ML19347F317
| ML19347F317 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 05/02/1981 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML19347F316 | List: |
| References | |
| NUDOCS 8105180359 | |
| Download: ML19347F317 (33) | |
Text
. . . . . . . . . _. - . . ,
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ATTACIMENT B
$ License
. and Technical' Specification '. hanges a
f
-Lic. .Page 3.
T.S. Pages 6 ;
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12 13' !
14 I 15 :
- 18
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127 I t
12iA 127A-1 ,
I 205A- "
205B 205C ;
205C-2 '[
205C-3 205C-4 :
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4- B. . The Technical Specifications contained in Appendices A r . 4, as revised through Amendment No. ~ 48. are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
C. Records Boston Edison shall keep facility operating records in accordance with t'.te requirements of the Technical Specifications.
' D. Equalizer ' Valve Restriction The valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor operation.
E. (Deleted)
F. Fire Protection The licensee may proceed with and is required to complete the modifications identified in Paragraphs 3.1.1 through 3.1.19 of the NRC's Fire Protection Safety Evaluation (SE), dated December 21, 1978 for the facility. These modifications will be completed in accordance with the schedule in Table 3.1.
In addition, the licensee shall submit the additional information identifled in Table 3.2 of this SE in accordance with the schedule contained therein.
In the event these dates for submittal cannot be met, the licensee shall sub-mit a report, explaining the circumstances, together with a revised schedule.
The licensee is required to implement the administrative controls identified in Section 6 of the SE. The administrative controls shall be in effect by December 31, 1978.
G. Physical Protection i The licensee shall fully implement and maintain in effect all provisions of the following Commission approved documents, including amendments and changes made pursuant to the authority of 10 CFR 50.54(p):
(1) " Security Plan for the Pilgrim Nuclear Power Station", dated Novmeber 7, 1977 with Revision 2 dated May 26, 1978 and Revision 3 dated January 8, 1979.
(2) " Pilgrim Nuclear Power Station Safeguards Contingency Plan", dated April 5, 1979 and revised by letters dated December 20, 1979 and April 22, 1980, submitted pursuant to 10 CFR 73.40. The Contingency Plan shall be fully implemented, in accordance with 10 CFR 73.40(b), within 30 days of the approval by the Commission.
(3) " Pilgrim Nuclear Power Station Guard Training cnd Qualification Plan",
i Revision 3, dated October 1980 includes pages dated August 18, 1979, ;
i May 28, 1980, and October 1, 1980. This Plan shall be followed in ,
L accordance with 10 CFR 73.55(b)(4), 60 days af ter approval by the Commis- g sion. All security personnel, as required in the above plans, shall be .
qualified within two years of this approval. The licensee may make changes to this plan without prior Commission approval if the changes do not !
decrease the safeguards effectiveness of the plan. The licensee shall l maintain records of and submit reports concerning such changes in the l
>1.1
_ S'AFETY LIMIT 2.1 tbfITIhG SAFETY SYSTDi SETTINO 1.1 TITEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY
_Asplicability: Applicability:
AP Plies to t'ne interrelated Applies to trip settings of the variables associated with fuel instruments and devices which are ther:r.a1 behavior. provided to prevent the reactor system refety limits from being exceeded.
Objective: Objective:
To establish limits belov which To define the level of ths irocess integrity of the fuel variables at which automatic pro-1
. , is preserved. tactive action is initiated to prevent the fuel cir.dding integrity safety limits from being exceeded.
Soecification: Specification:
A. Reactor Pressure > 800 psia and A. Neutron Flux Scram Core Flev >10% of Rated The existence of a minimum critical The limiting safety systea trip power ratio (MCPR) less than 1.07 settings shall be as specified for two recirculation loop opera- below: ,
tion (1.08 for single-loop opera-tion) shall constitute violation of 1. Nuetrer, Flux Trip Settins:s the fuel cladding integrity safety ~
limit. This value of the MCPR is hereinaf ter referred to as the Safety Limit MCPR.
Pressure 1800 psia and/or Core Flev s10%)
- When the Mode Switch is in the RUN position, the iraen the reactor pressure is $800 AFRM flux scran trip psia or core flov is less than or setting shall be:
equal to 10% of rated, the steady Sf.65W + 55% 2 loop state cord thermal power shall not Sf.65W + 51.7% 1 loop exceed 25% of design thermal power.
Where:
C. Power Transient S = Setting in percent The safety limit shall be assumed of rated thermal to be exceeded when scram is known power (1998 MWt) to have been accomplished by a ceans other than the expected W = Percent of drive scram signal unless analyses flow to produce demonstrate that the fuel a rated core flow cladding integrity safety of 69 M lb/hr.
limits defined in Specifi- .
cations 1.1A and 1.13 vere not exceeded during the actual transient.
6
e . , -
. a 1.1 . SAFEIT LIMIT 2.1 LIMITING SAFETY SYSTEM SETIING D. Whenever the reactor is in the '
In the event of operation with a ccid shutdown condition with maxistan fraction of limiting power irradiated fuel in the reactor density (NFLPD) greater than the ;
vessel, the water level shall not fraction of rated power (FRP),
be less than 12 in. above the top the setting shall be modified as of the normal active fuel sone. follows: -
ppy ;
S f (0.65W + 55% ) . MFLPD b E
, S g (0.65W + 51.7%) ~
FRP = fraction of rated thermal power (1998 MWt)
MFLPD = maximum fraction of limiting >
power density where the limiting power density is '
13.4 KW/f t for 8x8 and P8x8R fuel.
The ratio of TRP 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 i the actual operating value will be used.
For no combination of loop recircula-tion flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thernal power.
When the reactor mode switch is !
in the REFUEL or STARTUP position, the APRM scram shall be set at less than or equal to 15% of rated power.
B. APRM Rod Block Trip Setting The APRM rod block trip setting shall be:
3 RE 6 O.65W + 42% 2 loop i S
- , 7 RB f 0.65W + 38.7% 1 Loop I
~..
1.1 S4?ETY LIMIT k.1 LIMITING SAFETY SYSTEM SETTING Where, b = Rod block setting in percant of rated thermal power (1998 MWt)
W = Percent of drive flow required -
to produce a rated core flow of 69M lb/hr.
In the event of operating with a maxinum fraction limiting power ,
density (MFLPD) greater than the fraction of rated power (FRP), the i setting shall be modified as follows:
~
S I' FRP 2 bop RB g (0.65 W + 42%) 'l r LPD ,
1 Loop l RB f (0.65 11 + 38.7%~ FRP 5 MFLPD ,
Where, -
FRP = fraction of rated thermalpower MELPD= maximum fraction of limiting power density where the limiting power density is 13.4 KW/f t for 8x8 and P8x8R fuel.
The ratio cf 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 operating value will be ,
i used. ,
C. Reactor low water level scram setting shall be2 9 in. on level i instruments. ;
D. Turbine stop valve closure scram setting shall be s.10 percent valve closure.
E. Turbine control valve fast closure setting shall be 2.150 psig con-trol oil pressure at acceleration relay.
F. Condenser low vacuum scram setting shall be 2 23 in. Eg. vacuum.
G. Main steam isolation scram sr.tting shall be $ 10 percent valve clo- !
sure. ;
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-=- .AP. .R.M.. Flow p '/ gs._.-
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E 901 -
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ff , g -
1-s-/-
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- . 80. --/
.# f-f--v
_ __ -- , ,, .AP.R.M.. . Rod. Blo.ck --
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.-. :- - 4c* *5
~:- - -
5" - (Normal)31_* ~___
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ac = 2- ot*<_ .- sq? - =- -
y
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_-- e = are varied by the rs.tio N - + - - !
-_ __- MFLPD w
=
[.2~!/ See Specifications 2.1.A and 2.1.3 ;
. ko/% __ _ _.__
/-. ,
m.
w -
--- *2 When in the refuel cr sta.rtup/ hot stanity __' -
5 .-- - - - n= des, the 'APRM scram shall be set at :G.
e:
3v.. . ._
15% of design power ;
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sa -
a: :. 2 C_- _- - .
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Figure 2.1.1
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M .= _ . _ _ . . . - . . . .
- ==_ ._A_PR.M_.Scr.ad an'd lied 'Blo'cE Trip Limiting Safety System Settings -
-- :=. ::.- .
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P00R OR BA.
n-BASES:
' 1.1 FUEL CLADDING INTEGRITY A. Fuel Cladding Integrity Limit at Reactor Pressure 2 800_ psia and Core Flowl 10% of Rated The fuel cladding integrity safety limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since
. the parameters which result in fuel damage are not directly observable during reactor operation the 'hermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling-transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedure used to calculate the critical power result in an uncertainty in the value of the critical power.
Therefore the fuel cladding integrity safety limit is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling tran-sition considering the power distribution within the core and all un-certainties. .
The Safety Limit MCPR is generically determined in Reference 1 (page 13) for two recirculation loop operation. This safety limit MCPR is increased by 0.01 for single loop operation as discussed in Reference 2 (page 13).
B. Core Thermal Power Limit (Reactor Pressure <800 psig or Core Flow T10%
of Rated)
Since the pressure drop in the bypass region is eocentia 1y all elevation head which is 4.56 psi the core pressure drop at los pov<c and all flows will always be greater than 4.56 psi. Analyses show tnac with a flow of 28x103 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 28x103 lbs/hr irraepective of total core flow and independent of bundle power for the range of bu. ile powers of concern. Full scale ATLAS 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 the 3.35 MWt bundle power corresponds to a core thermal power of more than 50%.
Therefore a core thermal power limit of 25% for reactor pressures below 800 psia, or core flow less than 10% is conservative.
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4 (THIS PAGE INTENTIONALLY LEFT SLANK) 12
C. Power Transient Plant safety analyses have shown that the scrams caused by ex-ceeding any safety setting vi u assure that the Safety Limit of Specification 1.1A or 1.13 vill not be exceeded. Scram times ' -
are checked periodically to assure the insertion times are adequate. The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g.,
scram from neutron flux following closures of the main turbine stop valves) does not necessarily cause fuel damage. However ,
for this specification a Safety Limit violation vill be assumed when a scram is cely accomplished by means of a backup feature o f the plant design. The concept of not approaching a Safety Linit provided scrsa signals are operable is supported by the extensive plant safety analysis. .
The computer provided with Pilgrim Unit 1 has a sequence annunciation program which will indicate the sequence in which events such as scram, APRM trip initiation, pressure scram initiation, etc. occur. This program also indicates when the scram setpoint is cleared. This will provide infor=ation on how long a scram condition exists and thus provide some measure ,
of the energy added during a transient.
D. Reactor Vater Level (Shutdown Condition)
During periods when the reactor is shutdown, consideration must also be given to water level requirements due to the effect of decay heat. If reactor water level should drop belev the ;
top of the active fuel during this tine, the ability to cool I the core is reduced. This reduction in core cooling capability i could lead to elevated cladding temperatures and clad perforation.
The core can be cooled sufficiently should the water level be reduced to two-thirds the core height. Establishment of the safety li=1t at 12 inche.;. abcve the top of the fuel provides adequate margin. This level vill be continuously monitored.
References
- 1. " General Electric Boiling Water Reactor Generic Reload Fuel Application", NEDE-24011-P - A-1, July 1979. -i
- 2. " Pilgrim Nuclear Power Station Single-Loop Operation", NEDO-24268, June 1980.
b 13
. i 1
l l
-s BASES: ~ LIMITING SAFETY SYSTEM SETTINGS RELATED TO FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY The abnormal operational transients applicable : to operation of the ~ PNPS 1 Unit have been analyzed throughout the spectrum of planned operating con-ditions up to the thermal power condition of 1998 MWt. The analyses were based upon plant operation in accordance_ with the operating map given in Figure 3.7-1 of tlu FSAR. In addition, 1998 MWt is the licensed maximum.
power level of PNPS 1, and this represents the maximum steady-state power which shall not knowingly be exceeded.
Transient analyses performed each reload are given in Reference 1 (page 20).
Models and model conservatisms are alco described in this reference. As discussed in Reference 2 (page 20), the core wide transient analyses for one recirulation pump operation is conservatively bounded by two-loop operation analyses and the flow-dependent rod block and scram setpoint equations are-adjusted for one-pump operation.
Steady-state operation without forced recirculation will not be permitted, except during startup testing.
4 14
2.1 BASES
The bases for individual set points are discussed below:
A. Neutron Flux Scram Trip Settings APRM The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of design power (1998 MWt). Because fission chambers provide the basic input signals, the APRM system responds rectly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power) is less than the instantaneous neutron flux due to the time constant of the fuel.
Therefore, during abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting. Analyses demonstrated that with a 120 percent scram trip setting, none of the abnormal operational transients analyzed violate the fuel safety limit and there is a substantial margin from fuel damage. Therefore, the use of flow referenced scram trip provides even additional margin.
The flow biased scram plotted on Figure 2.1.1 is based on recirculation loop flow. Figure 2.1.3, which shows the flow biased scram as a function of core flow, has also been included.
An increase in the APRM scram setting would decrease the margin present before the fuel cladding integrity safety limit is reached.
The APRM scram setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurious scrams, which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM setting was selected because it provides adequate margin for the fuel cladding integrity safety limit yet allows operating margin that reduces the possibility of unnecessary scrams. ,
P 15
l
2.1 BASES
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. As with the APRM scram trip setting,'the APRM rod block trip setting is adjusted downward if the -N=
fraction of ifmiting power density exceeds the fraction of rated power, thus preserving the APRM rod block safety margin. Definition of single loop setpoints is given in Reference 3 (page 20).
C. Reactor Water Low Level Scram Trip Setting (LL1) ,
The set point for low level scram is above the bottom of the separator skirt. This level has been used in transient analyses dealing with coolant inventory decrease. The results show that scram at this level adequately protects the fuel and the pressure barrier, because MCPR remains well above the safety limit MCPR in all cases, and system pressure does not reach the safety valve settings. The scram setting is approximately 25 in, below the normal operating range and is thus ,
adequate to avoid spurious scrams.
D. Turbine Stop Valve Closure Scram Trip Setting ,
The turbine stop valve closure scram anticipates the pressure, neutron flux and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram ttf p setting of $ 10 percen't of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR renains above the safety limit MCPR even during the vorst case transient that assumes the turbine bypass is closed.
E. Turbine Control Valve Fast Closure Scram Trip Setting The turbine control valve fast closure scram anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection exceeding the capability of the bypass valves. The reactor protection system initiates
- a scram when fast closure of the control valves is initiated by the acceleration reizy. This setting and the fact that control valve closure time is approximately twice as long as that for the stop valves means that resulting transients, while si=1lar, are less severe than for stop valve closure. MCPR remains above the safety limit MCPR.
i F. Main Condenser Tew Vacuum Scram Trip Sett*_ng !
To protect the main condenser against overpressure, a loss of condenser vacuum initiates autematic closure of the turbine stop valves and turbine 6
- r
__ bypass valves. To anticipate the transient and automatic scram resulting 1 f ron the closure of the turbine stop valves, low condenser vacuum '
initiates a setam.
The low vacuum scram set point is selected to initiate a scram befort the closure of the turbine stop valves is initiated.
18
. j
. 2.'1. BASES:
- Transient and accident analyses demonstrate that these conditions result in adequate safety margins for the fuel.
i References
- 1. " Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station, .
Reload 4", NEDO-24224.
- 2. Final Safety Analysis Report for Pilgrim Nuclear Power Station Unit #1. !
- 3. ' Attachment A, 'Hodified Rod Block Equation" to BECo letter (J. E. Howard) ,
to NRC (T. A. Ippolito) dated May 12, 1981.
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110 100 -
$b
[#
O 90 -
n g
E5 80 -
[# $
g 70 k ,
l 5 @A .
5 60 -
5 -
y 50 - -
NATURAL CIRCULATION E !
E 40 -
20% PUMP SPEED LINE 8 j
- i 30
- 20 t 10 .
f I I I i l I 1 I I 0 - 70 80 90 -100 110 120 .
10 20 30 40 50 60 !
0 CORECOOLANTFLOWRATE(%OFDESIGN) r APRM FLOW BIAS SCRAM VERSUS REACTOR CORE FLOW i FIG. 2.1.3 :
l t
Figure 2.1.3 above represents the APRM two loop flow bias scram with {
neutron flux plotted against core coolant flow rate instead of recirculation loop flow as shown in Figure 2.1.1.
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mu t e ..i nr.m. a un a av a r.t. a tua aiaires ut,auvif ina i tune.nin a ton anyu t at 1r.re t .
Mina. N.cher Mod a in Which Function
- Optrable Inst. Murt Ba Operrbl9 Chinn21c ptr Trip Trip Function Trip Level Satting Refuel (7) Startup/ Hot Run Action (1)
(1) System Standby 1 Mode Switch in Shutdown X X X h.
1 Manual Scram X X X A IRM 3 High Flux 5120/125 of full scale X X' (5) A 3 Inoperative X X (5) A APRM 2 High Flux
- (14) (15) (17) (17) X A or B 2 Inoperative X X(9) X A or B 2 Downscale 2 2.5 Indicated on Scale (11) (11) X(12) A cr 3 2 - High Flux (15%) 515% of Design Power X X (16) A or 3 2 High Reactor Pressure $ 1085 psig X(10) X X A 2 High Drywell Pressure $ 2.5 psig X(8) X(8) X A 2 Reactor Low Water Level 2 9 In. Indicated Level X X X A 2 High Water Level in Scram Discharge Tank - 539 Callons X(2) X '
X A 2 Turbine Condenser Low Vacuum 123 In. Hg Vacuum X(3) X(3) X A or C 2 Main Steam Line High 17X Normal Full Power Radiation Background X X X A or C 4 Main Steam Line Isolation Valve Closure 110% Valve Closure X(3) (6) X(3) (6) X(6) A or C
~
2 Turb. Cont. Valve Fast 1150 psig Control Oil Closure Pressure at Acceleration Relay X(4) X(4), X(4) A or D 4 Turbine Stop Valve Closure 510% Valve Closure X(4) X(4) X(4) A or D U *APRM high flux scram setpoint s(.65W + 55) FRP - Two recirc. pump operation l MFLPD or j(.65W+51.7)MfLPD. one recire. pump operation
PHPS TABLE 3.2.C -
INSTRUMENTATION 111AT INITIATES ROD BLOCKS Minimum i of Operable Instrisment Trip Level Setting Instrument Channels Per Trip Systems (1)_ F
(}
Two Loop (0.65W + 42) gFLPD" APlot Upscale (Flow -
2 Biased)
One loop _ (0.65W + 38.7) D 2.5 indicated on scale APRM Downscale 2 .
l Two Loop (0.65W + 42) --
(2)
Rod Block Monitor ,, .
1 (7) (Flow Biased)
F One Loop (0.65W + 38.7) MFLPD - ,
Rod Block Monitor 5/125 of full scale
- j 1 (7) Downscale IRH Downscale (3) 5/125 of full scale ,,
3 IRN Detector not in (8) 3 - Startup Potiition IRM Upscale <108/125 of full scale .
3 SRM Detector not in (4)
$ 2 (5) Startup Position .
SRM Upscale 110 counts /sec.
2 (5) (6) .
O vo-- m e- .-, -
m-
_ +m es e.e-- -- ~ , '-
--w e ---em.
- LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT 3.6.D Safety and Relief Valves (Cont'd)
(
pressure shall be below 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
E. Jet Pumps ,
E. Jet Pumps ,
- 1. Whenever the reactor is in the Whenever there is recirculation flow startup or run modes, all jet with the reactor in the startup or
- pumps shall be operable. If it is run modes, jet pump operability shall determined that a jet pump is be checked daily by verifying that i
inoperable, an orderly shutdown the following conditions do not oc-shall be initiated and the reactor cur simultaneously:
shall be in a Cold Shutdown Condi-tion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 1. The two recirculatioa loops have a flow imbalance of 15% .
or more when the pumps are opetated at the same speed.
- 2. The indicated value of core flow rate varies from the value derived from loop flow measurements by more than 10%. !
4 I
- 3. The diffuser to lower plenum differential pressure reading on an individual jet pump ,
varies from established ' jet pump P characteristics by ,
, more than 10%.
F. Jet Pump Flow Mismatch F. Jet Pump Flow Mismatch
- 1. Whenever both recirculation pumps Recirculation pump speeds shall.tne are in operation, pump speeds shall checked and logged at least once be maintained within 10% of each per day. ,
other when power level is greater than 80% and within 15% of each -
other when power level is less ;
than or equal to 80%. ;
- 2. Single loop reactor operation is 1ot .
permitted for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> :
unless the following designated ad-justments are made for ApRM rod r block and scram setpoints (Tech. !
Spec. 2.1.A and B) RBM setpoint ,
(Table 3.2.C), r^"R fuel cladding i integrity safeLi :. nit and operating. [;
limits (Tech. :~. 1.1.A and 3.11.C.
respectively),andMAplHGR(Tech ,
Spec. 3.11.A). !
127 l
c-1
.' , LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENT G.- Structural Integrity G. , Structural Integrity
- 1. The structural integrity of 1. The nondestructive inspections the primary system boundary listed in Table 4.6.1 shall be shall be maintained at the performed as specified. The ,
level required by the ASME results obtained from compliance Boiler and Pressure Vessel with this specification- will be -
Code,Section XI,." Rules evaluated af ter 5 years and the for Inservice Inspection of- conclusions of this evaluation' Nuclear Power Plant will be reviewed with AEC.
Components," 1974 i
r 7
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i 127A l
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1.S.C Structurn1_IntenritY (C!n't)
Edition (ASNE Code,Section XI). In the interim until the nuclear system piping inspection evaluatton Icvel criteria of the ASNE Boiler end Pressure Vessel Code,Section XI, 1974 Edition .are ,
completed, the applicable evaluation level provisions of the ASHE '
Boiler and Pressurc Vessel Code,Section XI, 1971 Summer Addenda shall be used in the Inservice Inspection of nucicar piping.
Components of the primary systen boundary whose in-service ,cxamina-tion. reveals the absence of flaw indications not in excecs of the '
allowable indication standards of this code are acceptabic for continued service. Plant operation with components which have ,
in-service examination flaw indication (s) in exenas of the allowchle indication standards of the Code shall be subject to NRC approval.
- a. Components whose in-scrvico examination reveals flaw
~
indication (s)in excess of the allowable indication !
standards of the ASNE Code,Section XI, are unaccep- i table for continued service unicss the following "
requirements are met: -
(i) An analysis and evalu'*. ion of the detected flaw indication (s) shall b. submitted to the' NRC that demonstrace that the component structural integrity ;
justifies continued service. The analysis and evaluation shall follow the procedures outlined in Appendix A, 'Tvaluation of Flaw Indications," I of ASME Code,Section XI.
t (ii) Prior to the resumption of service, the NRC shall review the analysis and evaluation and aither I approve resumption of plant operation with the affected components or require that the component (
be repaired or replaced. ;
- b. For components approved for continued service in accordance uith paragraph a, above, reexamination of the area containing !
the flaw indication (s) shall be conducted during each scheduled ~
successive in-service inspection. An analysia and evaluation ;
shall be submitted to the NRC following each in-service i inspection. The analysis and evaluation shall follow the i proceduras outlined in Appendix A, 'Tvaluation of Flaw i Indications," of ASFE; Code,Section XI, and shall reference l 4
prior analyscs submitted to the NRC to the extent applicable. !
Prior to resumption of service following cach in-service ;
inspection, the NRC shall review the analysis and evaluation ;
and either approve resumption of plant operation with the 1 affected components or require that the component be repaired ;
or replaced. )
r r
- c. Repair or replacement of components, including re-examiaations, f
- shall conform with the requirements of the ASME Code, Section i i
XI. In the case of repairs, flaws shall be either removed !
! or repaired to the extent necessary to meet the allowabic j indication standards specifica in ASIE Code,Section XI.
i j
6 127A-1 l
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- SURVEILLANCE REQUIRDENTS LIMITING CONDITIONS FOR OPDATION 4.11 REACTOR FUEL ASSDtBLY 3.11 REACTOR FUEL ASSMBLY Applicability Applicability The surveillance Requirements De Limiting Conditions for Operation apply to the parameters which ,
associated with the fuel rods apply the fuel rod operating condi-to those parameters which monitor the tions.
l fuel rod operating conditions. ,
Objective _
Objective The Objective of the Surveil-The Objective of the Limiting Condi- lance Requirements is to ,
tions for Operation is to assure the specify the type and frequency performance of the fuel rods. of surveillance to be applied to the fuel rods.
Specificati_ons_
Specifications _ _
A. Average ?Ianar Linear Best A. Average Planar Linear Heat Generation Rate (APillGR)
Generr. tion Rate (AFLHCR)
The APLEGE for each type of During power operation with both fuel as a function of average recirculation pumps operating, the planar exposure shall be APLEGR for each type of fuel as a determined daily during function of average planar exposure reactor operation at 1 15 shall not exceed the applicable rated thermal power.
limiting value shown in Figures i 3.11-1 through 3.11-5. The top :
cutves are applicable for core flow ,
greate.r than or equal to 90% of rated core flow. When core flow is '
less than 90% of rated core flow, the lower curves shall be limiting. For l
greater than 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> operation with one 1 recirculation pump, values from these curves are to be multiplied by 0.84 for 8x8 and 8x8R fuel. If at any time during operation it is detennined by nonnal surveillance that the limiting value for APLHGR is being exceeded, )
action shall be initiated within 15 minutes to restore operation to with-in the prescribed limits. If the APLHGR is not returned 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 correspond-ing action shall continue until re-actor operation is within the pre-scribed limits.
i
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20SA . - - - . - __
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-..wo . ..- - . , . . . - _ . . _
LIMITING CONDITIONS FOR 0PERATION SURVEILLANCE REQUIREMENTS C. Minimum Critical Power Ratio (MCPR) C. Minimum Critical Power Ratio OfCFR)
During power operation MCPR for two MCPR shall be determined daily recirculation loop operation shall during reactor power operation at
> 25% rated themal power and be A 1.35 for 8x8 and P8x8R fuel. If I'110"iD5 any change in power at any time during operation it is level or distribution that would determined by normal surveillance cause operation with a limiting that the limiting value for MCPR is control rod pattern as described being exceeded, action shall be in the bases for ipecification initiated within 15 minutes to re- 3. .B.5.
store operation to within the prescribed limits. If the steady state MCPR is not returned to with-in the prescribed limits winin 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 witnin the prescribed limits.
For core flows other than rated the MCPR shall be 2 1.35 for 8x8 and P8x8Rfuel ti=es Kg, where Kg is as shown in Figure 3.11-8.
As an alternative ;-thod providing equivalent thernal-hydraulic protec-tion at core fitws other than rated, i the calculated MCPR may be divided by Kg, where Kg is ss shown in Figure 3.11-8.
For one recirculation loop operation, of greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, tne PCPR limits at rated flow are L.01 higher e than de comparable' two-loop values.
?
205B i
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BASES 3.11A Average Planar Linear Beat Ceneration Rate (APLHCR1
(- This specifications assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident l will not exceed the limit specified 14 the 10 CFR $0, Appendix K.
The peak cladding temperature (FCT) following a postulated loss-of-coolant accf dent is primarily a funetton of the averag'e heat generation rate of all the rods of a fuel, assembly at any axial location and is only dependent, secondarily on the rod tc rod power distribution within an assembly, n e peak clad temperature is calculated assuming a MGR for the highest powered rod which is equal to or less than the design' MGR.
This LEGR times 1.02 is used in the heat-up code along with the exposure depewient steady state gap conductance and rod-to-rod local peaking factors. ne limiting,value for
. APLEGR is this LEGR of the highest powered rod divided by l its local peaking factor. l
)
The calculational procedure used to astablish the APLEGR limit l for each fuel type is based on a loss-of-coolant accident analysis.
The emergency core cooling system (ECCS) evaluation models which l are employed to determine the effects of the loss of coolant '
accident (LOCA) in accordance with 10CFR50 and Appendix K are The models are identified as LAMB, discussed in Reference 1.
SCAT, SAFE, REF'AOD, and CHASTE. 'The IAMB Code calculates the short term blowdown response and core flow, which are input into I the SCAT code to calculate blevdown heat transfer coefficients.
The SAFE code is used to deter =ine longer Where term system appropriate,response the and flows from the various ECC systems.
output of SAFE is used in the REF1 BOD code to calculate liquid levels. Tne results of these codes are used in the CHASTE code to calculate fuel clad temperatures and maximum average planar linear heat generation rates (MAPLBGR) for sach fuel type.
The significant plant input parameters are given in Reference 2.
MAPLHGR's for the present fuel types were calculatedThe curvesby in the Figuresabore procedure and are included in Reference ~4. These multipliers 3.3.11-1 through 3.11-Reference 3 by factors E iven in Reference were developed assuming no core spray heat transfer credit in the LOCA analysis.
Reduction f actors for one recirculation loop operation were derived in Reference 5.
l 205C
REFERENCES )
- 1. General Electric BWR Ceneric Reload Fuel Application, NEDE-24011-P.
- 2. Loss of Coolant Accident Analysis Report for Pilgrim Nuclear Power Station, NEDO-21696, August 1977.
- 3. " Supplemental Reload Licensing submittal for Pilgrim Nuclear Power Station Unit 1 Reload 4", NEDO-24224, November 1979.
- 4. " Supplement 1 to Supplemental Reload Licensing Submittal for Pilgrim Nuclear Power Station Unit 1 Reload 4" NEDO-24224-1 March 1980.
- 5. " Pilgrim Nuclear Power Station Single Loop Operation", NEDO 24268,
, June 1980.
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BASES:
l 3.11C MINDCM CRITICAL POVER RATIO (MCFR)
Operatinat Limit MCP ,R .
For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not ,,
decrease below the safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.1. The required operating limit MCPR at steady scste conditions '
in Specification 3.11.C was chosen conservatively at a value higher than MCPR's of past analysis with the objective of establishing an operating limit MCPR which is fuel type and cycle independant.
The difference between the specified Operating Limit MCFR in Specification 3.11C and the Safety Limit MCPR in Specification '
1.1A defines the largest reduction in critical power ratio (CPR) permitted during any anticipated abnormal operating transient.
To ensure that this reduction is not exceeded, the most limiting trancients are analized for each reload and fuel type (8x8 and P8x8R) tr stermine that transient which yields the largest value of A CPR. This value, when added to the Safety Limit MCPR aust be less than the =4n4=m operating limit MCPR's of Specification ,
3.11.C. The result of this evaluation is documented in the "
" Supplemental Reload Licensing Submittal" for the current reload.
Models used in the transient analyses are discussed in Reference 1 (page 205D).
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MCPR LIMITS TOR CORE FIDW5 otter THAN RATED.
The purpose of the Kg factor is to define operating limits at other than rated flow conditions. At less than 100% flow the required MCFR is the product of the operating limit HCFR and the Kg factor. Specifically, the K, factor provides the required thermal margin to protect 31 ainst a flow in-crease transient. The most limiting transient initiated from less than rated flow conditions is the recirculation pump ;
speed up caused by a motor-generator speed control failure. ,
For operation in the automatic flow control mode, the K g factors assure that the operating limit MCPR given in Specification 3.11C will not be violated should the most limiting transient occur at less than rated flow. In the manual flow control mode, the Kg factors assure that the Safety Limit MCPR will not be violated .
for the same postulated transient event.
The K, f actor curves shown in Figure 3.11-8(4) were developed generleally which are applicable to all BWR/2, &*R/3, and %*R/4 i resctors. Tha Kg f actors were derived using the flow control line corresponding to rated thermal power at rated core flow as described in Reference 1 (page 206D) .
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. j l The g factors shown in Figure 3.11-8(2) are conservative for the Pilgr.a Unit 1 operation because the operating limit MCPR given in Specification 3.11C is greater than the original 1.20 ,
I operating limit MCFR used for the generic derivation of Eg.
4.11.C MINIMUM CRITICAL POk'ER RATIO (MCPR) - SURVEII. LANCE REQUIREMENT At core thermal power levels less than or equal to 25%, the reactor will be operating at =4n4="= recirculation pump speed and the moderator void content vill be very small. For all designated control rod patterns which may be amployed at this point, operating plant experience indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place operation in a more conservative mode relative to MCPR. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% thermal power level wit'n minimum recirculation pump epeed. The MCPR margin vill thus be demonstrated such that future MCFR evaluation below this power level v111 be shown to be unnet.essary. The daily re-guirement for calculating MCPR above 25% rated thermal power is .
suf fic.ent since power distribution shif ts are very slow when there have not been .cignificant power or control rod changes.
The requirement for calculating MCPR when a liniting control rod pattern is approached ensures that MCPR v111 be known following a change in power or power shape (regardless of magnitude) that could pit.ce operation at a thermal limit.
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- REFERENCES _
1
- 1. General Electrie But Generic Reload Fuel Application. IEDE-24011-P.
- 2. Letter from J. E. Howard, Boston Edison Company to D, L. Ziemann I USNRC, dated October 31, 1975. . ,
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FIGURE 3.11-1 '
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, htAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE VERS 05 PLANAR AVERAGE EXPOSURE FUEL TY PE BD B 219 L 4
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PLANAR AVERAGE EXPOSURE (MWft)
- For two recirculation loop Reduction factors for one recirculation loop were derived in
" Pilgrim Single-Loop Operation", NEDO 24268, June 1980, t
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F IG U R E 3.11- 2. '
MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE VERS 05 PLANAR AVERAGE EXPOSURE
~ FUEL TYPE BDB219H
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PLANAR AVERAGE EXPOSURE (MWft) l *For two recirculation loop Reduction factors for one recirculation loop werederived in
" Pilgrim Single-Loop Operation", NEIX) 24268, June 1980. ,
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MAXIMUM AVERAGE PLANAR LINEAR HEAT GENEftATION RATE l VERSOS
. PLANAR AVERAGE EXPOSURE f
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PLANAR AVERAGE EXPOSURE (MWft)
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" Pilgrim Single-Loop Operation", NEDO 24268, June 1980.
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- For two recirculation loop Reduction fact?rs for one recirculation loop were derived in "Pilgriin Single-Loop heration, NEDO 24268, June 1980
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NUCLE AR .E NEltGY SUSINESS GROUP e GE NE R AL E LE CT RIC COMP ANY SAN JOSE, CALIFORNI A 95125 GENER AL h ELECTRIC APPLICABLE TO:
" " ' ' ' ' ^ " "" EF 2 ATA And ADDENDA T .1. E . N O 80NED215 5HEET TIT LE PILGRIM NUCLEAR POWER 3 STATION SINGLE-LOOP OPERATION September 1980 DATE NO W. Conect all cops'es of we applicable ISSUE DATE NEDO-24268 _
publication as specsfred below.
REFERENCES INSTRUCTIONS ITEM (SE C T ION. P AG E (CORRECTIONS AND ADDITIONS)
PAR AG R APH LINE) 01 Page 1-1/1-2 Replace with new page 1-1/1-2.
NOTE: Change is indicated by change bar in right-hand ptargin.
PAGE I Of I
- a. .D
- .~_
e e NED0-24268 l
- 1. INTRODUCTION AND
SUMMARY
The current technical specifications for Pilgrim Nuclear Power Station do not allow plant operation beyond a relatively short period of time if an idle recirculation loop cannot be returned to service. Pilgrim Nuclear Power Sta-tion (Facility Operating License 3.E) shall not be operated for a period in excess of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with one recirculation loop out of service.
The capability of operating at reduced power with a single recirculation loop is highly desirable, from a plant availability / outage planning standpoint, in the event maintenance of a recirculation pump or other component renders one loop inoperative. To justify single-loop operation, the safety analyses docu-mented in the Final Safety Evaluation Reports and Reference 1 were reviewed for one-pump operation. Increased uncertainties in the core total flow and TIP readings resulted in an 0.01 incremental increase in the MCPR fuel cladding /
integ-ity safety limit during single-loop operation. This 0.01 increase is reflected in the MCPR operating limit. No other increase in this limit is required as core-wide transients are bounded by the rated power / flow analyses a performed for each cycle, and the recirculation flow-rate dependent rod block and scram serpoint equations given in the technical specffications are adjusted for one-pump operation. The least stable power / flow cond. tion, achieved by tripping both recirculation pumps, is not affected by one-pump operation.
During single-loop operation, the flow control should be in master manual, since control oscillations might occur in the recirculation flow control system under automatic flow control conditions.
The derived MAPLHGR reduction factor is 0.86 for the 8x8 and 8x8R fuel types.
~
The analyses were performed assuming the equalizer valve was closed. The dis-charge valve in the idle recirculation loop is normally closed, but if its closure is prevented, the suction valve in the loop should be closed to prevent the partial loss of Low Pressure Coolant Injection (LPCI) flow through the recirculation pump into the downcomer degrading the intended LPCI performance. ,
1-1/1-?
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