf is as shown in Figure 3.11-8.
D.
Power / Flow Relation::hio Durina Power D.
Power / Flow Relationship During Ooeration Power Operation The power / flow relationship shall not Compliance wit.h the power / flow exceed the limiting values shown in relationship in Section 3.ll.D shall Figure 3.11-9.
If at any time during be determined daily during reactor power ope' ration it is determined by operation.
normal surveillance that the limiting value for the power-flow relationship is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the power / flow relation-ship is not re arned 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 corres-ponding action shall continue until reactor operation is within the pre-i935 017 scribed limits.
Amendment No. 39
factors shown in Figure 3.11-8 (5) are conservative for the The Kg Pilgrim Unit 1 operation because the operating li=it MCPR given in Specification 3.11C is greater than the original 1.20 operating limit MCPR used for the generic derivation of Kg.
- 4. llc MINIMUM CRITICAL POWER RATIO (MCPR) - JURVEILI.ANCE REQUIRE N At core thermal power levels less than or equal to 25%, the reactor will be operating at minirm recirculation pu=p speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the resulting MCPR value is in eNeess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only piace operation in a more conservative mode relative to MCPR. During initial start-up testing of the plant, MCPR evaluation will be made at 25% thermal power level with t 11 mum recirculation pump speed. The MCPR =argin will thus be demonstrated such that future MCPR evaluation belcw this power level will be shown to be unnecessary. The daily re-quirement for calculating MCPR above 25% rated thermal power is sufficient since power distribution shifts are ver/ slow when there have not been significant power or control rod changes.
The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in power or power shape (regardless of magnitude) that could place operation at a thermal 'imit.
Power / Flow Relaticnshia Bases P
The power / flow curve is the locus of core thermal power as a function of flow frem which the occurrence of abnormal opera-ting transients will yield results within defined plant safety limits. Each transient and postulated accident applicable to operation of the plant was analyzed along the power / flow line.
The analysis justifies the operating envelope bounded by the power / flow curve as long as other operating limits are satis-fi ed. Operation under the power / flow line is designed to enable the direct ascension to full power within the design basis Tor the plant.
1935 018 205C-6 Amendment No.
39
7 CORE THERMAL POWER MW6 A
I I
k h
o 0
0 8
o o
4 g
l As
)
a -o.
33 g3 En\\ ii 11 ih i
19 4
sA n
tw k
' \\;h,i i
L [g 1(5
\\
nt s
is i
l o
h kN
~
I a *l 3
i i
i o J l
i I
~\\
I 1
I I
g Ll b
I
- ft [
\\
\\
I I
g g
I
'kh 1,
I I
l 8'.%
i i
it t
I 5; b.
R
-j-t s
j 7
o"
\\1 I
h-Yh
%% %I E
w g
I i R
I 4 'N
's s
k
-\\-g i ut -4 l
l 1
x o
B is a
a e
e s
e 8
j g
g P&ecENroFmrfo powen lYo ecrp y9)5 0\\9 Amendment No.
39 205H
_ _...... _...