ML20028G807
| ML20028G807 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 08/21/1990 |
| From: | BOSTON EDISON CO. |
| To: | |
| Shared Package | |
| ML20028G806 | List: |
| References | |
| NUDOCS 9009050023 | |
| Download: ML20028G807 (118) | |
Text
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1 l Marked-uo Technical Snecification Paaes - l- i i 4 1,
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l L 1 i s 9009050023 900821 !: PDR ADOCK 05000293 P PDC
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L'- TABLE Oy C0trTDely O ; r./m.
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\ $9s DEFIN1730NS
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b- l 2.0 . SAFETY LIM 1?S m _. l a4. _ _??12 ETl= ? A* , g[1 2.
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'LIMITINC CONDITIONS FOR OPERATION FURVE!LLUtCE REQUIRDG3rf i *
- 4.1 36 3.1 reactor PROTECTION SYSTDI ,
--i 4.3 A1 3.2- F80TECTIVE USTRUNDITATION 4.3 80
-3.3- REACTIv!TT CONTROL ,
.A 80 A. Reaetivity Limitations 81 S. 'Centrol Reds 3 l
C 83 i- C. Scram faserties Times -36 -
'D. Centrol Red Ace mulators 3
! 35-E. Renetivity Amenalias- I . 85 F. Alternate Requirements C SS >
]G.'ScramDist.hargeVolme:
- 95
-A.4 .
34 ' STAND 3Y L2QUD CONTRDL SYSTEM Ak 5e:3a1 Systen Availability A 95'
'O (q ! 3. Operaties with looperable Components 3 96 C 97 fj C. Sodius Pantaborate Soluties
- 9. . Alternate Requir ments ,
97. 4.5 . 103 3.5 CORI AND CONTAINMDff C00LDC SYSTDS A 103 A. ~ Core Spray and LPCI subsystems 106
- 3. , Containment
- Cooling Subsystem 3 C 107
- C.~ EPCI Subsystem 108 D. RCIC Subsystem; 9 5 109~
E. Automatic Depressurisaties.Systen F. Minia n low Pressure Coe11a8 System and .
. Diesel Generator Availability S
i S. '(Deleted)- . 5 - 112 I . R. Maintenance of Filled Discharge Pipe 4.4 123 3.4 FRDERY -SYSTDI 30pNutr A 123 A. Thermal and Pressurisaties Limitations 3 i' 124-
- 3. Coolant Chemistry C 125 C. Coolant Leakage * '
126 Safety and Relief Talves D
. :9. 3 127 '
E. Jet Fmps 127 F F. Jet Pop 71ew Misantshi -tH-P IN ( S.~ Struetural 1stegrity Flyte 'f,.teih L ;.i a . y1 C
, -1Htt~ l D-3.}%;t E., - 137a
- 1. Shock Suppresa, ors (Soubbt.1i) i a..ne.ns 3s.@,
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f, a ., W/2eem ku q?e' . M @ Surveillance Page No. 3.7 CONTAINHENT SYSTEMS ~ 4.7 152 j 4 -) A. Primary Containment A 152 B. S tandby' Gas Treatment System M B 158 < t C. Tecondary Containment C 159 l 3.8 RADIOACTIVE EFFLUENTS 4.8 177-A. Liquid Effluents Concentration' A 177 , B. Radioactive Liquid Ef fluent B. 177
-Instrumentation I;' C. Liquid Radwaste-Treatment C 178 D. Gaseous Effluents Dose Rate D 179 E. Radioactive Gaseous Effluent E 180 Instrumentation '
F. Gaseous Effluent Treatment F 181 G. Main. Condenser' G 182 H. Mechanical Vacuum Pump H 183 ; 3.9 AUXILIARY ELECTRICAL SYSTEM 4.9 194 A. Auxiliary Electrical Equipment A 194 B. Operation with Inoperable Equipment M ~ 19 (p - 3.10 CORE ALTERATIONS 4.10 202 3
^,m A. Refueling Interlocks A 202
) B. Core Monitoring B 202~
d C. S ent Fuel Pool Water Level C 203 - 1). W IWR DNe R D 703 3.11 REACTOR F L ASSEMBLY 4.11 05A 0 7 osm , A. Average Planar Linear Heat A -E s e 2 c Ca. Generation Rate (APLHGR)
- 8. Linear Heat Generation Rate (LHGR) B -EG5A.4p z.os b a C. Minimum Critical Power Ratio (MCPR) C -20;0 e 2 W b q.
1 D. Power / Flow Relationship D -205B-if 2 066 3.12. FIRE PROTECTION 4.12 206 l A. FireDetectionInstrumqndtion SW[ A 206-B. Fire GeperMMHater4 System B 2064.' ; E C. Spray and/or Sprinkler Systems C 206c L D. Halon System D 206d L E. Fire Hose Stations E 206e L F. Tenetre44ee Fire Barriers % F 206e-1 G. Alternate Shutdown Panele G 206e-1 l
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-Revi; ion 118 Amendment No. AdB,84,89,113,/[/, ii
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> Paae No.
- 14. 0 -_ MISCELLANEOUS RADIOACTIVE MATERIALS. SOURCES 206k
' 4.1 : Sealed Source Contamination- 206k:
O , 4 12 4.3:
. Surveillance Requirements-Reports' 206k
.2061 4.4 Records Retention. 2061 5.0 MAJOR DESIGN FEATURES- 206m 5.1 Site Features 206m 5.2 Reactor h 206m 5.3 . Reactor Vessel 206m 5.4 Centainment 206m 5.5- Fuel Storage- 207 5.6 Seismic Design 207-6.0- ADMINISTRATIVE CONTROLS' 208 6.1.
_ Responsibility. ~208 6.2 Organization 208' 6.3 Facility Staff Qualifications -20Ba l-6.4: Training 208a-6.5 Review and Audit 212' 6.6 Reportable Event Action 216
' 6.7 -S;fety Limit Vic1*t4tef M 217 6.8 Procedures 217 - '6.9 Reporting Requirements 214 6.10. Record Retention 224 6.11 Radiation Protection Program -226 6.12 -(Deleted) 21(p 6.13 High Radiation Area 226 6.14 Fire Protection Program- =227
{QparationalObiettives Surveillance 7.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 8.0 229 7.1 Monitoring Program 8.1 229 7.2 Dose - Liquids 8.2 232 7.3 Dose - Noble Gases 8.3 233 7.4 Dose - Iodine-131, Iodine-133, 8.4 234 Radioactive Material in Particulate Form, and Tritium 7.5 Total Dose 8.5 234'
--Rev4t4en-4EF Amendment No. 35,45,88,89,95,/M, iii
e ?. _ f .. - , p l.0- DEFINIT 20NS he succeeding fregulnT1'y used terms are explicitly defined so that ' a uniform interpretation of-the specifications may be achieved. , A. Safety Limit - The safety limits are limits below which the rea-
. sonable maintenance of the cladding and primary systems are as- :
sured. . Exceeding such a limit is cause_for unit shutdown and review by the Nuclear Regulatory Commission before resumption of unit operation. Operation beyond such a limit may not in itself re-sult in serious consequences but it' indicates an' operational de-ficiency subject ,to regulatory review. B. Limiting Safety System setting (LSSS) - The limiting safety system settings are settings on instrumentation which initiate the auto- ! matic protective action at a. level such that the safety limits will l not be exceeded. The region between the safety limit and these
+ settings reprosent margin with normal operation lying below these
', _ l i
settings. De margin _has been established so that with proper op - eration of the instrumentation the safety limits will never be ax. l coeded. j C. Limiting Conditions for Operation (LCOT - The limiting conditions ' for operation-specify the minimum acceptable levels of system per. formance necessary to assure safe startup and operation of the facility. When these conditions are met, the plant can be opera-ted safely and abnormal situations can be safely controlled. l , ! A I y- - D. ' Col'E o p Eggt /U4 LtMiTS E4: OnR3" q Aw cse. ofEmnNG um crs wswr 4. m W- c74 spapi. M w, % uy W ~ Mc"^^<~t~ &> ther fu~ cev % JA fe -cAs ope % & cy&. Dr & M k & M % M ly th M. % eys spu</a. w m .susourn w p
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p.: y-My ke~CM g th # nt W ) t b,f : 1.1 A SAFETY LIMIT yy ru. Jm'2.1- LIMITING SAFETY SYSTEM SETTING x. ( () - 1.1 ' FUEL CLADDING INTEGRITY Applicability:
,. 12.1- FUEL CLADDING INTEGRITY-Applicability. ;
Applies to the interrelated
~
Applier, to tr'p settings of the variables associated with fuel instruments d devices which are thermal behavior, provided to event the reactor system saf ty ?imits from being ' exceeded. Objective: 0bject' e: To establish limits below which- To d ine the level of the pro-X- the integrity of the fuel clad- < ce variables at which a tomatic ding is preserved, p otective action is initiated to revent.the fuel eladding inte-- grity safety limits from being-exceeded.
.Syecification: Specification:
A.- Reactor Pressure >800 psia and .A. . Neutron Flux Scram Core Flow >10% of Rated , The existence of a minimum The lim ting safety system trip critical power ratio (MC ) settings shall be as specified
.less'than 1.07 shall e sti- below:
'tute violation of th fuel Neutron Flux Trip Settings claddings integrity afety 1.
['.^>A[h : l imi t. . A MCPR of .07 is here-
- a. APRM Flux Scram Trip inafter referre to as~the Setting (Run Mode)
Safety Limit PR. B. Core The al Power Limit (Reat- When the Mode Switch is ; D [toIS('rpffsure .5800 psia and/or in the RUN position,
/ CoreAlow 110%) the APRM flux scram trip setting shall be:
/
klen t e reactor pressure is S 5.58W + 62% 2 loop U 5.800 sia or core flow is less , than equal to 10% of rated, 7 l' the s cady state core thermal Where: lg powe shall not exceed 25% of des gn thermal power. S= Setting in percent i' of rated thermal l power (1998 MVt) C. Power Transient The safety limit shall be as- W= Percent of drive sumed to be exceeded when scram flow to produce a is known to have been accomplished rated core flow of by a means other than the expected 69 M lb/hr. scram signal unless analyses demon- " strate that the fuel cladding integrity safety limits defined in Specifications 1.lA and 1.lB were I (qv
)
not exceeded durinE the actual transient.
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Mt M 72
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LIMITING SAFETY SYSTEM SETTlHG1 hl SAFETY L1 HIT D. Whenever the reactor'is in the- -In the event of operation with a: cold shutdown condition with. maximum fraction of limiting pow . irradiated fuel in the reactor: density (MFLPD) greater.than t > vessel, the; water level shall not fraction of rated power (FRP), the. be less than 12 in, above the top. setting shall'be modified a follows: of the normal active fuel' zone.
'F S 1 (0.5BW + 62%) LPD '2 Loop Where,
-FRP = fr ion of-rated thermal power
- 98 MHt)
MFLP - maximum fraction of limiting: power density where the-C limiting power density is 13.4 KW/f t for all 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, it, which case the. actual operating value will be used.
For no combination of-. loop recirculation flow rate and core thermal power shall the APRM flux scram trip setting be allowed to exceed 120% of rated thermal power,
- b. APRM Flux Scram Trio Setting (Refuel or Start and Hol Standby Mode)
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. C . 18.8 The IRM flux scram setting shall be 1120/125 of scale. O Revision 139 _m Amendment No. IE, 42, 72, 105, 129 s
-mmmmmm mm g.g.mii ii.i............. i- 1 ; 1' SAFETY L1 HIT- 2.1 LIMITING SAFETY- SYSTEM,_SillU"' B, APRM Rod Block Trio Settina.
- 1. APRM Rod Block Trio Settdq (Rur;
/ A \ When the mode switt is in the run-
[ Wh H5'u t t
";han F * " *
$ SRB 1 0.5BH 50% 2_Lorp .
. T4-c.A..Jp. Hhere, SRB - Rod block setting in-percent of rated thermal power (1998 MHt.).
H - Percent of drive flow required to produce a rate" core flow of 69 Mlb/hr. In the event of operating with a-maximum. fraction limiting power density-(MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as 9 follows: FRP ~ S 1 (0.58H + 50%) _MFLPD. LLO.on Where, FRP - fraction of rated thermal power MFLPD - maximum fraction of limiting power density where the limiting power density is 13.4 KH/ft for all fuel. The ratio of FRP to MFLPD shall be set. equal to 1.0 unless the attesi operating value is less than the design value of 1.0, in which cu e the actual operating value will be used. O Revision 139 ! Amendment No. 75, 42, 72, 105, 129 %
6 ,
! .s 1,1 SAFETY LIMIT 2.1 LIMITING SAFETY SYSTEM SETTING ._-.
1 4 2.- APRM Rod Block Trio Setting- _f (Refuel and Startuo Modes)- l When the. reactor mode swit ris in l' ' the efuel.or startup p tions, the- DM rod block;tr setting-shall -set at-les than or equali to 131. 6f rated ge er, but always j p, less than ~ the J#RM, flux scram trip,
-V setting in ;4cification 2.1.A.I..t.
C. Reactor ow water level scram
.settj r shall be 1 9 in._on level:
i frunents. D .' - Turbine stop valve-closure scram settinis shall-be 1 10 perc'ent-valve closure. E. Turbine *ontrol valve fast closurt-setting shall be 2 150.psig. control oil _ pressure at
/ acceleration relay.
F. Condenser low vacuum scram setting j
,m shall be 1 23 in. Hg..' vacuum.
. )
G. Main steam isolation scram setting shall be-1 10 percent valve , closure.
=
H.- Main steam isolation on main steam ! line low pressure at inlet-to l turbine. valves.. Pressure setting j shall be 1 880 psig. I l I. Reactor low-low water level l initiation of CSCS systems setting l m shall be'at or above -49 in. Indicated level. l- : u -:
,- 7
'N s Revision 139 Amendment No. 42, 129
W [Jh & s Y p s WOff U "W ~ *. ih2 mesmeer q;%%+L3}Xng - _ .= .~m... ..
= -
7 - g ~ _- _-= - - _ lg -
'30' ?~ :
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- g- -
100 h g 2}$" ' 1 -S ~
-S -
APRM Flow liased Scram = E_#
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=
90
+
" (Normal) *1.2 E j -7 f _
-: g- -
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1 Rod Block 80 - 1 36-A g L"'"" _3 i
* (Normal) *1
- r& ,EE 9 f -
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w _ , . - _: _
* - =.__ _
m , 60 = _- ya* O jwIh. gg -
*1 f /MFLPD greater than FRP. the intercepts e varied by the ratio FRP WTPU me:::===:
[ -- - 1 See Specifications 2.1.A and 2.1.E g
]
~ *2 When in the refuel or startup/ hot standbv modes, the APRP. scra'n shall be set at g = - ~
=
[N n--
" ' - 15% of design power
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E d-- , F " --_ - K g gg {- i _ i 10 i Figure 2.1.1 @
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- Mfety h-f t: f '
F .j q- \; 2.0 SAFETY LIMITS
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(,b 2.1 SAFETY LIMITS F 2.1.1 With the reactor steam dome pressure < 785 psig or core- flow
< 10% of rated. core flow, q THERMAL POWER shall be 125% of RATED THERMAL POWER.
2.1.2 With the reactor steam dome pressure 2 785 psig and core. flow 210% of rated core flow: , E -1.07 ) fer t;;; C
MINIMUM CRITICAL 1 :p recirculetica POWER er 1 [1.08 ) fersi.,RATIO shall be 2{le 1
-cecircul: tier eper4tian p q
- 2.1.3 Reactor vessel water-level shall be > the top of active irradiated fuel.
2.1.4 Reactor steam dome pressure shall be i 1325 psig.
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./ 2.2 SAFETY LIMIT VIOLATION ,
'o- With any Safety Limit not met the following actions 'shall be met:
2.2.1 Within one hour notify the NRC Operations Center in
~
j
- w. accordance with 10CFR50.72. .!
2.2.2 Within two hours: A. Restore compliance with all Safety Limits, and 1 B. Insert all inurtable control rods. . SC&G h3 fh3 2.2.3 The[G =:1::r P1:nttandAVice President - -)
+ . NucTear;r,erai nouai;cr[andthe-t8Ekshallbenotifiedwithin24 h a Y J NSR&C.] 1 2.2.4 A Licensee Event Report shall be prep eu pursuant. to ' ..
l 10CFR50.73. The Licensee Event Repo shall be submitted to ,
/^ the Commission, the W , the andthefC=:r:1M:n:;erfT i
" % clear' Plant- and Vice President - Nuclear 1 within 30 days ) -
of the violation. h] P g]} 2.2.5 Critical operation of the unit shall not be resumed until authorized by the Commissio.; w -- og rueemcawa) ,
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.B_-2.0iSAFETYLIMITS' 'f.
j [%;' BASES _ cQ .
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INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piping 'are the principal barriers to the release of radioactive materials to the environs. Safety Limits are' established to protect the integrity of these barriers-during, normal plant operations and anticipated transients. The fuel ' cladding integrity Safety Limit is set such that no fuel damageLis calculated to occur if the limit is not violated. . Because fuel' damage -is not directly observable, a stepback ,- approach is used-to establish a Safety Limit such that the ' Minimum Critical Power Ratio (MCPR) is not less than the >
. limit specified in Specification 2.1.2g:r [hth GE nd AT C C"=11 - MCPR greater than the specified limit represents a conservative margin relative to the conditions' required to maintain fuel cladding integrity.
The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some t corrosion or use-related cracking may occur during the life 1 of the cladding, fission product migration from this source L is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal y
) stresses which. occur from reactor operation significantly L .V above design conditions.
While fission product migration from. cladding perforation is-just as measurable as that from use-related cracking, the: thermally caused cladding perforations signal a-threshold-beyond which still greater thermal stresses may cause' gross
.rather than incremental cladding deterioration. Therefore,.
the fuel cladding Safety Limit is defined with a-margin- to the conditions which would produce onset of transition. boiling (i.e., MCPR of 1.0) ~These conditions represent a-significant departure from the-condition intended by design for planned operation. The MCPR fuel cladding . integrity. Safety Limit assures that during normal operation and.during anticipated operational occurrences, at least 99.9% of the fuel rods in the core do not experience transition boiling. FUEL CLADDING E critical power correlations are applicable for all INTEGRITY critical power calculations at pressures at or above 785 psig (2.1.1) or core flows at or above 10% of rated flow. For operation at low pressures and low flows another basis is used as follows: (continuec) n U 7 kM Ho, $ l Q g je 'r ^l=/= r ____________.__l--
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hfQt Ly t q ' yi ' BASES (continued) m. , j Since the pressure' drop in the bypass region is essentially:
, G^y . FUEL CLADDING all elevation head, the core pressure drop at low power and Q '
' INTEGRITY (2.1.1) flows will- always be greater than 4.5 psi.: Analyses show.-
that with a bundle flow of 28 x--103 lbs/hr, bundle pressure
. (continued)-
f- - drop is nearly independent of bundle- power and has a-value of 3.5 psi. .Thus, the bundle- flow with. a 4.5 psi driving head will be greater than 28 x 103'lbs/hr. . Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia. indicate ' l that the fuel assembly critical power at this flow.is a approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED-THERMAL POWER. Thus, a THERMAL POWER 1imit.of.25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.y GThe use of the XN-3 correlation is valid for critical power -l calculations at pressures greater than 580-psig and b die I mass fluxes greater than'0.25 x 105 lbs/hr-ft2, Fo H cladding operation et low pressures or low flows, the fue ingcondition{ integrity Safety Limit is established by a li on core THERMAL POWER with the following is: L Provided-that the water level in t vessel.downcomer is I maintained ve the top of t etive fuel. natural circul on is s ficient assure a minimum bundla flow for all- el assemblie have a relatively high power and- I-- pote tially c reach a critical heat flux condition.- For
,v\. the uel d sign, the minimum bundle flow is greater than 30 x 103 lbs hr. For the ANF and GE 8x8 fuel, the ,
minimum bundle f ow is greater than 28 x 103 lbs/hr. .For.all ~ designs, the co lant minimum bundle flow and maximum flow , area is such at the mass flux is always. greater than 0.25 x 105 lbs/hr-f . Full scale critical power tests taken;at l pressures wn to 14.7 psia indicate that'the fuel assembly 1 critical ower at 0.25 x 105 lbs/hr-ft2 is 3.35 MWt or
-greate . At 25% thermal power a bundle power of 3.35 MWt corr ponds to a bundle radial peaking factor of greater than.
3 which is significantly higher than the expected peaking q - actor. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor. pressures below 785 psig is. conservative.] , (continued) O v g
,, _ n
. . - - .. - - . - . _ .~
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' BASES (con'tinuedI j 1[7
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MINIMUM . CRITICAL POWER e fuel cladd'in integrity Safety Limit is set such that no
, fuel damage is ca$c,ulated to occur if the. limit is not ,
RATIO . violated. Since:the parameters which resultiin-fuel damage
- (2.1.2) are not directly. observable during reactor. operation, the J thermal 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 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 procedures ' used to calculate the critical power result .in an uncertainty _- in the value of the critical power. Therefore, the fuel -l cladding integrity Safety Limit is defined astthe.CPR in'the limiting fuel assembly for which more:than;99.9% of the fuel , rods in the core are expected to avoid boiling.-transition 4 considering the power distribution within.the core and all uncertainties. The Safety Limit' MCPR-is determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical-power. The' probability of the occurrence of boiling
-transition is determined using:the approved General Electric .
Critical Power correlations. Details of the fuel cladding integrity Safety Limit calculation are given in Reference 1. 4 Reference 1 includes a tabulation of the uncertainties used. I in the determination of the_ Safety Limit MCPR and of the nominal values of the. parameters used in the Safety Limit- 3 ' MCPR. statistical analysis.g (continued) jf -t ' l L i
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= CRITICAL ,W S[The Safety perating MCPR/ limit Limit MfPR that in the event of assures an anticipatedsufficient 1 -
6 erational occurrence from'the limiting cc,ndition for. (L1 ? o eration,- at least 99.9% of the fuel' rods in the' core wouldi
- RAM)s"7 ~
- (gontin 2 be expected to avoid boiling transition. 'The pargin betweenf .'
cal ulated boiling' transition (MCPR.= 1'.00) and the. Safety _. Limi MCPR is based on a-detailed stati.itical p"ecedur n consi ers the uncertainties in monitoring the_ cort operating
' state. - One specific uncertainty included in the : safety limit-is the neertainty. inherent in the XN-3 critical . power correlat'on. Reference 2 describes the methodoiogy used in- ,
J' determini g the Safety Limit MCPR.- , The XN-3 c tical power correlation is based on a significant t body of pra ical test data, providing a high degree of 9 assurance tha the critical power as' evaluated by the correlation. is within a small percentage of the actual . critical power ing estimated. Aflong as the. core pressurt f and flow are with the ranM validity of the XN-3 correlation - (refer etion B 2Q .1), the assumed reactor conditions used in d ing the Saf ty Limit introduce conservatism into t li ' because ounding high radial power factors and btunding f peaking distributions are used to estimate the number of rods in boiling transition. Still ~urther conservatism is induced by the _ s W tendency of the XN- correlation to overpredict the number of rods in-boiling tran ition. These conservatisms and the~
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inherent accuracy of e XN-3 correlation provide a 'f E f reasonable operation.at the degree Safety imitof a g(urance-that MCPR there would be no during sustaine transition boiling in_the ore. If boiling transition were to occur, there is reason t believe that the-integrity'of' i the fuel would not be comprom ed. Significant test-data' accumulated by the U.S. Nuclear egulatory CommissionL and 4 1 l- private organizations indicate tha the use of a boiling transition limitation to protect aga st cladding failure is a very conservative approach. Much o he data indicates L that BWR fuel can survive for an extende period of time in-an environment of boiling transition.] ( e ,,1 g ( 3: l lL E l w TW **y
1 02.0% . BASES-(continuedt t REACTOR' VESSEL- With fuel in the reactor vessel during' periods when the r n WATER LEVEL' reactor is shutdown, consideration must be given to water I (2.1.3) level requirements due to the effect of decay heat. If the . . water level should drop below the top of the active irradiated fuel'during this period, the ability to remove: decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures!and clad - perforation in the event' that the water level became less than two-thirds of the core height. The Safety Limit-has. been established at the top of the active irradiated' fuel to ' provide a point which can be monitored and also provide adequate margin for effective ctin1%5A') _ REACTOR The Safety Limit for the reactor eam dome pressure ~has been-STEAM DOME selected such that it is at a pres re below which it can be PRESSURE shown that the integrity of the sys m,is not endangered.. (2.1.4)- The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel CodeXTf474-Edition, including:
- - Adt rA thr
- :;h "ir,t:r 1070}P vhich permits a maximum-
^ pressure transient of 110%, 1375 psig, of design pressure tIdWY}g 1250 psig. The Safety Limit of 1325 psig, as measured by the reactor steam dome pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant' system.
- ,o The reactor coolant system it signed to the USAS Nuclear Power Piping Code,' Section B31. {l959 Editicr., f r.cluding q
-R Ad4nd: thrr,qh Aly 1, 1970} for the. reactor recirculation l J 1 pipina, which normits a maximum pressure transient of'ilM el?O %
.P6qg g er SYo2*Fy of design pressures ofl{ 1250 ) piipfor-suction piping and >
Q - [ 1500 ) priggfor discharge piping. The pressure Safety. Limit is seiected to be the lowest transient overpressure p yrsb2.*F allowed by the applicable codes. l REFERENCES 1. " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A ptett :pproved revitip. I 12d_xon Nuclear Cpttgal Power Methodology for Boiling l' Water Reactors,MMTNF&B+(*7T~ Revision 1, November 1983.] 3 A&WW s-e, ce.e. gewnve omrs mer (l Ject- %se U & R O hh> ~1HHO-5-f ^ 4j2BlS9f l-
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Amendment No. 15~ ?j i ,i. ~.?
^ 'W'4 M, J -; _g , '< ; <'
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g YY 3., n ; .. e "I B ASES : [ 1.1 rUEL CLADDING !NTEGRITY_ $ - LL
~
A. Fuel Claddine Inteority Li . at Reactor Pressure 18M isia' ; ano Core Flow 210% of htec , , , W The fuel cladding integrity safety limit is set such that o fuel damage is calculated to ucur if the limit is not v lated... ! Since the parameters which result in fuel damage are n l directly observable during reactor operation the the .ial and ! p' hydraulic conditions resulting in a departure from ucleate
- boiling have been used to mark the beginning of e region
- where, fuel damage could occur. Although it is ecogni:'ed thetia departure from nucleate boiling wouldf.ot necessarily ;
l r result in damage to BWR fuel rods, the cri cat. power at which i boiling transition is calculated te occ as pen adopted lg as a convenient limit. However, the up rtaint.ies in monitoring , the . core operating state and in the ocedure used to calculate the critical power result in an una rtainty in the value of the critical power. Therefort t : fuel cladding integrity safety limit is define 64 the critici power ratio in the fuellimit? ng$* fuel assembJydor w mor han 99.9;. of the rods in - oiling transition considering the core are pfpected to 'oi power d (tribution wit y the core and all' uncertainties. . The 5 fety Limit
- s detert ged using the General Electric g ' Thenfnal Analysie asi , GETAB l U which is a statistical model Y tha contine 11 of he uncertainties in operating. parameters V. and{uMet!ures the u ed to calculate critical.pcwer.
The probability of + e occurrence of boiling transition ' it, determined .usin the General Electric Critical Quality (X) - P. oiling Length (L), GEXL, correlation., ; , The GEXL correl . ion is valid over the range of conditions used in the te s of the data used to develop the correlction. Theseconditj s are: Pr/ssure: 800 to 1400 ps 0.1 to 1.2Sx10{glb/h.-ft 2 Hix flux: niet Subcooling: 0 to 100 Btu /lo i Local Peaking: 1.61 at a corner rod to a 1,47 at an interior rod Axial Peaking: Shape Max / Avg. Uniform 1.0 Outlet Peaked 1.60 Inlet Peaked 1.60 Double Peak 1.46 and 1.35 Cosine 1.39 ' Rod Array 16. 64 Rods in an 8x8 array
- r. 49 Rods in a 7x7 array i l ,
l' H p) Q, Amendment No. 15 L
.s .. _ . _ _ _ - -
s * ' o .
/ .
- / 4
+
. , J /
4 L s ne required 1:put to .. e statistical ::ofel are the uncertainties y listed on Table 5-1, Reference 3, the noninal values of the core paraneters listed in Table 5-2, Ref erence 3, and the relative r assently power distribution sbeve in Tigures 5-1 s.nd $-1A of Keference 3. Tables 5- A and .421. Ref seence 3, show the . R-f actor distributions that are 6put to the statistical nedel vhich is used to establish the saf ety lir.it MCSK. Le R-factor distributions abovn are taken near the betinting of the fuel 27 me. { E ne basis for(4}be and in NEDO 20340 uncertain .iesfer the buis in thethecors parameters uncertainty is are the giDLen correlation is given in NDD-1095E(1). ne power distri tier. is based on a typical 764 assembly core in which the red pattern was arbitrarily ebesen to produce a skeved power distrf.bution having
- n the greatest number of asse=blies at the highest povdir levels. ne
! vorst distribution in Pilgris Nu:le.ar Power Stat 1ptl1Jnit i during any ' uel cycle would net be as' sever' as the di tribution used in
. the analysis.
- 1. Core Thermal Pever ti J: (Reacter pressure 800 pair or Oore T1ov
<10: of Rated) .
The use of the GD'). correlat . n is valid for the critical '~ power esiculation at pressures v B00 psig or cor e flows less than 10% et ated. Theref .ey the fuel claddir.g integrity saf ety 11=it is stablished , et er means. This is done by establishing a '
- ting e itio of core thermal power operation with the f olleving aris.
Since the pressure drop in the bypass region is essentially all elevation head which is 4.56 psi the core pressure drop at low power and a'.1 flows vill ays be greater thr.n 4.56 psi.
. Analyses sh9v that with a .lov of 28604 lbs/hr bundle flow, bundle pressure drop is arly independent of bundle power and nus, the bund has a value driving neadof vill3.5 bepsi / eater than OSx.10ge lbs/hrflev vith a 4.56 psi irrespective of total core flov independent of bur.dle power for the range of bundle powers of oncern. Tull scale. ATI.AS test data taken
- at pressures from . 7 psia to 800 psia indicate that the fuel asse=bly critical power at this flow is approximately 3.35 Wt. Vith the design eaking f actors ,the 3.35 Wt bundle power cor-responds to a re thernal power of are than 50%. Theref ore a - core thermal over li=it cf 25% f cr re. actor pressures below B00 psia, or ce* flov less than 10% is conservative.
= s O '
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g Amendmen: No. 42 n
! ' ./.:: f.: , :;. .- m -- . 2 , .. L + k . f l=1 -s / [. = C. ?over Transient _ Plant saf ety analyses have shown that the scrass caused by ex-ceeding any safety sett6 g vill assure that the Safety Limit o Specification 1.1A or 1.13 vill not be exceeded. Scram times ' - are checked periodically to assure the hsertien times are adequate. The thermal power transient resulting when a scra is accomplish.' other than by the expected scram signal K.g., rbine scram from neutron ilur following c.losures of the main vever, stop valves) does not ns:25=arily cause fue.1 dar. age. e assumed for this specifiestion a Saiety Limit violation vil vben a serac is only accomplished by means of a bapkup feature of the plant design. The concept of not approacyng a Saf ety L1:1: provided scrs= signals are operable is su orted by the , extensive plant safety analysis. . as a sequence ThecomputerprevidedwithPilgrhUnitifhesequenceinwhich annunciation program Wich will 6dicat events such as scra=p APF2f
- rip initia on, pressure scram initiation, etc. occu r. Thi progt so indicates when the s cran setpoint is' cit.ared. . is vi,11 provide inf ersa*. ion on hov long a ser " cen11 tion cristsAd thus provide some peasure of the ener added curing a
- 4dsient.
D. Reactor Va r Level (Shutdo conditioni 1
//
During p ..ods when the ea tor is shutdown, consideration must also be iven to vate- ev water requirements due to the effect ete level should drop halov the of decay heat. If top of tL actyedel i d ring this tina, the ability to cool the core i tiduced. . d s reduction in core cooling capability could lead to elevated cladding temperatures and clad perfora ton. sufficiently should the water level be The core can be cool Establishment of the s. f ety reduced to two-thir f the core height.ove the top of the fuel provides adequate 11=1: at 11 inches vill be continuously nenitored. cargin. This le References _ Data, 1. General Correlatio El[esric Ther=al Analysis Basis (GETA3):and Design A 3'.7, Syst s Department, Nove=ber 1973 (hto-10958). 2 .' Proces Computer Perf o: ance Evaluation Accuracy, June, 1974 General Eleefric Cc=pany I'a7. Systems Department, (h- 20340).
- 3. eneral Electrte Boiling Vater Reactor Generic Reload y al App 11:stie, NEDE- 24 011-?.
i Anen hen: No. 42 ,, L 4
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BASES: ' LIMITING SAFETY SYSTE71 SETTINGS RELATED TO FUCL CLADDIN3 IN7EGRITY . 2.1 FUEL CLADDING INTEGRITY g # :
.he abnormal operational transients applicable to operation of the NPS 1
'(J Unit have been analyaed throughout the spectrum of planned operat' .g i conditions up to the thermal pcwer condition of 1998 MWt. The .alyses were based upon plant operation in accordance with the operati , map .
.given in Figure 3.7-1 of the FSAR. In addition,1998 MW: is Je licensed i maximum power level of PNPS 1, and this represents the maxi .:m steady-state power which shall not knowingly be exceeded.
Conservatism is incorporated in the transient analyses n estimating th'e controlling factors such as void reactivity coef' cient, control i rod scram worth, scram celay time, peaking factors and axial power shapes. These factors are selected conservative 1 with respect to their effect on the applicable transient results as dyermined by the current analysis model. This transient ver many years, has been substantiateo in operation asfa,podeh, conserydtive s evolved
.o01 for evaluating reactor l
dynamic performance. Resul,tf obtained from a General Electric boiling-L water reactor have been cc::: pared with prop etions made by the model. The comparisons and results are sum.ari id in Reference 1. The absolute value o the void react i y coefficient used in the analysis is conservatively ' ,timated to be a 25: greater tnan the nominal maximuth value exppted to occur d in the core lifetime. The scram worthusedhasbeenderatedtopfeq valent to approximately 80 of the total scram k The scram delay time and rate of rod in /rtionorthallowegby of the control rods. are conservatively set th analyses Q equal to the 1 ngest delay ind slo :est insertion rate acceptable by , V Technical Spe ifications The ef ect of scram wortn, scram delay time and rod insertio ate, all conservatively applied, are of greatest l significance ivQt e arly porti n of the negative reactivity insertion. The rapid insertion of negati 4 reactivity is assured by the time i requirements for 10t and 30: nsertion. By the time the rods are 60 L
. inserted, apprcximately fou dollars of negative reactivity have been
, inserted which strongly tu ns the transient, ud accomplishes the l desired effect. The tim for 50t and 90" insertion are given to assure , l proper completion of the expected performance in the earlier p0rtien ' j of the transient, and +6 establish the ultimate fully shutdown steady state conditio . l This choice of usi conservative values of controlling parameters anc initiating transi answers tnan woud presult s at the cesignexpected by using power level, produces values more of control pessini5tic paramete and analyzing a nigher power levels. Steady-state peration without forced recirculation will not be permitted, except curi ; startup testing. M n ~
-Amendac e c4 {
2.1 RASES
I
- In summary:
- 1. The abnormal operational t:wsients were analyzed r level of 1998 Wt. t
- 11. The lic nsed maximum power leve & 998 Wt.
iii. alyses of tra cat employ adequately conservative values of the ontrolli or parameters. 4 iv. The analytical procedures now used result in a more logical answer than the alprnative method of assuming a higher starting power in conjunctiorfwith the expected values for the parameters. The bas individual set points are discussed below: Neutron Flux Scram Trip Settings APM The average power range monitoring (APM) system, which is cali-brated using heat balance data taken during steady-state coaditions, h reads in percent of design power (1998 ffWt). Because fission cham- i bers provide the basic input signals, the APM system responds directly to average neutron flux. During transients, the instan-taneous rate of heat transfer from t.be fuel (reactor thermal power) 9(med
,36 ) is less than the instantaneous neutron flux due to the time constant -m of the fuel. Therefore, during abnormal operational transients, the (U) 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 feel damage. Therefore, the use of flow referenced scram trip provides even additional margin. O 4he-Htw-btned strim piv&d ee Figun 2.1.1 1.1,ei.e4 ee r;c-ked r \ l
.
- 4 nn_te,p t}
An increase in the ApM scram setting would decrease the margin pre-sent before the -the(' fuel cladding integrity safety limit is reached. l The APM scram setting was determined by an analysis of margins re-i quired to provide a reasonable range fr? maneuvering during opera-L tion. Reducing this operating margin w 4uld increase the frequency ! of spurious scrams, which have an adve- e effect on reactor safety because of the resulting thermal stresses. Thus, the APM setting was selected because it provides adquate margin for the fuel clad-ding integrity safety limit yet allows operating margin that reduces the possibility of unnecessary scrams. O .nds.ntw g
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- 3 - %y Ce!EofERA1M& LVNTS WJdCF The scraa tr setting must be adjusted to ensure that laGR i transient ye is not incrasood for any combination of amminua l.
fraction of 11 hating power density (HFLPD)and reactor core thermal power. K scraa setting is adjusted in accordanaa with the iormula ing;ni incir Wgl
*.1. A.Evben the NFLPD is
( graatar than the fraction of rated power 7EP).j .
~
yeas of the limiting treasients show that no screa adjustasat ] is required to assure MCPR greater than the Safe'.7 Limit McFR when the transione is initiated from EPJL above the operating Limit McFA. ' l For operation in the startup anda while the reactor is at low pressure, the AFIN ocrea setting of 15 percent of rated power providas edeguate .tharmal sargia between the setpoint sad the safety 11ad,t,15 percast et rated. The margia la adequate to ( accomandata anticipated mammuvers associated with power plant , startup. Effects of increasing pressure at sero or low void
- content are minor, sold water from sources availabla during startup is not asch soldar than that already in the system, temperatura coefficients are maan, and sentrol rod patterms are constrained to be uniform by operating procedures backad up by the rod worth miniminar.
Worth of individual rods is very low in a uniform red pattern. Thus, of all possible sources of reactivity taput, unifera control I rod withdrawal is the most probable case of significant power rise. Because the flum distribution associated with miform red withdrawals does not involve high local peaks, and haemune several rods must be anved to change povar by a significant percentage af rated power, the rata of power rise is very slow. Gemara11y tha
, heat fluz is in the near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the screa level, the rate of power rise is no more than five percent of rated pesar per mianta, and the APRM system would be more than adequate to
' assure a scram before power could ascoed the safety limit. The 151 APRM screa remains active stil the mode switch is placed in the 121 position. TLis switch occurs when reactor prosaura is greater than 880 peig.
The analysis to support operation at various power sad flow re-lationships has considered operation with either one or two re-circulatica pumps. M h the IRM system consists of a chambers, 4 in each of the reactor protection system logic chanaals, the IBM is a 5-decada instrument which covers the range of povar laval b-" - m . 1
=_ . . . . . ,
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rme . - , Q between that covered by the 3RN and the AFIM. The 5 decadas are covered by the IBM by means of a range switch and the 5 decades are broken down into 10 ranges, nach being one-half of a decade
. I is sise. The IBM scras setting of 120/115 of full scale is active i is each range of the IRE , For azample, if the Lastrummat were an range 1, the scraa settias would be a 120/115 et full sca2a for that range likewise, if the Lastrument were on range 5, the seras ,would be 120/125 of full scale sa that range. Thus, as the ZIN is reased '
- sp to acconnodata the sacrease la power level, the scram setting is !
also ranged up. The most significant sources of reactivity abange , durias the power increase are due ta nontral nd wf tMr*=el. Fer i in-sequence control rod withdraum1, the rate of change of povar is slow enough due to the physical limitation of withdrawing control rods that hast flus is in equilibrium with the neutros. flux, and an , IIH scram would result in a reactor shutdova well before any safety 4 limit is escoeded. In order to ensure that the IIM provided adequate protection agatast *
. the single rod withdrawal arror, a range of rod withdrauni accidents ;
was analysed. This analysis imeluded starting the accident at various ! power levels. The most severe case involves an initial condition in ' i which the reactor is just suberitical and the IBM system is not yet i on scale. This condition entses at quarter rod density. Additinaal conservatism was takan in this analysis by assuming that the IIM ' . channal closest to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak oore power limited ' O to one percent of rated power, thus main **== ICFR above the Safety Limit MCF1. Based on the above analysis, the IRM providas protection against local control rod withdrawal envors and continuous withdrawal Q control rods in sequence and prom des backup protection for the AFEM.]
- a. = = = = = g + p .7 ; ;
7 Reactor power level any be varied by moving control rods or by varying the recirculation flow rata. The a?IM rysten providas a sontrol rod block to prevent red withdrawal beyond a given point at constant recir_culation flow rata, and thus to protect against the condition of ; a MCFA ese the Safety Limit WCPR. 'Xnis rod block sec point, unich is automati y varied with recirculation loop flow rata, prevents an increase in the reactor power Inval to arcassive valbes due to control rod withdrawal. De flow variable trip setting provides substantial martin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the safety limit increases as the flow decreases for the specified t: rip setting versus flow relationships therefore, the worst case MCFR which could occur during steady-state operation is at 107I of rated thermal power because of Ge APRM rod block. trip h@ to tm y /2
.-Amondaant E 17 f .e.-.. .m-,.- e----,,_.w..w.. . _ _ _ _ , _ _ _ , . - . _ . _ , _ _ , _ _ _ _ . _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . , _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ , _ _ _ , _
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t = < feetting. The actual power distribution la the core is established by specified control rod sequemees and is monitored coatinuously by , t the in-core LFRN systas. As with the AFEM scram trip setting, the AFRM rod block trip setting is adjusted downward if the maximum fraction of limiting power density escoeds the fraction of rated power, thus preserving the AFRM rod block safety margia.f - fReactor/Watoih Lee Lovel -
= ,
l y De set potat for low inval scram is above the bottom of the separats. '
- skirt. This level has been used La transient analyses dealiar with 1
coolant inventory decrease. Se results show that scram at this laval adequately protects the fuel and the pressure barrier, because McFR i remains well above the safety limit WCFR in all ca,ses, and oystsat 3 pressure does not reach the safety valve settings. De scram setting is approziantely 25 is belev the moraal operating range and is thus , - adequate to avoid spurious scrama. . Turbine Stoo Valve Closure. "a WS *::"- f , The turbine stop valve closure scram anticipatas the pressure, neutron i fluz and hast flus increase that could result from rapid closure of l the turbine stop valves. With a scran trip setting of 1 10 percen't gf valve closure from full open, the resultant increase in surface hast ( flux is limited such that MCPR r h am above the safety limit WCFR even during the worst case transient that assumas the turbine l fs bypass is closed. Turbine Centrol valve Tast closurs Seesa-0 2 ?: W - p De turbine control valve fast closure scram anticipates the pressure, asutron flux, and heat flux increase that could rasult from fast closure of the turbine control valves due to load rejection escoading the capability of the bypass valvow. De reactor protection system initiates a scram when fast camours of the control valves is initiated by the acceleration relay. Bis setting and the fact that control valve closure time is approximately twice as Imag as that for the semp valves means that resulting transients, while sta11ar, are less severe than for stop valve closure. McFR remains above the safety limit McFR. Main Condenser lav vacuum "x__ Tr5 St, tina-f To protect the main condenser against overpressure, a loss of condenser vacma initiates automatic closure of the turbine stop valves and turbine
.. bypass valves. To anticipate the transient and automatic scram resulting -
from the closure of the turbine stop valves, low condenaar vacuum 1 initiates a scran. The low vacuum scram set point is relected to initiate (a scram before the closure of the turbine stop valves is initiated. j O , rNec k c 3(p - k
-Ame dment-wm-*ef 13 9 l
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r
}-4 + b aelet' ea 5ces= p The low pressure isolation of the main steen.1ines at 880 psig was provided to protect against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage is taken of *he scram feature that occurs when the main steam line 1 solation valves are closed, to provide for reactor shutdown so t 'that high power operation at low reactor pressure does not occur; thus providing protection for the fuel claddin integrity mit. Operation of the reactor at pressures ower than h" pos i requires that th- reactor mode switch be in the STARTUP.
), where protection of the fuel cladding integrity safety
' - limit is provided by the IRF, high neutron flux scram.amaAne(15 scram < i Etans, the acabisatian of nata stema aina lov Saeasure* isolation and i isolation valve closure scram assures the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit.- In addition, the isolation. valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation 1 valve closure. With the scrams set at 10 percent of valve '
closure neutron flux does not increasey CReactor Low-towlater Leve1~$eFP6 int for~Attisatf 6ref Core 5tandby Coo 11no 5ystem The core standby cooling subsystems are designed to p ' de Q-- sufficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident and to linikfsel clad
' temperature, to assure that core geometryMins intact and ' '~
.to limit any , clad _ metal-wat,er reactigado less than 15. To ace'omplish-their iiitended functipn, the capacity of each Core .
Standb ooling Syst componefit was established based on the. { reacto low water le 1-scram set point. To lower the set point t of the ow water lev ' scram would increase the capacity require-ment for ia W of the SCS components. Thus, the reactor vessel low water level sera was set low enough to permit margin for operation, yet will t be set lower because of CSCS capacity requirements. ) The design of CSCS components to meet the above guidelines t wasdependent,vponthreepreviouslysetparameters: the maximum break sige., low water level scram set point and the CSCS initiatiop' set point. To lower the set point for initiation of the CSCS'may lead to a decrease in effective core cooling. To raisn'the CSCS initiation set point would be in a safe direction, but'it would reduce the margin established to prevent actuation tif the CSCS during normal operation or during normally expected transients.- r di Amendment E . 1 @
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J i . . 4 I ea mie==.
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nsient and accid 1 Semonstrate that these
-(* tInerfeiHill in tr anrstas for the -
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- 7~" l 1.2 SAFETY LIMIT .
2 LIMITING SAFETY SYSTEM SETT!$
) 1.2 REACTOR COOLANT SYSTEM INTEGRITY 2.2 REACTOR COOL ANT SYSTEM INT DITY Applicability: Applicability:
~~ '
., . _ . __ _. Applies to limits-on reactor - - - -
,.,) plies to trip se) ings of the'in coolant system pressure. struments and de ices which are pro-vided to preve the reactor system-safety limi rom being exceeded.
Objective: Objective.. To establish a limit below which To f e the level of the process i the integrity of the reactor cool. varf bles at which automatic pro-ant system is not threatened due t,ective action is initiated to to an overpressure cond' itinT'N
/ prevent the pressure safety limit from being exceeded.
}
Specification: f Specification: The reactor ssel.done pressur, The limiting safety system settings shall not e eed1325psigat/ny shall be as specified below: time when ) radiated fuel y pre-sent in t e reactor vess . f) Limiting Safety () , , Protective Action System Settino l A. Scram on Re- $1085 psig actor vessel high pressure B. Relief / Safety Nominal setpoint valve settings will be selected be-tween 1095 and 1115
/ psig. All valves
/ shall be set at this nominal setpoint .+,
L /
\
l 11 psi. C. Safety valve 1240 psig i 13 psi settings 9 1
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r['; : gjSES : -: SATETY 1.1MIT tt REACTNi C0011,KT SYSTEM 1NTECAITY a e o . . -
. 1 1.2 The'rcactor coolant system. integrity is an important barrier:1 esserr the j
prevention'of-uncon_ttelled release of fission products. It ! r
, tial .that; the : integrity. c f this system be -protected by esta' ishing'a " ,
pressurc . limit: to be observed for all operating 'cenditions and= vhen.- -)
- cver there.is irrediated. fuel in.the reactor vessel. 1i The pressure safety limit,of 1325 psig as measured by he vessel _ steam'
- spacc jressure ineicator-ir, equivalent to 1375 psig the levest ele. ~,- Ll
- vation of the: reactor coolant system.; The 1375 ps ..value_is derived ,
W ci; from t.he desitn pressuren of the reactor: pressure essel'(1250 psigf ~ at $750p) and coolant syst<.m piping (suction' pip np 1148 psig at y 5620 p; discharge pipings: L1241 psig at 5620p). The pressure safety
~
l
- 11mit vas; chosen as the lover of the pressure transients perstitted by ~
the applienble det.ign codes t AS?2 Loiler a pressure vessel Code, l' Seetion III-for the press ' ves 1_and 1F SI-L31.1 Code for'the reac-- _ ter coolant: system ' pip g. The ASI: to er and presourc Vessel Code 3 y N 1 ' permits pressure.tra tents up to 1 1*4 'er design pressure (110%'X- ' 1250 = 1375 psig), nd the USASI cc ) perv.its pressure transients up
.J
- K to 20% over_the . sign pressure (1* >/. X 1148 = 1378 psig; 120*/. X 1241' = ,;
0 1489 psig). , The Station S fety Analysis ett on 14.5.1). states that the turbine- g fp :t.rtp from hi.h power witho' byp s is ,the most severe abnormal opera-
.tional tranatent resuitietnrdire ly in a renetor coolant'systen pres . ,
W V'- -
'sure increase. 'The r ves el pressure code limit of 1375 psig, J given in Sub eti .2 of th Safety Analysis' Report, is vell above- '
x~ the peak pressure produced b' the vorst; overpressure transient above. , Thus, the pressure safety 1 4.it is well above_the peak pressure that-can result from reasonably expected overpressure transients. L
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i Thevalvasizinganalysisc:nsideredfcur,107.capasityrelicf/safetyvalds ! a.nd two 81, capacity safety valves. Ibese are sited and set pressures are os. tablished in accordance with the following three requirements of Section III i ; of the AS:2 Coder j ! t
'Og '
- 1. Tr.e lowest safety valve n'ust be set to open at or below vessel design pressure and the highest safety valve be set at or below 109% of design i
pressure. - ,
- 2. The valves must limit the reactor' pressure to no more than 3,10$ of design ,
pressure. 3 Protection systems directly related to the valve sizing transient unst not be credited with action (i.e., an indirect screa must be assumed). I A win stsam lin ischtien with flux cerem har been relected to be used as l the safety valve sizing transient since this transient results in the. highest peak vessel pressure of any trsnsient when analysed with an indirect scram. ' The original ySAR analysis concluded that the peak pressure transient with indirect scram would be caused by a loss of condenser vacuum (tudine trip with failure of the bypass valves to open). However, later observations have shown that the long lengths of steam lines to the turbine buffer the faster i stop valve closure isolation and thereby reduce the peak pressure caused by this transient to a value below that produced by a me,in steam line isolation with flux scram. Item 3 above indicates that no credit be taken for the primary scusa signal generated by closure of the r.ain steam isohtion valves. Two otner scram initiation sicnnis would be generated, one due to high neutron flux and one j due to high reactor pressure. . Thus item 3 vill be satisfied by assuming a : ( screa due to high neutron flux. ; Relieving capscity of kOT, (4 relief / safety valves) results in a peak pressure during the transient conditions used in the safety valve sizing analysis which is well below the pressure safety limit.' N- . E w do p. IW - s m O .
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' he relief / safety valve settings satisfy the Code requirements that the .
lowest safety valve set point be at or below the vessel design pressure range to prevent unnecessary cycling caused by minor trr.nstants. Tne I results of postulated trenstents where inherent relief / safety valve 1
. actuation is required are given inY! " '- -
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*
- of theg safety Analysis Report. l
- (Frpox.fian. (W-A @ ltnok . -
l the expected case, wheresti tp 'am sign {tl generatedly c.losure of.
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the va n gr f eaintion valvat dom fact initiate a scram, the
.i ' . *
' set point and capacity of the re7tef/;;hh vatves is suG1crenWemove .
8 j '. . enough energy frcm the reactor to prevent the safety valves from lifting l
.u (Figure 14.0-7 of TSAR).J ,
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S'
. LIMITING CONDITIO% FOR OPERATION O ' SURVEILLANCE REOUIREMENis REACTOR PROTECTION SYSTEM i 3.1 REACTOR PROTECTION SYSTEM i l ,
Aeolicability: Aeolicability: Applies to the instrumentation Applies to the surveillance of the l
- and associated devices which instrumentation and associated i initiate a reactor scram.- devices which initiate reactor !
scram, j l Objective: Obiettive: 4 To assure the operability of the To specify the type and frequency reactor protection system. of surveillance to be applied to ' the protection instrumentation. Soecification: Specification: ! The setpoints, minimum number of A. Instrumentation systems shall lA. trip systems, and minimum number be fui.:tionally tested and ; of instrument channels that must calibrated as indicated in be operable'for each position of Tables 4.1.1 and 4.1.2 the reactor mode switch shall be retnectively, as given in Table 3.1.1. The system response times from the B. Verify the maximum fraction of, opening of the sensor contact up limiting power density is less e n to and socluding the opening of than or equal.to the fraction : the trip actuator contacts shall of rated power once within 12 (U)- not exceed 50 milli-seconds, hours after thermal power is . greater than or equal to 25% of rated thermal power and B. The maximum fraction of limiting power density (MFLPD) shall be every 24 hours thereafter. less than'or equal to the QQ WKQ] fraction of rated power (FRP) when thermal power is greater than or equal to 25% of rated thermal power. gj g gg
- 1. IfMFLPDisgreaterthanFRPf gg@A/(; a (M IT3 T6fDRT adjust the APRM high flux '
setpoints to the relationships s, _ n ~.n.n 4 OAM c,rF,4"ft1TATFT go Ajg & t/.s 6 hours. If the required actions and 2. associated completion times of Specification 3.1.B.1, f jdA M " AA G d above cannot be met, reduce . g ggg[ thermal power to less than 5% of ratcd thctm:l pcwer within 4 hours. v) .
. IO"! $ ! ^^ I N -k AmendmentNo.42,[/b 20 L
TABLE 3.7.1 REACTOR PROTECTIO % .,rSTEM (SCRAM) INSTR!)MENTATI0tl REQtHREMENT Minimus Number Modes in Which function Operable Inst. Trip Function Trip Level Setting Muit_Be Operable Action (I) Channels per Refuel (7) Startup/ Hot Run Trio (l) System Standby _ 1 Mode Switch in Shutdown X X X A 1 Manual Scram X X X A IRM 3 High Flux 1120/125 of full scale X X (5) A 3 Inoperative X X (5) A APRM A or B
^
2 High Flux s G " (15) (17) (17) X 2 Inoperative (13) X X(9) X A or B g 2 High flux (15%) 115% of Design Power X X (16) A or B ' 2 High Reactor Pressure 11085 psig .X(10) X X A 2 High Drywell Pressure 12.5 p:;ig X(8) X(8) X A 2 Reactor Low Hator Level 19 Ir.. Indicated Level X X X A tM Vefume.3 2 High Waterftevel in Scram Discharge W 139 Gallons X(2) X X A
< ES >
2 -TToir.4 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 (18) X X X(18) A or C 4 Main Steam Line Isolation Valve Closure 1101 Valve Closure X(3)(6) X(3)(6) X(6) A or C s=u ret 2 Turb 2150 psig Control Oil Clos [ureCont[Valvefast Pressure at Acceleration Relay X(4) X(4) X(4) A or D 4 Turbine Stop Valve Closure) 1101 Valve. Closure X(4) X(4) X(4) A ar D ub_ n )N
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y - S6. 92, ///, e
- 27
^5dmentNo.twm o
NOTES FOR TABU. 3.1.1 (CONT'D) !
- 10. Not required to be operabic when the reactor pressure vessel head is not i bolted to the vessel.
ld 11. Deleted
- 12. Deleted .
- 13. An APRM will be considered inoperable if there are less than 2 LPRM inputs per level or there is less than 50% of the normal complement of ,
LPRM's to an APRM. : 14, E d+iS/hr. i: pereetaf drive Trip livri fin icyviiru sciiing in pivvvie in percent e 7:ted of n :ign powercere '!oe=ef
'1000 . p50 "\
- 15. pk:Sectic=>2.1.A.I.C ,
- 16. The APRM (15%) high flux scram is bypassed when in the run mode.
- 17. The APRM flow biased high flux scram is bypassed when in the refuel or startup/ hot standby modes. '
i
- 18. Within 24 hours prior to the planned start of hydrogen injection with the reactor power at greater than 20% rated power, the normal full power
- radiation background level snd associated trip setpoints may be changed based on a calculated value of the radiation level expected during the injection of hydrogen. The background radiation level and associated trip setpoints may be adjusted based on either calculations or ,
measurements of actual radiation levels resulting from hydrogen
- O injection. The background radiation level shall be determined and V associated trip setpoints shall be set within 24 hours of re-establishing normal radiation levels after completion of hydrogen injection and prior to withNawing control rods at reactor power levels below 20% rated power.
b ffKY 'N A .&NW
, y w tL, ape-.oraArmc, wurs psner, a m n-nuamspus r t f
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, 3.1 BASES (Cont'd) 4.1 BASES (Cont'd) +
m been provided to allow for To facilitate the implemen-(V) tation of this technique, Figure bypassing u... o,.f..u one such channel,
. - - . - ' ' ' =
)
4.1.1 is provided to indicate an j^ur the -!M, ^^^", h! ? rea c tar ~p appropriate trend in test pr+swrc, reecter ics w;t:r c 1 interval. The procedure is as 1 v:1, "SIV cicivir, sir.ireter p follows: Joad-rijic iion, tur birrt ttopp ! w4+e-e4oiw w w6 ici; cf I Like sensors are pooled into
- 1. i
{cedenser VGtm ei Glicvind 'np one group for the purpose of ' 6pe444tatier, 2.1 er.d 2.2 '~ data acquisition. i Hth Orwdl fNooW r Instrumentation (pressure 2. The factor M is the exposure switches) for the drywell are hours and is equal to the provided to detect a loss of number of sensors in a - coolant accident and initiate the group, n, times the elapsed core standby cooling equipment, time T (M . nT). A high drywell pressure scram is provided at the same setting as 3. The accumulated number of the core cooling systems (CSCS) unsafe failures is plotted initiation to minimize the energy as an ordinate against M as i which must be accommodated during an abscissa on Figure 4.1.1. a loss of coolant accident and to prevent return to criticality. 4. After a trend is This instrumentation is a backup estabitshed, the appropriate , to the reactor vessel water level monthly test interval to '
/c\ inttrumentation. satisfy the goal will be the 'd A she b RM ILON test interval to the left of High radiation levels in the main the plotted points.
l steam line tunnel above that due l to the normal nitrogen and oxygen 5. A test interval of one month i radioactivity'is an indication of will be used initially until leaking fuel. A scram is a trend is established. Initiated whenever such radiation level exceeds seven times normal Group (B) devices utilize an background. The purpose of this analog sensor followed by an , scram is to reduce the source of amplifier and a bi-stable trip such radiation to the extent circuit. The sensor and necessary to prevent excessive amplifier are active components I turbine contamination. Discharge and a failure is almost always of excessive amounts of accompanied by an alarm and an radioactivity to the site environs indication of the source of is prevented by the air ejector trouble. In the event of off-gas monitors which cause an failure, repair or substitution isolation of the main condenser can start immediately. An off-gas line. "as-is" failure is one that b A /4r1,Aab1 " sticks" mid-scale and is not A reactor mode switch is provided capable of going either up or which actuates or bypasses-the down in response to an varinne erram rimettnns not nf.ltmite i n,nn t . Thje tune I appropriate to the particular of failure for analog devices is plant operating status. Ref, a rare occurrence and is l l
'tp) ' Section 7.2.3.7 FSAR. detectable by an operator who observes that one signal does not track the other three. For purpose of analysis, 36 AmendmentNo.[(
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3.1 BASES (Cont'd) 4.1 BASES (Cent'd)
/NLM ScNvm i (n) The manual scram function is active it is assumed that.this rare 'v in all modes, thus providing for a failure will be detected within '
manual means of rapidly inserting two hours. control rods during all modes of reactor operation. The bi-stable trip circuit which is a part of the Group ) IRMsysemandJPRM(15%) scram) (B) devices can sustain unsafe provl ec-tMgainst failures which are revealed excessive powe ev s and short only on test. Therefore, it is reactor periods in the tug and necessary to test them l Qntermediate Dower ranges.f- periodically. _ h #.04eeAaw shrenar # W W The control rod drive scram system A study was conducted of the is designed so that all of the instrumentation channels water which is discharged from included in the Group _(B) ; the reactor by a scram can be devices to calculate their ' accommodated in the discharge " unsafe" failure rates. The ploing. The two scram discharge analog devices (sensors and volumes accommodate in excess of 39 ampitfiers) are predicted to gallons of water each and are at have an unsafe failure rate of i' the low points of the scram less than 20 X 10-' failure / discharge piping. No credit was hour. The bl-stable trip taken for t5ese volumes in the circuits are predicted to have f, design of the discharge piping as an unsafe failure rate of less ; than 2 X 10-' failures / hour. 1 concerns the amount of water which (T must be acccemodated during a scram. Considering the two hour ) V monitoring interval for the i' During normal operation the scram analog devices as assumed s discharge volume system is empty; above, and a weekly test however, should it fill with interval for the bi-stable trip 1 water, the water discharged to circuits, the design ; the piping could not be reliability goal of 0.99999 is i accommodated, which would result attained with ample margin. j in slow scram times or partial l control rod insertion, To preclude The bl-stable devices are l this occurrence, redundant and monitored during plant l diverse level detection devices operation to record their l in the scram discharge instrument failure b* story and establish a volumes have been provided which test interval using the curve will alarm when water level reaches of Figure 4.1.1. There are 1 4.5 gallons, initiate a control rod numerous identical bi-stable block at 18 gallons, and scram the devices used throughout the reactor when the water level plant's instrumentation reaches 39 gallons. As indicated system. Therefore, significant above, there is sufficient volume data on the failure rates for in the piping to accommodate the the bl-stable devices should be i scram without impairment of the accumulated rapidly. ; scram times or amount of insertion I of the control rods. This function The frequency of calibration of shuts the reactor down while the APRM Flow Biasing Network l n sufficient volume remains to has been established as each I accommodate the discharged water ( and precludes the situation in which a scram would be required but not be able Amendment No My 37
7 i 3.1 [ASES (Cont'd) 4.1 BASES (Cont'd) to perform its function adequately.. refueling outage. The flow i A source range monitor (SRM) blasing network is functionally !
^ tested at least once per month '
system is also provided to supply and, in addition, cross k' additional neutron level calibration checks of the flow information during start-up but has input to the flow blasing network no scram functions. Ref. Section can be made during the functional ; 7.5.4 FSAR. The APRM's cover the test by direct meter reading.
" Refuel" and "Startup/ Hot Standby" There are several instruments modes with the APRM 15% scram,~and the power range with the flow which must be calibrated and it i will take several days to perform
/ biased rod block and scram. The the calibration of the entire' IRM's provide additional protection network. While the calibration in the " Refuel" and "Startup/ Hot
# Standby" modes. Thus, the IRM and APRM 15% scram are required in the is being performed, a zero flow signal will be sent to half of the APRM's resulting in a half
/ '
- Refuel" and "Startup/ Hot Standby" scram and rod block condition. i modes. In the power range the APRM Thus, if the calibration were
& system provides the required performed during operation, flux t
/ protection. Ref. Section 7.5.7 shaping would not be possible.
\g FSAR. Thus, the IRM system is not Based on experience at other required in the "Run" mode. generating stations, drift of instruments, such as those in the The high reactor pressure, high Flow Blasing Network, is not drywell pressure, reactor low water significant and therefor 2, to level and scram discharge volume avoid spurious scrams, a l high level scrams are required for calibration frequency of each
, Startup/ Hot Standby and Run modes refueling outage is establ1shed. of plant operation. They are, i therefore, required to be Group (C) devices are active only
- operational for these modes of during a given portion of the reactor operation. .
operational cycle. For example, the IRM is active during startup The requirement to have the scram and inactive during full-power i functions, as indicated in Table operation. Thus, the only test 3.1.1, operable in the Refuel mode that is meaningful is the one is to assure that shifting to the performed just prior to shutdown Refuel mode during reactor power or startup; 1.e., the tests that operation does not diminish the are performed just prior to use need for the reactor protection of the instrument. system. Group (D) devices, while similar The turbine condenser low vacuum in description to those in Group scram is only required durinQ power (B), are different in use because operation and must be bypassed to they (the analog transmitter / trip start up the unit. Below 305 psig unit Cevices) provide alarms, trips e turbine first stage pressure or scr u functions. An availability l . (45% of rated), the scram analysis is detailed in NEDO-216i7A (12/78). l Surveillance frequencies for the SDV tvstem indennentation 15 detailed in Amendment Number 65. NRC concurrence with this surveillance pro-
-Revision-103P 38 pwx m. R l
-3.)
t BASES (Cont'd) 4.1 BASES (Cont'd) , signal due to turbine stop valve gram is contained in the Safety closure is bypassed because flux Evaluation Report and its associated G and pressure scram are adequate to Technical Evaluation Report ,.h ! protect the reactor. (TER-C-5506-66) dated 11/10/82. The requ.lrement that the IRM's be Calibration frequency of tht, i inserted in the core when the instrument channel is divided APRM's read 2.5 Indicated on the into two groups. These are as scale assures that there is proper follows: - overlap in the neutron monitoring i systems and thus, that adequate 1. Passive type Indicating coverage is provided for all ranges of reactor operation. devices that can be compared with like units on a continuous basis. The provision of an APRM scram at 115% design power in the " Refuel" 2. Vacuum tube or se.91 conductor and "Startup/ Hot Standby" modes devices and detectors that > and the backup IRM scram at drift or lose sensitivity, i 1120/125 of full scale assures that there is proper overlap in Experience with passive type the neutron monitoring systems instruments in generating and, thus, that adequate coverage stations and substations is provided for all ranges of indicates that the specified
-( .
reactor operation. calibrations are adequate. For those devices which employ L. - h $ L A g fro., 9 N amplifiers, etc., drift
'q *F specifications call for drift to f] j vv .. A dc +. ~ ~c e, A ANf be less than 0.4%/ month; 1.e., in V the period of a month a drift of r<d Hve.k / 9 sc # j 4 @ /ect .4% would occur and thus d - - 4 // M Aro execeds providing for adequate margin.
For the APRM system, drift of
/wP , /L f+c.3emig /4 AITW electronic apparatus is not the ecd 4/cc 4 M 4 x<a / 3 ,, only consideration in determining a calibration frequency. Change In power distribution and loss of
/ P ,4 , s // c A c. o c 4 ~. /* chamber sensitivity dictate a ed
/ or b calibration every seven days. '
,(ce<4 g. su/c N.,e 7 / /L./1,/cr f~.,/<< t6 % %assures plant Calibration on this frequency operation at or
%y g< u / p /c d. s d R d below thermal limits.
/4< Al'O( fer d .e ye. /er
/-/,- e- p / /j3a c 4 L ea c
A compari an of Tables 4.1.1 and 4.1.2 indicates that two afr4 r*/> > f.wJ , dul M / A_ instrument channels have not been cd p 44 ,4/ W r re< d dee# ^*j included in the latter Table. These are: mode switch in e med /co % c,f g42 shutdown and manual scram. All [. , /1c %/ /ber ,.2 _ -v 4c. e of the devices or sensors associated with these scram (' 4 4duefA f-m /ed. c - /k functions are simple on-off recc /or ew A v/ ge /. switches and, hence, calibration during operation is not applicable, i.e., the switch is either on or off.
~ ~ ' "
AmendmentNo.Jh 39
, . , - , , . - _ - . - - -,.,,n,.,,,.,-,,-.,,,,-.,,-,-,a.n ... , _ , , _ . . . , , , , -n. ~ , , . , , , --- ,,,-.e. - , . . - ,
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}(.1 legg, (cent'd) l> 3 The Namiansa Praction of limiting Power Density (MrLPD) '
shall be checked once per day to determine if the s APRK scram requires adjustaast. This will neraally be deae.by checking the 1 Pat readings. Caly a small aanber of sentrol rede are moved daily and thus the MFLPD ta not aspected to change signifteastly and
- thus a daily shock of the MFLPD is adequate.
~
, The sensitivity of LPRM datestore decreases with surposure to nantess flun at a elev and apprestantaly epastaat rate. This is esapensated fet ta the APRM syetam by calibrating every three days estag heat balance data and by calibrating indivf. dual LPRN's every 1000 affective in11 power hours usias TIP troverse data.
t 4 I 1
-(L AmendmentNo.[[ .,
[ 4
[ G/ h
\q'0 '
i' m FNPS TABLE 3.2.C-2 CCWTROL POD BLOCK INSTPtM NTATION SETPOINTS Trip Function Trio Setroint APRM Upscale - 2: !---:'"--t:-- ? ' ? O ( 1 ) ( ~2. APRM Inoperatiee Not Applicable APPM Downscale 1 2.5 Indicated on Scale Rod Block Monitor "M' 'N "
' ' ,T (
(Flow Biased) fihi- I Rod Bicek Monitor Inoperative Not Applicable Rod Block Monitor Downscale 1 5/125 of Full Scale IRM Downscale 2. 5/125 of Full Scale IRM Detector not in Not Applicable Startup Position IRM Upscale i 108/125 of Full Scale IRM Inoperative Not Applicable SRM Detecter not in Not Appi, cable Startup Position SRM Downscale 1 3 counts /second SRM Upscale 5 i 10 counts /second SRM Inoperative Not Applicable Scram Discharge Instre 2nt Volume i 18 gallons Water Level - High Scram Discharge Instrument Volgene - Not Applicable Scram Trip Bypassed Recirculation Flow Converte'. - Upscale i 120/125 of Full Scale Recirculation Flow Comerter - Inoperative Not Applicable Recirculation Flow Converter - Comparator i 10% Flow Deviation for > 8 N. Rated Powr, and ch . < 15% Flow Deveation for < 83% Rated Power Mism4* o y ('u M h. W N Dn .";'"I pd.. aCe.~.$,,, n ~@@,,G,, SAT ,,6-nn WE, sis, 5_ L &M ITS
, ,._ ,~ , , ,,_
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N A , Y A~.ne .nt no.' az, 1:e. M % & M L K% ~ y h r3 % ep - Ssa _ _ _ _ i _
pra A 3.2 EASES (Cont'd) The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not decrease to the Safety Limit MCPR. The trip logic for this function is I out of n: e.g., any L trip on one of six APRM's, eight IRM's, or four SRM's will result in a rod block. The minimum instrument channel seguirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements for the RBM x Jy be reduced by one for maintenance, testing, or calibration. This time period is only 3% of the operatin; time in a month and does not significantly increase the risk of pretenting an inadvertent control rod withdrawal. The LDBi red bimt functier i; ' low bia;;j in the-rgn nd; nd pr ' o
/ m i r ,1 . -
. cs m i, m m g ,,,, m 1 y: ;it p.I, -flee QIn the startup and refuel modes the APRM rod block function i setcoin below the APRM flux scram triphs spec'ficd in-Specifia; . ^-c
-?1 n ?. Mc An'" prov d s gros: core prott' tion; '.e.. ' "'ts t'. g
+ros; core ;ower incre,'st- f ro withdrewal of control rods " t" nom' 'a wiiis a ai Lequence. The trips are set 50 ihsi HCFR i:, maintained f 9r4 ate-thn-Re Safety Lid i HCFR. (
The RBH rod block function provides local protection of the core, for a single rod withdrawal error from a limiting control rod pattern. The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less O than a factor of 10 above the indicated level. A downstale indication on an APRM or IRM is an indication thc instrument has feiled or the instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and thus, control rod motion is prevented. The downstale trips are as shown in Table 3.2.C-2. The flow comparator and scram discharge volume high level components have only one logic channel and are not required for safety. The refueling interlocks also operate one logic channel, and a e required for safety only when the mode switch is in the refueling position, for effective emergency core cooling for small pipe breaks, the HPCI system must function since reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function is provided as a backup to the O _ .m AmendmentNo.IE,42,110,/Q 71
..:. . g.....
s 4
-31 2 BASES (Cont'd)
HPCl!in the event the HPCI:does.not operate. The arrangement of the
~
-tripptrig-cont *, cts is such as to provide this function when necessary and-minimite smrious operation. The trip settings given in the O. specification ars adequate to assure the above criteria are met. .The specification preserves the effectiveness of the system during periods i ,
of. maintenance, testing or calibration, and also minimizes thy isk of .
- inadvertent operation 1.e., only one instrument channel out of service..
.Four rtdiation monitors are provided which initiate the Reactor Building Isolation and Control System and optration of the standbyegas treatment system. The.instrutent channels monitor the radiation from the refuellig area ventilation exhaust ducts.
,, four instrument channels are arranged in a 1 out of 2 twice trip logic.
Trip settings of < 100 cr/br for the monitors in the refueling. area ventilation exhaust ducts are based upon initiating normal ventilation isolation and standby gas treatment system operation so that none-of the-activity released during the refueling accident leaves the Reactor Building.via the normal ventilation path but rather all the activity is proces' sed by the standby gas treatment system. Flow integrators are used to record the integrated flow of liquid from the drywell. sumps. The_ alarm' unit-in each integrator is set to annunciate before the values specified in Specification 3.6.C are exceeded. A system whereby the time interval to fill a known volume will be utilized to provide a back-up to the flow integrators. An air
' sampling system is also provided to detect leakage'inside the primary containment.
- q.
e. _5_' 72 AmendmentNo.k
LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS
. 3.6.CJCoolant; yMCont'd) 4.6 _ ,
power' operation is permissible y during the succeeding seven b g% g g
#hhTAn ed4tif/
-10% W lil5pg. A4 g/
A bo c et an rderly - U #N shutdown shall be-initiated and M t I! p<A. Tkg.:t;; A < the reactor shall be in a Cold ( qgggg# a Shutdown Condition within 24 hours. O% 1 3 % ,- D. Safety and Relief Valves D. Safety and Relief Valves 1.'Ouring reactor power operating 1. At least one safety valve'and two conditic s and prior to reactor relief / safety valves shall be-startup from a Cold Condition, checked or replaced with bench or whenever reactor coolant checked valves once per operating pressure is greater than 104 cycle. All valves will be tested psig and temperature greater every two cycles. than 340*F, both safety valves ~ and the safety modes of all k set point shasi-be as s he safety valves 44-46 S reliefvalvesshallbeoperable.) K ' Specification 2.2.
/-
2.IfSpecification3p.1isnot
.A F met, an orceriy snutdown shell 2. At least one of the relief / safety-T A h be initiated and the reactor valves shall-be disassembled and coolant pressure shall be below inspected each refueling outage./-
104 psig within 24 hours. Note: Technical Specifications-3.6.D.2 3. Whenever the safety relief valves T 3.6,0.5 apply only when two are required to be operable, the Stage Target Rock SRVs are discharge pipe temperature of each installed. safety relief valve shall be logged daily.
- 3. If the temperature of any safety relief discharge pipe exceeds 4. Instrumentation shall be 212*F during normal reactor calibrated'and checked as power operation for a peric' 3f indicated.in Table 4.2.F.
greater than 24 hours, an " engineerirg evaluation shall be 5. Notwithstanding the above, as a y performed justifying continued operation for the corresponding minimum, safety relief valves that have been in service shall be increases. tested in the as-found condition tem @Y during both Cycle 6 anri Cycle 7.
- 4. Any safety relief valve whose discharge pipe temperature exceeds'212*F for 24 hours or more shall be removed at the next cold shutdown of 72 hours O or more, tested in the as-found condition, and recalibrated as 7 necessary prior to reinstallation. Power operation shall not continee' beyond 90 days AmendmentNo.52,55,J( I 126 l
u... - ,, . . , , LIMITING CONDITIONS FOR OPERATION- SURVElllAg r REQUIREMENTS- l 3.6.D Safety Relief Valvu frnn't) m E. Jet Pumps
-i j 3a- ,
-from the initial discovery of Whenever there is recirculation discharge pipe temperatures in flow with the reactor in the i . excess of 212*F for more than startup or run modes.. jet pump.
24 hours without prior NRC operability.shall be. checked daily approval of the engineering by verifying that the following evaluation delineated in-3.6.D.3. conditions do not occur-simultaneously.
- 5. The limiting conditions of operation for the instrumentation 1. The two recirculation loops have'a that monitors tall pipe tempera- flow imbalance of 15% or more wher, ture are-given in Table 3.2.F. the pumps are operated at the sar.ie speed.
E. Jet Pumps-
- 2. The indicated value of core flow
- 1. Whenever the reactor is in the rate varies from the value derived startup or run modes, all jet from loop flow measurements by more pumps shall be operable.. If it than 10%.
is determined that a jet pump is inoperable, an orderly = shutdown 3. The diffuser to lower plenum
~ shall_be initiated and the differential pressure reading on an.
reactor shall be in a Cold individual jet pump varies fr Shutdown Condition within established jet pump [ F'~~~ 24 hours. characterictics by more than'10%. . F. Jet Pump Flow Mismatch F. Jet Pump Flow Mismatch
- 1. Whenever both r: m culation Recirculation pump speeds shall be pumps are in operation, pump checked and logged at least once speeds shall be maintained within per day.
10% of each 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. If Specification 3.6.F.1 is exceeded imediate corrective action shall be taken. If recirculation pump speed mismatch is not corrected within 30 minutes, an orderly shutdowh-shall be initiated and the reactor shall be in the Cold Shutdown condition within 24 hours unless the recirculation pump speed 'mistaatch is brought within limits soontr.
uo AmendmentNo.7J,y, 127
. LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS _
4 ,
.f x ;rs, 3.6.G Structural' Integrity 4.6.G Structural Integrity
- D .
q [ ')/ 1. The structural-integrity of Inservice inspection of
~ x- / -- the primary system boundary components shall be' performed '
'shall.be maintained'at the in accordance with the PNPS-level required by the'ASME' Inservice Inspection Program.
Boller and Pressure-Vessel The results obtained from Code,'Section XI " Rules for Inservice Inspection o (uclear compliance with this program .' Power Plant Component will be evaluated at the Articles IWA, IWB, IWC, IND and completion of.each ten year-interval. The conclusions of-INF and mandatory appendices as required-by 10CFR50, Section This evaluation will-be reviewed with the NRC.
. 50.55a(g), except where specific relief has been granted.by the NRC pursuant ice 10CFR50, Section th.55a(g)(6)(1).
.. s. o&*r.J HW
. p(Eh q; O}
4 4
\_ / ,
.s Amendment No. 19, 93 127A
a . ..
. g g,t',l- ,
1.6.D and b.6.n- , . .; . C:> safety and Relief vains .
/
O r- . discussed in Substation'k.k.6 of, ths Final Safety Analysis, Report,. ' 4 of the malsar systes ssure relief system is intanded to- .
- --h==
- . potect
* - the safety valve s <, ludn,^ $essurization on .
inAirect screain is the essened event of besanse.ASME Boiler. ~ Pressure vessel Code,'Sectica III,. requires-
. .that prot'ection systens etly related to the valve sizing treasient 'j anst not be credited with act 0: M m 3.jsive relieving
. onpacity. A total of 4 relief / safety ~velves,end a saan^v M-3rwided by the Ama4 _
W h W.
..
- fLP . ,
Erperience in safety valve operation shows that a testing of et least 4 (~5o5 of the safety valves per refueling entage is adeguate to, deteet- ;
-fafh'res or deterioration. The tolersace vaine of +15 is in'accor-dance with Section III of ths A3ME 3ciler"and Faessure Yessel Code.
An ane. Lysis bas.been performed which shows that with all safety-valves set 15 higher, the reactor ecolant pressure safety limit of 1373 Psig is not emneeded. . Therelief/safetyvalveshavetwofuma'tions;i.e.,powerreliefor self-4ctuated by high pressure. power relief is a solenoid actuated . ,: funct tea (Automatic Pressure Relief) in which external instrumen. - g 1- ' tation signals 'of coincident high dryvell pressure and low-lot veter . 1 ' level initiate the. valves to open. This function is discussed in Specification 3 5 3.. In addition, the valves can be operated.
,. h
. Pilgrim's'asperience with 2 stage safety /. relief velves has demonstra that =4a4== leakage exists when the: tailpipe temperature is 215* F enheit.
Therefore. a reporting requirement' triggered 1.- a temperacure..gf.2% F is conservative, and assures timely reporting be". ore leakage reaches signifi . . L
~'
east proportions'. , L . .. , [. . l gyg p . O - l Amendment No. k'. .
, (; . . .
, . . - .4. L . . - _ . . . . . . . l; , .. .. . . - - . . . . . . _ , .. . . . . . . . . . . . . . . . . . . . . o
,n- - - - -
4
-( -
Q . p * '.i d t'^ . G
.. 9
- 4
~ - en Wj 4
0 . A. II)
-* - w..t:2.-4 y ,
146 i e -_c.. . . _ _ - _-_-________-_-_-_-__.______________._-____.------.___--__-__-_--_a-.-_-=_u
-v x 4 -3 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
.g 4,11 REACTOR FUEL ASSEMBLY 3.11 REACTOR FUEL' ASSEMBLY Applicability: Appli cabil i t() -
- The Limiting Conditions.for The surveillance requirements Operation associated with feel apply to the parameters whict)( y rods' apply to-those parameters the fuel rod' operating which monitor the fuel rod conditions, operating conditions.
Objective: Objective: The Objective of the Limiting The Objective of the Conditions for Operation is to Surveillance Requirements-is to assure the performance of the specify.the type and frequency-fuel rods, of-surveillance.to be applied-to the fuel rods. Specifications: Specification @_ A. Average Planar Linear Heat A. Average Planar Linear Heat-Generation Rate (APLHGR) Generation Rate (APLHGR)- During power operation with both The APLHGR for each type of recirculation pumps operating, fuel as a function of average = the APLHGR for each type of fuel planar exposure shall be as a function of average planar determined daily during reactor exposure shall not exceed e operation at 1 5%~ 2 rated thermal applicable limiting value[fshown p-)wer. In t iguresl_.11-1 throu V I-3.11-7. The ur are applicpbfe ore flow greater than of-e to 90% of rated core fl ,ow When core flow is ( m M A t;(f C n .e less ttfan 90% of rated core flow, / t ower curves shall be C M. . limitinghWIf at any time during- t operation it is determined by normal surveillance that the limiting value for APLHGR is > being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits. If the APLHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown N condition within 36 hours. Surveillance and corresponding
. action shall continue until reactor operation is within the prescribed limits.
m p _ . 2 _ 2 - .. 2na ,, 205A P T No. )f >h ) 1h>
u J ; LIMITING CONDITION'S FOR OPERATION SUR'/EILLANCE REOUIREMENTS
.;F;
'^ B. Linear-Heat Generation' Rate (LHGR) 8. Linear Heat Generation Rate (LHGR)
,p t During reactor power operation; The LHGR as a function of core' the ' % ear heat gea*ratian rete - height shall be checked daily-
- -(LHCR) of :ny rod in .ny fuel-y \ during reactor operation at; es::-61y :t any Oxi;l lot:ti0n 125% ratG thermal power.
-;h:11notcxcccd13.4kwiftfor{3
-el'- fue! .c - l-If-at any time during operation -tk, LHGR aAcM #
.it is determined by norral surveillance that:the limiting -6XA 0A d N M..
value-for LHGR is being exceeded, - 4g@ action shall be initiated within 15 minutes to restore operation to within the prescribed limits, O MY l a If the LHGR is not returned to within the prescribed limits within two (2) hours, the' reactor shall be brought to the Cold
. Shutdown condition within 36 hours. Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.
--> W M kk C %N {' " . 34. --
O n - -/g#4,#m# ze
-hv44nn _ inn p -gc53 _p
e a. . }*- . Mwif Nr. count' rip gysugi(1,Ang REgt m pgg _p g- ] g Vt W N W Q LO Q 911minum' Critical Power Ratio (MCPR) C. Minimum Critical Power Ratio (MCPR) MCPR 'shall be determined daily during t6.. During
* -" "*" e- power opesatiopCPR
- ;; ieg .id t e pei.iii..l
--GLi- ?_!! _ e.?;Ilf at any time during operation it is determined by notinal surveillante that the limiting value for MCPR is being exceeded, action shall be[f reactor power operation at > 25% rated thermal power and following any change in power level or distribution that would cause operation with a limiting control rod pattern as described in shall be f.nitiated within 15 minutes the bases for Specification 3.3.8.5.
to rest'sre operation to within the
. prescribed limits. If the steady 2. The value of Tin Specification state MCPR is not returned to with- 3.11.C.2. shall be equal to 1.0 ,
in the prescribed limits within two unless determined from the resul 'I (2) hours, trie reactor shall be of surveillance testing of Spe fi- , I brought to the Cold Shutdown cation 4.3.C as follows: condition within 36 hours. S urveil-lance and corresponding action shall a) Yis defin6d
- as continue until reactor operation is -
within the prescribed limits. m [ve . 3 1 For core flows other than rated the y S1.U5- 7 l limita shall be the limits identifi above times Kg where Kg is as sh in ) The ave se scram time to the ] Tigure 3.11-3 artion position is deter- j mine as - follows : ' alte ative meth e providinS n - l cqui alent therna 7 Ng7g Sl As ct co t er draulic protection than rated, the cal- y y,.i=1
]
culated ))CPR may be divided by Kf, where n l Kg is ay'shown in TiRure 3.11-8. I Ng
- 2. operating limit MCPR values as a unction of fare given in Table 3.11-1 l
where: n = nunber of surveillance
- whereTis given by specification tests performed to date in the ,
4.11. C. 2. . cycle . pLws a 1 h W kt6 Gw u %% M Y. (
-so5.t O( /mendumt-No. 54 f
---++-+p-= = *. s . w. . . . , _ _ _ _ , , , ,, ,, , _ e
~~)
~ ~ ' ^ ~ ' ~
yi? [",/; . 3, q t*. ..
}y LIMITING CONDITIONS FOR'0PERATION. ~ SURVEILLANCE RE0VIREMENTS.
I J y; [ hi=numberof:activecontrolrods measured-in the ith: 1
- p .
. surveillance test.
A-
" ' i = average.s' cram time to
-t e 301 :
insertion position of 11 rods. ' 6 measured in the Ith- )
?
-survelliance test. 1 h[ -
c) The adjusted anal; is mean-
- scram time ( T ) 5 calculated as f lows:a N, 1/2 .
I t g -- v4 .65 'n * ' 4,
- EN. !i
,1 - 1 , l Where:
.l p = me of~the! distribution for
- av rage scram insertion: time I t the 30% position;0.945 sec.
Ni= otal number of active control rod measured in t specification-4.3.C. [ t
.. standard-deviation of the
( distribution for average scram , insertion: time: to the 301 ).. i posttion,' 0.064~sec. ) > D. Power / Flow-Relationship Durina D. Power / Flow Relationship Durina Power Operation Power Operation The power / flow. relationship shall
~
.Compilance with the power / flow not exceed the limiting values e relationship in Section 3.11.0-
" :he 'n Fige 3.1 L FIf at shall be~ determined daily: -
any time during power operation during reactor operation. It is-determined by normal surveillance that the limiting valueforthe' power $ flow relationship is being exceeded, ' 4 action shall be initiated within ^ ,
'15 minutes.to restore operation a -t h b - '
to within the prescribed limits. If the. power / flow relationship is O d, . not returned to within the ; prescribed limits within two (2) ~ hours, the reactor shall be brought to the Cold Shutdown - condition within 36 hours. - , r Surveillance and corresponding action shall continue until O- reactor operation is within the prescribed limits.
^ - - " " " - "
-C= :ted Febi u=i .y G W No. h , h , -2058-4 p lo.---1986f go g
4 s M , TABLE 3.11 ,, ' OPERATING LIMIT MCPR VALUES A._ MCPR Operating Limit-from Beginning of Cycle (BOCl to'BOC.+ 513 MWD /ST.
.EjhBR/IP8x8R For all values of-t-- .36
-_ B . - MCPR Operating Limit from BOC + 7,513 MHD/ST to En of C/cle.
t P8x8R/BP8x8R 11 0 1.39 0.0 < t 10.1 1.40 0.1 <t5 0.2 1.40
'O.2 <t 1 0.3 1.41 0.3 < t1 0.4 1.41 0.4 <11 0.5 1.42 0.5'< t1 0.6 1.42-
--; 0.6 < t1 0.7 1.43 0.7 <ti-0. 1.43 O- 0'- I
.9. 1.44
- 0. 9 ' <
l.0 1.44 M it s Dau h ton 108 p 905B y
, _ - ~ . ~~= * *
]'i g .>
-t
+n , 3 j s .'
<M :.. . ,
q
' . [n .3dy e
9 Aversee Planar Linear test Ceneration Rate (AFWC n his specification (9 j I. following the_ postulated design basis'1oss-of-coolantassur I newill set asteed the limit _ epecified LaK1{Q seeident Appendi a E.
. losspeak cladding j f of-coolant accident tempirature YE). following a.Ipostulated
- best ged ration rate'of tily a function of the average i antal loc 1 t od power d sand i rods enden af a fuel, sembly at any -
1
.2... econdarily L Pt@emperatureiscalculated within anThesesembly.
the ro O peak slad povered rod which ta egual using a 12CR for the highest . This LUCR times 1.02 is uand in lass than the design' LICE.
"- = _asAalon rod-to-rod local peaking factors. , The limitt.n 1Qss AFLBCR local peakf is this LBCR of the highest powered rod divided b a: factor.
_i , hETheca5ulational p.ocedure used to establish the A each fuel type is based on a less-of-coolant accident era e= ergency core cooling system (ECCS) evaluation analysia. models whi h . m:1 dent c ' yed. to determine the affects of the loss of co dis 6ussed in fereses'1. 3C47, SATI, and shcrt ters blevdown The models are identified-as MMB,'The case and em the SCAT code to cale at flow, which are~imput~1sto The SATI code is used t eat transfer coeffielents C and ficvs from the vario deterr.ine longer term system responsa. CC rystems.
'h output of SATI is used in the levels. De results of these cadas D code Where appropriate, the to-calculate liquid i . to calculate fuel elad te=peratures and aax ed in~the CEASTE cade 11.ne.ar haat generation rates 0(APLEGR)erage
. planarfu for each The significant plant input parameters an'd the MAPLHCR' t the present fuel types'. calculated by the' abov s for included'in, Reference,2 ~ .
e procedure are 1P 1b + .. W w +, C d., e a.
.ALxQ A A W u tL top 4~e e u n > m s @ A 6.9.A.4
- 3. it. e h Ha,x WLm (u e Ph T w. y d, $ - ewr de h ki# M * *.
q Nn%. jw .L tL cW'g W k.*^-T & A
, 3 i % w% . u co m W W 4e cLir M- t.A., LHGE M g Cha#p;s e paa a w e a M-w m./f.7% ft R tl .
mp FI. . . . . , . .
$ f
'. A'-
+ s O e
- D 9
9 e e I l
% CL 1h12.h 4 WhAJ.A.1 We ----i
...s O t
6 4 4 e 4 0 6 1 _3_
. 1_. . _ _ . . . ~. - 4vs us, N
O
, esse w.
4
.i..
g,,i
..- . p '
z..
' . d ii. '
RErBD(CES '* 9-3v., .}; } ' ' , . s .{
.. , General-Electr p -- - 1.
' SWR Generic aLFueF1typTfcation, NEDE-24011-P..
E { 2. '.tess df Cools ent Analysis Report for Pilgria Nuclear Powar. A '< Sta t 0-21696,' August 1977 as amended.
.j l 4
'e
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6
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( r 1 g'(# i
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t ., b i ( 4 I e . l t .n h o l; j ,l . - N ==.=E :L-3; Q --een-a-Q L ' e
****==*-..4 . . , .
*+ 4 .
~
BASES:L A 4 3.1 RITICAL POWER RA (MCPR) . cx f
~
Ooeratino limit MCPR j G-For any abnormal ~ operating transient analysis d=ti^^ with the initial- l condition of the.reactorW at the steady state operating limit, it is. i required that the resulting MCPR does not decrease.below the Safety Limit HCPR at any time during the transient assuming instrument trip settink given ingdru/n ficatien z.u' s2.1
. 2 tO
. A ,n.t
. r. . G d s .l i
l 1 Tdifference betw(en the'specifiefoperating Limit MCPR in Specification 3. llc and the Safety Limit MCPR in-Specification 1.11 J defines the. largest reduction in critical power ratio (CPR) ted ; T- during any anticipated abnormal operating transient. T sure that this- 1 reduction is'not exceeded, the most' limiting'trans s are analized for t eac. reload uel 3 Nto determine that s ent which yields the l
~ largest ue of ACPR. Thl. value, why ed to the Safety Limit MCPR must b less than the minim m op3Ating e limit MCPR's of Speci ication 3.11.C. The attit of this evaluation is documented in the i "SuppWntal- R21oad-td e ing Submittal" for the current reload. 1 The evaluation of a g en transient begins with the system Oput parameters shown i able 5-4, 5-6 and 5-8 of NEDE-240ll-PC 1, ..
Supplemented b load untaue inputs given in the current Supplemental. ' 1
' Reload Lipa ng Submittal. These values.are input to a GE beha,vter' transient computer program described in NEDO-10802('gpre > . The dy;
, -enstent code used for all pressurization events is described in '
y NEDE-24154-P (Reference 5). The MCPR analysis for pressurization events is done in accordance with the procedures given in Reference 6. H
' i,j l
l W W Y W ek 5 W & h o y W NPR A A tb 6 Op 5 A' , p yx D# A @n u. 4. V. l q % eca pu p tfu 0 Y' / 00'0 U y 1 kWS. [ p p 4. g M ,a.t.a.a w ~
, % m~r g c/v ~+ AMDb M b M "" '
l W No, h, g M [r/;d, 205h ) ilevi5icii IOP -205C=$ l i
.s.. .
, , 1;. . J .', .. ,i
, ',', t *
.T.
3
; s. ., , .
, +..- .
e,
- y. . .. . ,
- v. ;
'd< f
, . sN
^ l
?}
codes: are used to analyse the rod withdrawal error transient,
' The first: code simulates ths. three dimensional BW1 core nuclear
+ cad ~ tharsal-hidraulic characteristica. .Using this code a limi ,
?
control rod pattern is determinad;' the following assumption e o
, included in- this determinations ,.
F.~(1) The core is operating s't full power in the sen free. condition. N
- (2)_ The highest worth control rod is -ass be fully inserted.
'(3). Th analy3 rformed for th st reactive point in-the n (4)' a contro'1 rods ase- d to be the worst possible pattern- ,
thout exceedin armal limits. l(5) bund the cinity of the-highest worth control-rod is-assumed to be op rating- at the maximum allowable linear haat generation rat .
'J.(6). A bundle in-a vicinity of the' hi best worth control rod is l
fassumed to e operating ,the .minimun allowable -. critical . power ~t ratio. ' ,
. .'he three mensional BWR code then simulates. the core response to ',
,'{ l~the' con el rod ' withdrawal' error. 'The - second code calculates the 'e L Rod 31 ek Monitor response to' the rod . withdrawal error. This code , s tes the Rod Block Honitor under selected failure conditions > ( ) f or the core rod.use (calculated by the 3-dimensional SWR dacion code) for the control rod 'vithdrawal.- , 9 M' i
--- 4 Y 9 e
~ . . e _
5 ) e . -
... .- .~ ..: .. _
+.
p The analysis of the rod with rawa d l arror for Pilgria Unit 1 I 1 considers the continuous withdrawal of the mar'== A sunnary vorthf cent
- rod at its maxisma drive speed from the reactor. taristics I
the analytical methods used to deterrine the nuclaar char is gAv,en in section 5.2.1.5 of NzpI-24021-P. MCPR LIMITS FOR (DRE TIDk5 OT1LER TRAN RATED The purpose of the Eg f actor is to define operat g100% limitsflow At less at other than rated flow conditions. ing limit MCPR the required MCPR is the product of the oper or provides t and the Kg iactor. Specifically, the Kg fa t a flow in- ) che required thermal' margin zo protect a sient initiated from crease transtant. The most limiting recirculation p. asp less than rated flow conditions is thspeed control failura. ,
- speed up caused by a motor-ganarato ov control mode, the Kf factors for operation in the automatic MCPR given ta Specification 3.11C assurethatjha,,o vill not be viola'perating should a most limiting transient occur at t
e manual flow control mode, the Eg less raced fio . In
- f acto s assure that thav afsty Limit MC?R will not be viciated .
gd transient event.
~
for e same post ' e shown in Figure 3.11-B(4) vers developed The' factor e applicable to all BWR/2, BWR/3, and Sk'R/4 gana call ' ch actors vare derived using the flow control reacto The Kf lina correspondin to rated thernal power at rated core flow. For the manual f ow control moda, the Kg factors were calculated maximum flow stata (as 11mitad by the pump scoop such that at t and the corresponding core pswaY (along the rated tube set pointina), the limiting bundia's relativa povar was flow control > the MCPR was slightly above the Safety limit. adjusted unt tive bundle power.'the MCPR's vara calculated Using this at differ t points along the rated flowThe control line ratio of the MCFR-correspo ing to different cera flows. calcula ed at a given point of core flow, dividad by the { opera 3 limit MCPR deter"nas the Eg. To operation in the automatic flow control mode, the samawas employed e
.ocedura was established such that the MCPR was equal to the operating limit MCFR at rated power and flow.
~
4mdseth iw. ;2 et05c-5
Kr fact;rsshown in Figure 111 a* are conservative fer the ' Pil Specificatnit'loperationbec(se-theopedrtting.1131tMCPRgivenin
. E t greater M th; ami 1.20 an rating 11mitL l t
(. MCFR.used for the generic derivation.of K 4.11.C . MINIMUN CRITICAL. POWER RATI@(MCPR){-s-5vkWILLCS *E0"!" At core thermal = power levels less than or equal'to 251, the reactor-will l be operating at minimum rectrculation pump speed and the moderator void- a L . content will be very small. For all designated control rod patterns-P twhich may be employed 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. 4;rir,; hiti:1 start ;p t;;t...u ,, .. ,,,.n.,
. mrn 7 -
- ni;;ti= lli bc =E et 25Y tMr :! pcu:r I;ve 4:ctr;;iett= p r; sp:;d. The E"" ;;;;rgia vi!! th;;.ith s/,atz [ d C i t: 1.4,itre:e Ed thei futui e %"", :": Nit Na 5:!ou this g,.;E level will be siv.n ty{ ,
The daily requirement for calculating MCPR above:25% M :=:= :t y_ rated' thermal power 1s-sufficient since power distribution shifts are L , very slow when there have not been significant power or control rod- 1 I 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 oper1 tion at a thermal limit. ii.0 Power / Flow Relationship'il6EDs fu DyKk The power / flow curve is the locus of core thermal power as a function of j f'ow from which the occurrence of abnormal operating transients will .{ yleid results within defined plant safety limits. Each transient and .l ' ( postulated accident applicable to operation of the plant.was analyzed N along the power / flow'line. The analysis justifies the operating envelope ~ bounded by the power / flow curve as long as:other operating-limits are satisfied. Operation under the power / flow line is-designed to enable the direct ascension to full power within the. design basis for the plant. r a l' # g~ p ~ T . 4 *^' #
- zore.
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{ Morrecterfebruary-l&,1"Cf, @
. . 1 p ets ^N e j 1
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- 1. Samaral Electric BVR Generte Reload Fuel Application, NEDE-2&O11.F.
l 3.' L , aford, Analytical Math of Plant Transient tvaluattens - d.= ' fot' the VR, Feb 16( *
- 3. General Electric Ces ., f--ivrical Mod or tasedfeolant l Analysis in Accordanc with 10 crt 30, appendia E, Not-20566 * .l
. (Praf t), August 1974 . . 1
- 4. Letter from J. 3. Seward, eten Edison Cocyany to D. L. tianaan ]
tssc, dated October 31,197 . ]
- 5. Qualification of the One-Dimensionah Transient Model for !
Soiling Water Reactors October 1978 (NE'DE- 154-?) . '
- 6. 1Latter, R. P. Denise (NRC) to G. C. Sherwood,(CE), Je ry 23,1980: -l k
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% , FueyType 8DB219L
- Core Flow >,- 90% rated Jow rated 3 l l 12.3 11.9 12.1 l i s' # Ni\
Maximum s gN 1.5 i 11.5 Yk , F Average Planar P .2 + 10.9 Linear Heat 10.7 Generation N \\' 9.6 ! Rate (kw/ft) 9 1000 .h l-i 9.0 1 g 200 9.1 e s 5,000- 10,000 15,000 20,000 25,000 30,000 35,000-40,000 45,0
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- Coce Flow >, '90% rated
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Fuel es P8DRB265L and BPDRB265L
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m O 4 0 . O l' FIGURE 3.11-5 k \ MAP GR ersus Planar ' Average Exposure g: 3 gL\ _ e u,el N { pes P80RB282 and- BP8DRB282 ed Core Flow >,- 90% rated j Core Flow
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Q. + Q Q, ( FIGURE .3.11-6 - I. \ MAPLHG . Vers Planar Average Exposure-B- _ 1 2 , ce Ty es P8DR 265H and BP8DRB265H i [) T b l O Core Flow < 90% r Core Flow >,= 90% rated d 12.1-l' I 12.1 x 11.9 11.9 I h 12-. 1 16 l .. 11.5 # y-% ; Maximum C
/g- 11.5 11.5 \ \ 10.7 Average Planar 10.9 11 0 11.3 h -h '10.2 -
Linear Heat , .; N Generation N < g-Rate (kw/ft) V 10~2 1000 g. 9 7 s' , 1 200 ru 8 %> .! E O 5,000 10,000 15,000 20,000.25,000 30,000.35,000 40,000 45,000 [ - Planar. Average ' Exposure. (MWD /ST) -
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xx 3
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OdcnJ . sci wm =:n:cui:q _ i Nhe following actions shall be taken in the event a Safety Limit is ; viola 1ed: . A. T si ns of 10 CFR 50.36( L(ll{i) shall be complied with J- ,- immediately. K B. The Safety Limit Viola on s U M e.xaportiLt the Comission within , I hour per 10CFR50.36(c (6) and 50.72, and te the Station Director , l and the NSRAC Chairman ediately. 1 f ' i ! C. A Safety Limit Violation ReporQhall be prepared.- The report shall be reviewed by the ORC. This rep 6tt hall describe (1) applicable circumstances preceding the violation, 2 (effects of the violation l tron facility components, systems or structuresqnd (3) corrective : action taken to prevent recurrence. , D. The Safety Limit Violation Report shall he submitted to ommission' . i within 30 dayc in accordance with 10CFR50.36(c)(7) and 50.73 a to l' the NSRAC Chairman and the Station Director. l-L 6.8 PROCEDURES A. Written procedures and administrative policies shall be e'.tablished, implemented and maintained that meet or exceed the requitements and recommendations of Sections 5.1 and 5.3 of ANSI N18.7 - 972 and Appendix A" of USNRC Regulatory Guide 1.33, except as provided in - 6.8.B and 6.8.C below. ( 8. Each procedure of 6.8.A above, and changes thereto, shall be reviewed by the ORC rnd approved by the responsible department manager prior to l implementation. These procedures shall be reviewed periodically as set forth in administrative procedures. RQT.E: ORC review and approval of procedures for vendors / contractors, who have a QA Program approved by Boston Edison Company, is not required for work performed at the vendor / contractor facility. C. Temporary changes to procedures of 6.8.A above may be made provided: , L- 1. The intent of the original procedure is not altered. i l i 2. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's license on the unit affected. l l 3. The change is documented, subsequently reviewed by the ORC within ' 7 cays of implementation, and approved by the responsible L department manager. l l D. Written procedures to implement the Fire Protection Program shall be l established, implemented and maintained. l D l Amendment No. 29, 30, 46, 74, 88, /)[ 217 l
6.9 REPORTIE REQUIREMENTS In addition to the applicable reporting regairements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted to the Comission. l A. Routine Reports
- 1. Startup Repor_t summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the I Hense involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the .7uclear, thermal, or hydraulic performancs of the plant. The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the-operating conditions or characteristics obtair,ed during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any addit 19nal specific dethlis required in license conditions based on otner commitments shall be included in this report.
Startup reports shall be submitted within (1) 90 days followi1g completion of the startup test program, (2) 90 days following. resumption or ' commencement of commercial power operation, o' (3) 9 months following initial cr1tical1ty, wh1chever 1s earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test prog wii. and resumption or commencement of commercial power operation), suppl *mentary reports shall be submitted at least every three months untti all three events have been completed. 7
- 2. MonthlyOperatinaReporthoutinereportsofoperatingstatistics, shutdown experience and forced reductions in power shall be submitted on a monthly basis to the Comission to arrive no later than the 15th I of each month following the calendar month covered by the report.
s The Monthly Operatirg Report shall include a narrative summary of operating expertenre that describes the operation of the facility, including safety-ralated maintenance for the monthly report period.
- 3. OccupationalExposureTabulationhAtabulationofthenumberof station, utility snd other personnel (including contractors) receiving exposures greater than 100 arem/yr and their associated man-rem exposure sccording to work and job functions, e.g. reactor operations anit surveillance inservice inspection, routine maintenance, special maintenance (including a description), waste processing, and refueling shall be submitted on an annual basis.
This tabulattoi suppleraents the requirements of 20.407 of 10 CFR 20. The dose assigiment to various duty functions may be estimates based on pocket dositeter, TLD, or film badge measurements. Small exposures tota'11ng less than 207, of the individual total dose need not be etcounted toi. In the aggievale, at least 607, or liie total whole body dose received from external sources shall be assigned to specific major work functions. 1 4teyWorr-10tf' 218 Q wlm~f M k N, h--(-the-ntxt-page H 2M y
b.' Cw e D 4 h M u Cse. g W aholl b-t# & ed %u.~n (> k-1:f, COTE 0/2RRrlU& t -ti11tT.S ret 0AYlup M & y n a~g. m ?% & q.~ saea y ,nr 4tuenuM w-ecura Mw2Myy ts- A Q eAAZ/ h# "Ge 4 56.x Q Scr / d th NEC. L WEDE - 0 401I- P.- R,
/JyA gn M F*d, (>dpyM ,A M RepcAT) H g Q k th CohE. Ct'ERt9'rPJ& Unt ITS MEDO- QlM G, ' A Q W 0- W {c t ilgd,m W : #P-4*tet4oneH
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- 6. 9. e cactra n
L.) 9,\ % Amendment No.7Bp hj (A/pd y *w 2.2.3) --22OI
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h Attachment D to BECo 90-101 O Reolacement Technical Soecification Paaes O 4 O l
o h TABLE OF CONTENTS p Paae No.
. fm 1.0 DEFINITIONS 1 i 2.0 SAFETY LIMITS 2.1 Safety Limits 6 2.2 Safety Limit Violation 6 .
1 Limitina Conditions For Ooeration Surveillance ReauiregtILt 3.1 REACTOR PROTECTION SYSTEM 4.1 26 3.2 PROTECTIVE INSTRUMENTATION 4.2 42 3.3 REACTIVITY CONTROL 4.3 80 A. Reactivity Limitations A 80 B. Control Rods B 81 C. Scram Insertion Times C 83 D. Control Rod Accumulators D 84 , E. Reactivity Anomalies E 85 ! F. Alternate Requirements 85 G. Scram Discharge Volume G 85 3.4 STANDBY LIQUID CONTROL SYSTEM 4.4 95
- f. A. Normal System Availability A 95
'O B. Operation with Inoperable Components B 96 ,- Q . C. Sodium Pentaborate Solution C 97 O. Alternate Requirements 97 3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 103
- A. Core Spray and LPCI Subsystems A 103 -
B. Containment Cooling Subsystem B 106 C. HPCI Subsystem C 107 D. RCIC Subsystem D 108 E. Automatic Depressurization System E 109 F. Minimum Low Pressure Cooling System F 110 l and Diesol Generator Availability G. (Deleted) _ G 111 H. Maintensnce of Filled Discharge Pipe H 112 3.6 PRIMARY SYSTEM B0UNDARY 4.6 123 A. Thermal and Pressurization Limitations A 123 B. Coolant Chemistry B 124 C. Coolant Leakage C 125 D. Safety and Relief Valves 0 126 E. Jet Pumps E 127 F. Jet Pump Flow Mismatch F 127 G. Structural Integrity I G 127a
. H. Deleted H 127a I. Shock Suppressors (Snubbers) I 137a Amendment No. T5, 45, 65, i
Surveillance Pace No. I w fm (j 3.7 CONTAINMENT SYSTEMS 4.7 152 A. Primary Containment -A 152 ' B. Standby Gas Treatment System and B 158 ! Control Room High Efficiency Air Filtration System 1 C. Secondary Containment C 159
-l 3.8 RADIDACTIVE EFFLUENTS 4.8 177 A. Liquid Effluents Concentration A 177 !
B. Radioactive Liquid Effluent B 177
- Instrumentation :
C. Liquid Radwaste Treatment C 178 D. Gaseous Effluents Dose Rate D 179 E. Radioactive Gaseous Effluent E 180 ; Instrumentation .
-F. Gaseous Effluent Treatment F 181 G. Main Condenser G 182 H. Mechanical Vacuum Pump H 183 3.9 AUXILIARY ELECTRICAL SYSTEM 4.9 194 .
A. Auxiliary Electrical Equipment A 194 ! B. Operation with Inoperable Equipment 196 3.10 CORE ALTERATIONS 4.10 202 A. Refueling Interlocks A 202 B. Core Monitoring B 202 C. Spent Fuel Pool Water Level C 203
- D. Multiple Control Rod Removal D 203 l 3.11 REACTOR FUEL ASSEMBLY 4.11 205a A. Average Planar Linear Heat A 205a f
, Generation Rate (APLHGR)
- B. Linear Heat Generation Rate (LHGR) D 205b C. Minimum Critical Power Ratio (MCPR) C 205b D. Power /Finw o.:!;tionship D 205c 3.12 FIRE PROTECTION 4.12 206 A. Fire Detection Instrumentation A 206 B. Fire Hater Supply System B 206a l C. Spray and/or Sprinkler Systems- C 206c D. Halon System D 206d E. Fire Hose Stations E 206e F. Fire Barrier-System F 206e-1 G. Alternate Shutdown Panels G 206e-1 Amendment No. 75, 27, 45, 84, 89, 113. Ild, ii
l l
)
Paae No. 4
'kj) 4.0 MISCELLANEOUS RADIC1.CTIVE MATERIALS SOURCES 206k-4.1 Sealed Source Contamination 206k 4.2 Surveillance Requirements 206k ,
4.3 Reports 2061 1 4.4 Records Retention 2061 ! l 5.0 MAJOR DESIGN FEATURES 206m 5.1 Site Features 206m 5.2 Reactor Core 206m 5.3 Reactor Vessel 206m 5.4 Containment 206m 5.5 Fuel Storage 207 5.6 Seismic Design 207 6.0 ADMINISTRATIVE CONTROLS 208 l 6.1 Responsibility 208 6.2 Organization 208 6.3 Facility Staff Qualifications 208a i 6.4 Training 208a 6.5 Review and Audit 212 6.6 Reportable Event Action 216 6.7 Deleted 217 6.8 Procedures 217 l (m. s 6.9 Reporting Requirements 218 l 6.10 Record Retention 224 6.11 Radiation Protection Program 226 6.12 (Deleted) 226 6.13 High Radiation Area 226 6.14 Fire Protection Program 227 1 Ooeration11 Obiectlyn Surveillance 7.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 8.0 229 l L 7.1 Monitoring Program 8.1 229 L 7.2 Dose - Liquids 8.2 232 7.3 Dose - Noble Gases 8.3 233 7.4 Dose - Iodine-131, Iodine-133, 8.4 234 Radioactive Material in Particulate Form, and Tritium 7.5 Total Dose 8.5 234 O Amendment No. 35, 45, 88, 89, 95, 722, iii
.l.0 DEFINITIONS l l
O The succeeding frequently used terms are explicitly defined so that a i uniform interpretation of the specifications may be achieved. {) A. Safety Limit - The safety limits are limits below which the reasonable maintenance of the cladding and primary .,'tems are assured. Exceeding such a limit is cause for unit si,. N on and review by the 4 Nuclear Regulatory Commission before resumption of unit operation. Operation beyond such a limit may not in itself result in serious . consequences but it indicates an operational deficiency subject to i regulatory review. B. Limitina Safety System Settina (LSSS) The limiting safety system settings are settings on instrumentation which initiate the automatic protective action at a level such that the safety limi';s will not be exceeded. The region between the safety limit and these settings i represent margin with normal operation lying below these settings. - The margin has been established so that with proper operation of the instrumentation the safety limits will never be exceedad. C. Limitina Conditions for Ooeration (LCO) - The limiting conditions for , operation specify the minimur acceptable levels of system performance necessary to assure safe.startup and operation of the facility. When these conditions are met, the plant can be operated safely and abnormal situations can be safety controlled. D. CORE OPERATING LIMITS REPORT T'e CORE OPERATING LIMITS REPORT is a reload-cycle specific document, i,s supplements and revisions, that provides core operating limits for the current operating reload cycle. . It shall be administrative 1y controlled by the licensee. These cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.A.4. Plant operation withti these operating limits is addressed in individual specifications. LO Amendment No. I 1
2.0 SAFETY LIMITS'
/\
(j 2.1 SAFETY LIMITS r 2.1.1 With the reactor steam dome pressure < 785 psig or core i flow < 10% of rated core flow r THERMAL POWER shall be 1 25% of RATED THERMAL POWER. 2.1.2 With the reactor steam dome pressure 1 785 psig and core flow 2 10% of rated core flow: ; HINIRUM CRITICAL POWER RATIO shall be 1 1.04. 2.1.3 Roactor vessel water level shall be > the top of active irradiated fuel. 2.1.4 Reactor steam dome pressure shall be 1 1325 psig at any time when. irradiated fuel is present in the reactor vessel. 2.2 SAFETY LIMIT VIOLATION Hith any Safety Limit not met the following actions shall be met: 2.2.1 Hlthin one hour notify the NRC Operations Center in accordance with 10CFR50.72. , 2.2.2 Hithin two hours: A. Restore compliance with all Safety Limits, and B. Insert all insertable control rods. 2.2.3 The Station Director and Senior Vice President - Nuclear and the Nuclear Safety Review and Audit Committee : (NSRAC) shall be notified within 24 hours. 2.2.4 A Licensee Event Report shall be prepared pursuant to 10CFR50.73. The Licensee Event Report shall be submitted ' to the Commission, the Operations Review Committee (ORC), the NSRAC and the Station Director and Senior Vice President - Nuclear within 30 days of the violation. 2.2.5 Critical operation of the unit shcIl not be resumed until authorized by the Commission.
'A U
Amendment No. 75, 27, 42, 72, 6 1, . - -_. _ . . . --_ . . _ . . -
I B 2.0 SAFETY-LIMITS I I
-. BASES 1
L l\ INTRODUCTION The fuel cladding, reactor pressure vessel and primary system I piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these barriers during 4 normal plant operations an! anticipated transients. The fuel i cladding integrity Safety Limit is set such that no fuel l damage is calculated to occur if'the limit is not violated. Because fuel damage is not directly observable, a stepback ' approach is used to establish a Safety Limit such that the Minimum Critical Power Ratio (MCPR) is not less than the limit specified in Specification 2.1.2. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. J The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The , integrity of this claddin; 6 crier is related to its relative
- freedom from perft% tions or cracking. Although some ,
corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions. (Q Hhile fission product migration from cladding perforation is just as measuraSle as t ht from use-related cracking, the 2 thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross , rather than incremental cladding deterioration. Therefore, , the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling (i.e., HCPR of 1.0). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity Safety Limit assures that during norr'1 operation and during anticipated operational occurrences, at least 99.9% of the fuel rods in the core do not experience transition boiling. FUEL CLADDING GE critical power correlations are applicable for all INTEGRITY critical power calculations at pressures at or above 785 psig (2.1.1) or core flows at or above 10% of rated flow. For operation at low pressures and low flows another basis is used as follows: (continued) l0 Amendment No. 75, 42, 72, 105, 129, 7
BASES (continued) FUEL CLADDING Since the pressure drop in the bypass region is essentially O INTEGRITY (2.1.1) (continued) all elevation head, the core pressure drop at low power and flowswillalwaysbegreaterthanj.5 that with a bundle flow of 28 x 10 lbs/hr psi. bundle Analyses show pressure drop is nearly independent of bundle power and has a value of 3.5 psl. Thus, the bundle f willbegreaterthan28x10{owwitha4.5psidrivinghead lbs/hr. Full scale ATLAS test oi a taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MHt. With the design peaking factors, this corresponds to a THERMAL P0HER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL P0HER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative. MINIMUM The fuel cladding integrity Safety Limit is set such that no CRITICAL POWER fuel damage is calculated to occur if the limit is not RATIO violated. Since the parameters which result in fuel damage (2.1.2) are not directly observable during reactor operation, the thermal 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 result in damage to BHR 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 procedures 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 CPR in the limiting fuel assembly for which more tbn 99.9% of tne fuel rods in the core are expected to avoid boiling transition considerirg the power distribtttJon within the core and all uncertainties. The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity Safety Limit calculation are giver :n Reference 1. Reference 1 includes a tabulation of the uncertainties used in the determination of the Safety Limit MCPR and of the nominal values of the parameters used in the Safety Limit MCPR statistical analysis. (continued) O Amendment No. 15, 42, 72, 105, 129, 8 1
1 j BASES (continued) g REACTOR VESSEL With fuel in the reactor vessel during periods when the HATER LEVEL reactor is shutdown, consideration must be given to water (V (2.1.3) level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated l fuel during this period, the ability to remove decay heat is l reduced. This reduction in cooling capat'111ty could lead to , elevated cladding temperatures and clad perforation in the ! event that the water level became less than two-thirds of the - core height. The Safety Limit has been established at the top l of the active irradiated fuel to provide a point which can be , monitored and also provide adequate margin for effective action. ; REACTOR The Safety Limit for the reactor steam dome pressure has been - STEAM DONE selected such that it is at a pressure below which it can be PRESSURE shown that the integrity of the system is not endangered. - (2.1.4) The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel Code (1965 Edition, including the January 1966 Addendum), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig, f The Safety Limit of 1325 psig, as measured by the reactor ' steam dome pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The reactor coolant system is designed to the USAS Nuclear Power Piping Code, Section 831.1.0 for the reactor rectreulation piping, which permits a maximum pressure transient of 120% of O design pressures ~of 1148 psig at 562'F for suction piping and Q 1241 psig at 562*F for discharge piping. The pressure Safety Limit is selected to be the lowest transient overpressure allowed by the applicable codes, j REFERENCES 1._ " General Electric Standard Application for Reactor fuel," i NEDE-240ll-P-A (Applicable Amendment specified in the CORE ( OPERATING LIMITS REPORT). j l Amendment No. 15, 42, 72, (Next page is 26) 9
i LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS , 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Apj11cability: Anolicability: f Applies to the instrumentation Applies to the surveillance of - and associated devices which the instrumentation and , initiate a reactor scram. associated devices which initiate reactor scram. : Obiective: Obiective: To assure the operability of the reactor protection system. To specify the type and frequency ' of surveillance to be applied to Snecification: the protection instrumentation, e A. The :.' points, minimum number of Specification: , trip systems, and minimum number , of instrument channels that must A. Instrumentation systems shall be
- be operable for each position of functionally tested and the reactor mode switch shall be calibrated as indicated in Tables as given in Table 3.1.1. The 4.1.1 and 4.1.2 respectively, system response times from the opening of the sensor contact up B. Verify the maximum fraction of *
.to and including the opening of limiting power density is less i the trip actuator contacts shall than or equal to the fraction of ,
not exceed 50 milli-seconds. rated power once within 12 hours C. after thermal power is greater s B. The maximum fraction of limiting than or equal to 25% of rated > power density (MFLPD) shall be thermal power and every 24 hours less than or equal to the thereafter, fraction of rated power (FRP) when thermal power is greater than or equal to 25% of rated thermal power.
- 1. If MFLPD is greater than FRP, either adjust the APRM high ,
flux scram and rod block trip setpoints to the relationships specified in the CORE OPERATING LIMITS REPORT or adjust the APRM gain such that the APRM readings are greater than or equal to MFLPD within 6 hours.
- 2. If the required actions and associated completion times of Specification 3.1.B.1, above cannot be met, reduce thermal power to less than n 25% of rated thermal power
( within 4 hours. Amendment No. 42, 129, 26 l l
O O O TABLE 3.1.1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRtiMENTATION REOUIREMENT Minimum Number Modes in Which Function Operable Inst. . Trip Function Trip Level Setting Most Be Operable Action (l) Channels per Refuel (7) Startup/ Hot Run Trio (1) System Standby 1 Mode Switch '1 Shutdown X X X A 1 Manual Scram X X X A IRM 3 High Flux 1120/125 of full scale X X (5) A 3 Inoperative X X (5) A APRM 2 High Flux (15) (17) (17) X A or B 2 Inoperative (13) X X(9) X A or B 2 High Flux (151) 115% of Design Power X X (16) A or B 2 High Reactor Pressure 11085 psig X(10) X X A 2 High D m e11 Pressure 12.5 psig X(8) X(8) X A 2 Reactor Low Mater Level 19 In. Indicated Level X X X 'A 2 High Water Level in Scram Discharge Instr. Volume 139 Gallons X(2) X X A 2 Main Condenser low Vacuus 123 In. Hg Vacuum X(3) X(3) X A or C 2 Main Steam Line High 17X Normal Full PoJer Radiation Background (18) X X X(18) A or C 4 Main Steam Line Isolation Valve Closure 1 101 Valve Closure X(3)(6) X(3)(6) X(6) A or C 2 Turbine Control Valve 2.150 psig Control Oil Fast Closure Pressure at Accelerr' ion Relay X(4) X(4) X(4) A or D 4 Turbine Stop Valve Closure 110% Valve Closure X(4) X(4) X(4) A or D Amendment Nc. 75, 42, 86, 92 II7, 27
NOTES FOR TABLE 3.1.1 (CONT'D) [ 10.- Not required to be operable when the reactor pressure vessel head is not I bolted to the vessel.
- 11. Deleted l
- 12. Deleted
- 13. An APRM will be considered inoperable if there are less than 2 LPRM inputs per level or there is less than 50% of the normal complement of ,
LPRM's to an APRM. :
- 14. Deleted
- 15. The APRM high flux trip level setting shall be as specified in th6 CORE OPERATING LIMITS REPORT, but shall in no case exceed 120% of rated :
thermal power.
- 16. The APRM (15%) high flux scram is bypassed when in the run mode.
- 17. The APRM flow biased high flux scram is bypassed when in the refuel or
- startup/ hot standby modes.
- 18. Within 24 hours prior to the planned start of hydrogen injection with the reactor power at greater than 20% rated power, the normal full power radiation background level and associated trip setpoints may be changed
(' based on a calculated value of the radiation level expected during the ( injectionofhydrogen. The background radiation level and associated trip setpoints may be adjusted based on either calculations or measurements of actual radiation levels resulting from hydrogen injection. The background radiation level shall be determined and associated trip setpoints shall be set within 24 hours of re-establishing normal radiation levels after completion of hydrogen injection and prior to withdrawing contr)1 rods at reactor power levels below 20% rated power. l i Amendment No. 15, 27, 42, 86, II7, 718, 29
i
. 3.1 E (Cont'd) 4.1 E (Cont'd) !
s been provided to allow for To facilitate the implementation bypassing of one such channel. of this technique, Figure 4.1.1
\ is provided to indicete an "
EPE appropriate trend in test interval. The procedure is as l The average power range follows: monitoring-(APRM) system, which . is calibrated using heat balance 1. Like sensors are pooled into ; data taken during steady-state one group for the purpose of conditions, reads in percent of data acquisition. ! design power (1998 MMt). Because fission chambers provide the 2. The factor M is the exposure basic input signals, the APRM hours and is equal to the system responds directly to number of sensors in a average neutron flux. During group, n, times the elapsed transients, the instantaneous time T (M - nT). rate of head transfer from the ! fuel (react)r thermal power) is 3. The accumulated number of : ' less than the instantaneous unsafe failures is plotted neutron flux due to the time as an ordinate against M as constant of the fuel. Therefore, an abscissa on Figure 4.1.1. during abnormal operational transients, the thermal power of 4. After a trend is the fuel will be less than that established, the appropriate indicated by the neutron flux at monthly test interval to < the scram setting. Analyses satisfy the goal will te the G demonstrated that with a 120 test interval to the left of percent scram trip setting, none the plotted points. ! of the abnormal operational ! transients analyzed violate the 5. A test interval of one month fuel safety limit and there is a will be used initially until substantial margin from fuel a trend is established. damage. Therefore, the use of , flow referenced scram trip Group (B) devices utilize an provides even additional margin. analog sensor followed by an amplifier and a bi-stable trip : An increase in the APRM scram circuit. The sensor and setting would decrease the margin amplifier are active components present before the fuel cladding and a failure is almost always integrity safety limit is accompanied by an alarm and an , reached. The APRM scram setting indication of the source of was determined by an analysis of trouble. In the event of margins required to provide a failure, repair or substitution reasonable range for maneuvering can start immediately. An , during operation. Reducing this "as-is" failure is one that operating margin would increase " sticks" mid-scale and is not the frequency of spurious scrium, capable of going either up or , which have an adverse effect v down in response to an reactor safety because of the out-of-limits input. This type resulting thermal stresses, of failure for analog devices is l D G Amendment No. 79, 36 l e5 7 .i,,--e, --y,-*, - , .-.__.mm , - u g_ --. ,.--.,.,,-#,.--._w---,,. ,p--,,+w.,,--ww, a y 4---- -,-vy -
. .- . -. . .- =. . ..
I o 1 3.1 BASES (Cont'd) 4.1 MSIS (Cont'd)
._ Thus, the APRM setting was a rare occurrence and is I selected because it provides detectable by an operator who j
[V] adequate margin for the fuel observes that one signal does not ' cladding integrity safety limit track the other three. For yet allows operating margin that purpose of analysis, it is i reduces the possibility of assumed that this rare failure ! unnecessary scrams, will be detected within two hours. l i Analyses of the limiting The bi-stable trip circuit which j transients show that no scram is a part of the Group (B) l adjustment is required to assure devices can sustain unsafe- : MCPR greater than the Safety failures which are revealed only Limit MCPR when the transient is on test. Therefore, it is : initiated from ACPR above the necessary to test them , operating limit MCPR. periodically. For ope atica '.r. ihe startup mode A study was conducted of the I while tee reattor is at low instrumentation channels included , pressurt, the APRM scram setting in the Group (B) devices to ' of 15 percent of rated power calculate their " unsafe" failure provides adequate therc .1 margin rates. The analog devices between the setpoint and the (sensors and amplifiers) are . safety limit, 25 percent of predicted to have an unsafe rated. The margin is adequate to faigurerateoflessthan20X accommodate anticipated maneuvers 10- failure / hour. The bi-stable associated with power plant trip circuits are predicted to r startup. Effects of increasing haveanunsafefaglurerateof ( pressure at zero or low void less than 2 X 10-content are minor, cold water failures / hour. Considering the from sources available during two hour monitoring interval for startup is not much colder than the analog devices as assumed that already in the system, above, and a weekly test interval temperature coefficients are for the bi-stable trip circuits, small, and control rod patterns the design reliability goal of L are constrained to be uniform by 0.99999 is attained with ample 1 operating procedures backed up by margin. the rod worth minimizer. The bi-stable devices are North of individual rods is very monitored during plant operation > low in a uniform rod pattern, to record their failure history Thus, of all possible sources of and establish a test interval reactivity-input.-uniform control using the curve of Figure 4.1.1. , rod withdrawal is the most There are numerous identical probable case of significant bi-stable devices used throughout power rise. Because the flux the plant's instrumentation distribution associated with system. Therefore, significant uniform rod withdrawals does not data on the failure rates for the l involve high local peaks, and bi-stable devices should be because several rods must be accumulated rapidly. moved to change power by a significant percentage of rated l O Amendment No. 79, 37
,5 i 3.1 E (Cont'd) 4.1 E (Cont'd) power, the rate of power rise is The frequency of calibration of I p )- very slow. Generally the heat the APRM Flow Biasing Network has
(# 1 flux is in the near equilibrium been established as each l L with the fission rate. In an refueling outage. The flow I ~ assumed uniform rod withdrawal biasing network is functionally approach to the scram level, the tested at least once per month rate of power rise is no more and, in addition, cross than five percent of rated power calibration checks of the flow
, per minute, and the APRM system input to the flow biasing network .'
would be more than adequate to can be made during the functional assure a scram before power could test by direct meter reading, exceed the safety limit. The 15% There are several instruments APRM scram remains active until which must be calibrated and it the mode switch is placed in the will take several days to perform RUN position. This switch occurs the calibration of the entire when reactor pressure is greater network. While the calibration ' than 880 psig. is being performed, a zero flow . signal will be sent to half of ! The analysis to support operation the APRM's resulting in a half at various power and flow scram and rod block condition, relationships has considered Thus, if the calibration were operation with two recirculation performed during operation, flux pumps, shaping would not be possible. Based on experience at other
- i. EM genu . ting stations, drift of i instruments, such as those in the >
The IRM system consists of 8 Flow Biasing Network, is not ' pd chambers, 4 in each of the significant and therefore, to reactor protection system logic avoid spurious scrams, a channels. The IRM is a 5-decade calibration frequency of each instrument which covers the range refueling outage is established, of power level between that covered by the SRM and the APRM. Group (C) devices are active only , L The 5 decades are covered by the during a given portion of the IRM by means of a range switch operational cycle. For example, l and the 5 decades are broken down the IRH is active during startup into 10 ranges, each being and inactive during full-power one-half of a decade in sire, operation. Thus, the only test that is meaningful is the one The IRM scram setting of 120/125 performed just prior to shutdown of full scale is active in each or startup; i.e., the tests that I range of the IRM. For example, are performed just prior to use if the instrument were on range of the instrument. 1, the scram setting would be a l 120/125 of full scale for that Group (D) devi us, while similar range; likewise, if the in description to those in Group l (B), are different in use because instrument were on range 5, the scram would be 120/125 of full they (the analog transmitter / trip scale on that range. Thus, as unit devices) provide alarms, the IRM is ranged up to trips or scraw functions. An , accommodate the increase in power availability analysis is detailed n in NED0-21617A (12/78). , Q Amendment No. 79, 38
1 3.1 BASIS (Cont'd) 4.1 BASIS (Cont'd) ,
. 7- - level, the scram setting is also Surveillance frequencies for the 1
(' ranged up. The most significant SDV system instrumentation is ' sources of reactivity change detailed in Amendment Number 65. during the power increase are due NRC concurrence with this , to control rod withdrawal. For surveillance program is contained i in-sequence control rod in the Safety Evaluation Report ' withdrawal, the rate of change of and its associated Technical power is slow enough due to the Evaluation Report (TER-C-5506-66) , physical limitation of dated 11/10/82. ; withdrawing control rods that heat flux is in equilibrium with Calibration frequency of the the neutron flux, and an IRM instrument channel is divided scram would result in a reactor into two groups. These are as shutdown well before any safety follows: limit is exceeded.
- 1. Passive type indicating '
In order to ensure that the IRM devices that can be compared ' provided adequate protection with like units on a against the single rod withdrawal continuous basis, error, a range of rod withdrawal accidents was analyzed. This 2. Vacuum tube or semiconductor analysis-included starting the devices and detectors that accident at various power drift or lose sensitivity, j levels. The most severe case l involves an initial condition in Experience with passive type which the reactor is just instruments in generating i subcritical and the IRM system is stations and substations 1:' not yet on scale. 1his condition indicates that the specified l exists at quarter rod density, calibrations are adequate. For Additional conservatism was taken those devices which employ in this analysis by assuming that amplifiers, drif t specifications call for drift to be less than l-the IRM channel closest to the withdrawn rod is bypassed. The 0.41/ month; i.e., in the period
- results of this analysis show of a month a drift of .4% would that the reactor is scrammed and occur and thus providing for ,
peak core power limited to one adequate margin. For the APRM percent of rated power, thus system, drift of electronic ! maintaining MCPR above the Safety apparatus is not the only Limi+ HCPR. Based on the above consideration in determining a anaiysis, the IRM provides ctbration frequency. Change in protection against local control power distribution and loss of rod withdrawal errors and chamber sensitivity dictate a L continuous withdrawal of control calibration every seven days, i L rods in sequence and provides Calibration on this frequency . l backep protection for the APRM. assures plant operation at or I
- below thermal limits.
Reactor low Hater Level , The set point for low level scram is above the bottom of the separator skirt. This level has O ; Amendment No. 79, 39 A _.m________m._. _ _ _ __ _ _ . . _ __ . - - _ _ _ _ _ _ . _ _ _ _ - _ . _ ._m - , - , -c-, ,-.. . - .. ,,--,-_-,,.m .-.-m. --
+ w
i 3.1 BASES (Cont'd) 4.1 BASIS (Cont'd) i n been used in transient analyses A comparison of Tables 4.1.1 and dealing with coolant inventory 4.1.2 indicates that two (V) decrease. The results show that instrument channels have not been ' scram at this level adequately included in the latter Table, protects the fuel and the These are: mode switch in pressure barrier, because MCPR shutdown and manual-scram. All remains well above the safety of the devices or sensors limit MCPR in all cases, and associated with these scram l system pressure does not reach functions are simple on-off the safety valve settings. The switches and, hence, calibration l scram setting is approximately 25 during operation is not ; in, below the normal operating applicable, i.e., the switch is i range and is thus adequate to either on or off. > avoid spurious scrams. . B. The Maximum Fraction of Limiting Turbine Ston Valve Closure Power Density (MFLPD) shall be checked once per day to determine ; The turbine stop valve closure if the APRM scram requires scram anticipates the pressure, adjustment. This will normally , neutron flux and heat flux be done by checking the LPRM increase that could result from readings. Only a small number of , rapid closure of the turbine stop control rods are moved daily and valves. Mith a scram trip thus the MFLPD is not expected tn , setting of i 10 percent of valve change significantly and thus a closure from full open, the daily check of the MFLPD is resultant increase in surface adequate. , heat flux is limited such that
' MCPR remains above the safety The sensitivity of LPRM detectors limit MCPR even during the worst decreases with exposure to case transient that assumes the neutron flux at a slow and turbine bypass is closed, approximately constant rate.
This is compensated for in the ' Turbine Control Valve Fast Closure APRM system by calibrating every > three days using heat balance The turbine control valve fast data and by calibrating-closure scram anticipates the individual LPRM's every 1000 pressure, neutron flux, and heat effective full power hours using ' i flux increase that could result TIP traverse data. from fast closure of the turbine control valves due to load rejection exceeding the l repability of the bypass valves. The reactor protection system initiates a scram when fast closure of the control valves is initiated by the ac'eleration relay. This settlag and the fact that control valve closure time is approximately twice as long as O Amendment No. 42, 40
F l l 3.1 MSIS (Cont'd) ; "gi. that for the stop valves means that resulting transients, while similar, are less severe than for stop valve closure. MCPR remains above the
- t
\ safety limit MCPR. J L Main Condenser Low Vacuum f i To protect the main condenser against overpressure, a loss of condenser vacuum initiates automatic closure of the turbine stop valves and turbine bypass valves. To anticipate the transient and automatic scram resulting from the closure of the turbine stop valves, low condenser vacuum ,
initiates a scram. The low vacuum scram set point is selected to initiate i a scram before the closure of the turbine stop valves is initiated. ' Main Steam Line Isolation Valve C1glyn The low pressure isolation of the main steam lines at 880 psig (as ; specified in Table 3.2.A) was provided to protect against rapid reactor dcoressurization and the resulting rapid cooldown of the vessel. Advantage is taken of the scram feature that occurs when the main steam ; line isolation valves are closed, to provide for reactor shutdown so that high power operation at low reactor pressure does not occur; thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 785 psig requires that l the reactor mode switch be in the STARTUP position, where protection of l the. fuel cladding integrity safety limit is provided by the IRM high
- neutron flux scram and APRM 15% scram. Thus, the combination of main .
~, steam line low pressure isolation and isolation valve closure scram 4 assures 'the availability of neutron flux scram protection over the entire range of applicability of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram anticipates the pressure and flux transients that occur during normal or inadvertent isolation valve closure. With the scrams set at 10 percent of valve closure, neutron flux does not increase.
Hiah Reactor Pressure The high reactor pressure scram setting is chosen slightly above the maximum normal operating pressure to permit normal operation without spurious scram -yet provide a wide margin to the ASME Section III allowable reactor coolant system pressure (1250 psig, see Basis Section 3.6,0). Hiah Drvwell Pressure Instrumentation (pressure switches) for the drywell are provided to detect a loss of coolant accident and initiate the core standby cooling equipment. A high drywell pressure scram is provided at the same setting as the core cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return-to criticality. This instrumentation is a backup to the reactor vessel water level instrumentation. O Amendment No. 40a
k . 3.1 BASES (Cont'd) Main Steam Line Hioh Radiation
'~ High radifcion levels in the main steam line tunnel above that due to the normal nitrogen and oxygen radioactivity is an indication of leaking fuel. A scram is initiated whenever such radiation level exceeds seven times normal background. The purpose of this scram is to reduce the source of such radiation to the extent necessary to prevent e,cessive turbine contamination. Discharge of excessive amounts of radioactivity to ,
the site environs is prevented cy the air ejector off-gas monitors which t.ause an isolation of the main condenser off-gas line. : Reactor Mode Switch A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status. . Ref. Section 7.2.3.7 FSAR. t, Manual Scram l The manual scram function is active in all modes, thus providing for a manual mean; of rapidly inserting control rods during all modes of reactor ' operation. Scram Discharae Instrument Voluma The control rod drive scram system is designed so that all of the water
.f* which is discharged from the reactor by a scram can be accommodated in the discharge piping. The two scram discharge volumes accommodate in excess L of 39 gallons of water each and are at the low points of the scram discharge piping. No credit was taken for these volumes in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram.
During normal operation the scram discharge volume system is empty; however, should it fill with water, the water discharged to the piping could not be accommodated, which would result in slow scram times or-partial control rod insertion. To preclude this occurrence, redundant and . I diverse level detection devices in the scram discharge instrument volumes L have been provided which will alarm when water level reaches 4.5 gallons, i initiate a control rod block at 18 gallons, and scram the reactor when the water level reaches 39 gallons As indicated above, there is sufficient volume in the piping to acccc Adate the scram without impairment of the scram times or amount of insertion of the control rods. This function shuts the reactor down while sufficient volume remains to accommodate the discharged water and precludes the situation in which a scram would be required but not be able to perform its function adequately. A source range monitor (SRM) system is also provided to supply additional neutron level information during start-up but has no scram functions. Ref. Section 7.5.4 FSAR. The APRH's cover the " Refuel" and "Startup/ Hot Standby" modes with the 5O APRM 15% scram, and the power range with the flow
, Amendment No. 40b
.~ -. .. .. - - - .-. - .
i l 3.1 BASES (Cont'd)
]
[ A biased rod block and scram. The IRM's provide additional protection in the " Refuel" and "Startup/ Hot Standby" modes. Thus, the IRM and APRM 15% i scram are required in the " Refuel" and "Startup/ Hot Standby" modes. In ! the power range the APRM system provides the required protection. R n' . ] Section 7.5.7 FSAR. Thus, the IRM system is not required in the "Run" ' mode. - The high reactor pressure, high drywell pressure, reactor low water level , and scram discharge volume high level scrams are required for Startup/ Hot - Standby and Run modes of plant operation. They are, therefore, required . to be operational for these mooes of reactor operation. The requirement to have the scram functions, as indicated in Table 3.1.1, operable in the Refuel mode is to assure that shifting to the Refuel mode during reactor power operation does not diminish the need for the reactor protection system. l The turbine condenser low vacuum scram is only required during power operation and must be bypassed to start up the unit. Below 305 psig turbine first stage pressure (45% of rated), the scram signal due to turbine stop valve closure is bypassed because flux and pressure scram are adequate to protect the reactor. The requirement that the IRM's be inserted in the core when the APRH's read 2.5 indicated on the scale assures that there is proper overlap in the neutron monitoring systems and thus, that adequate coverage is s provided for all ranges of reactor operation. The provision of an APRM scrar., at 15% 1 design power in the " Refuel" and "Startup/ Hot Standby" modes and the backup IRM scram at 1120/125 of full scale assures that there is proper overlap in the neutron monitoring systems and, thus, that adequate coverage is provided for all ranges of reactor operation. l l The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of maximum fraction of limiting power density (MFLPD) and reactor core thermal power. The scram setting is adjusted in accordance with the formula in the CORE OPERATING LIMITS REPORT when the MFLPD is greater than the fraction of rated power (FRP). ; In a similar. manner, the APRM rod block trip setting is adjusted downward if MFLPD exceeds FRP, thus preserving the APRM rod block safety margin. As an alternative action providing equal or greater protection from exceeding safety limits, the APRM gain may be adjusted such that the APRM readings are greater than or equal to 100% times MFLPD, provided that the . adjusted APRM reading does not excoed 100% of RATED THERMAL POWER and a ! notice of adjustment is posted on the reactor control panel. Amendment No. 40c
z__________.______._.______________ . . . . . a PNPS-TABLE 3.2.C-2,
, CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS' Trio Function Trio Setooint
'APRM Upscale (1) (2)
APRM Inoperative Not Applicabis APRM Downscale 1 2.5 Indicated on Scale Rod Block Monitor (Flow Biased) (1) 9l Rod Block Monitor Inoperative Not Applicable Rod Block Monitor Downscale 1 5/125 of Full Scale
-IRM Downscale 1 5/125 of Full Scale IRM Cetector not in.Startup Positica Not Applicable
'IRM Upscale i 108/125 of Full Scale t
-IRM Inoperat a Not Applicaba
.SRM Detector not in Startup Position Not Applicable SRM Downscale 1 3 counts /secotd SRM Upscale 1 105counts /second SRM Inoperative Not Applicable Scram Discharge Instrument Volume i 18 g3llons
~later r Lavel - High.
Scram Discharge Instrument Volume - Not Applicable Scram Trip Bypassed Recircula': ion Flow Converter - Upscale 1 120/125 of Full Scale Recirt.niation Flow Converter - Not Applicable Inorot tive Recirculation Flow Converter - 1 10% Flow Deviation for > 80% Comparator Mismatch Rated Power, and 1 15% Fiow Deviation for 1 8nt. Rated Power (1) The trip level setting shall be as specified in the CORE OPERATING LIMITS REPORT. (2) -When the reactor mode switch is in the refuel or startup positions, the APkM rod block trip setpoint shall be less than or equal to 13% of rated thermal power, but always less than the APRM flux scram trip setting.
- Amendment No. 42, 110, 729, 55a
.- . ~ .
f,
- al 5
-3d BASES'(Cont'd)
The control rod block functions are provided to prevent excessive control rod withdrawal so that MCrA does not decrease to the Safety-Limit MCPR. The trip logic for this function is 1 out of n: e.g., any trip o, one of.six APRM's, eight IRM's, or four SRM's will result in a 7 rod block. The minimum instrument channel requirements assure sufficient instrumentation to assure the single failure criteria is met. The minimum instrument channel requirements for the RBM may be reduced by one for maintenance, testing, or. calibration. This time period is only 3% of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal. Reactor power ',evel 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 recirce'ation flow rate, &nd thus to protect against the condition of a MCPR 1ess than the Safety Limit MCPR. This rod block set point, which
-is automatically varied with recirculation loop ilow rate, prevents an increase in=the reactor power level to excessive values due to control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting'. over the entire recirculation flow range. The margin to the safety *tt increases as the flow decreases for the specified trip x setting versus flow relationship; therefore, the worst case MCPR which could occur during steady-state operation is at 107% of rated thermal power because of the APRM rod block trip setting. The actual power distribution 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 maximum fraction of limiting power density exceeds the fraction of rated power, thus preserving the APRM rod block safety margin.- In the startup and refuel modes, the APRM rod block function is setdown below the APRM flux scram trip.
The RBM rod block function provides local protection of the core, for a single rod withdrawal error from a limiting control rod pattern. The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. A downscale indication on an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough. In either case the instrument will not respond to c5anges in control rod motion and = v thus, control rod motion is prevented. The downstale trips are as shown in Table 3.2.C-2. O Amendm9nt No. 15, 42, !!0, 129, 71 1 l
3 . 2 -- MSf3 (Cont'd) The flow comparator and scram discharge volume high-level components-(: have only one logic channel and are not required for safety,
-The refueling interlocks also operate one logic channel, and are required for safety only when the mode switch is in the refueling position. -
For effective emergency core cooling for small pipe breaks,-the HPCI
- system must function since reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate. The arrangement of the tripping-contacts is such as-to provide this-function when necessary and minimize spurious operation. The trip settings given in the specification are.
adequate to; assure the above criteria are met. The specification preserves the effectiveness of the system during periods of maintenance, testing or calibration, and also minimizes the risk' of inadvertent-operation; i.e., only one instrument channel out of service.. Four radiation monitors are provided which initiate the Reactor Building
-Isolation and Control System and operation of the-standby gas treatment system. The instrument channels monitor-the radiation from ths refueling area ventilation exhaust ducts.
Four: instrument channels are arranged in a 1 out of 2 twice' trip logic. Trip settings of < 100 mr/hr for the monitors in the refueling area 4 ventilation exhaust ducts are based upon initiating normal ventilation isolation and standby gas treatment system operation so + hat none of the activity released during the refueling accident leaves the' Reactor - Building via the normal ventilation path but-rather all the activity is processed by the standby gas treatment system.- Flow integrators are used to record the integrated flow of liquid from the' drywell sumps. The alarm unit in each integrator is set to annunciate before the values specified in Specification 3.' C are exceeded. A system whereby the time interval to fill a k vn volume will be utilized to provide a back-up to the flow integrar>rs. An air sampling system is also provided to detect leakage inside the primary containment. 1 F u O Amendment No. 89, 72 i
o ; h,7 LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS-
;.g .3.6.C Coolant Leakaae (Cont'd) 4.6 ,
power o eration is permissible g only du ing the succeeding seven , s days. >
- 3. - If the conditions in 1 or 2 above cannot be met .an orderly shutdown t shall be initiated and the' reactor shall be in a Cold Shutdown Condition within 24 hours. ;
D. Safety and Relief Valves. D. Safety and Relief Valves 1.. During reactor power operating 1. At-least one safety valve and two conditions and prior to reactor relief / safety valves shall be- ' startup from a Cold Condition, or checked or replaced with bench whenever reactor coolant pressure checked valves once per operating is greater than 104 psig and cycle. All valves will be tested temperature greater than 340*F every two cycles. both safety valves and the safety 7 modes of all relief valves shall At least one of the relief / safety.
~
2. be operable, valves shall be disassembled and inspected each refueling outage. The nominal setpoint for the t relief / safety valves shall be 3. Hbenever the safety relief valves selected between 1095 and 1115 hre required to be operdle, the
'( .psig. All relief / safety valves discharge pipe temperature of each
.shall be set at this nominal safety relief valve shall be logged W setpoint i 11 psi. The safety daily. '
valves shall be set at 1240 psig i 13-psi. 4. Instrumentation shall be calibrated and checked as indicated in Table 4 2. If Specification 3.6.D.1 is not 4.2.F. .
' met, an orderly shutdown shall be 4
,: , initiated and the reactor' coolant 5. Notwithstanding the above, as.a L pressure shall be below 104 psig minimum, safety relief valves that within 24 hours. Note: Technical have been in service shall be
- Specifications 3.6.D.2 - 3.6.D.5 tested in the as-found condition 3 apply only when two Stage Target during both Cycle 6 and Cycle 7.
Rock SRVs are installed.
- 3. If the temperature of any safety relief discharge pipe exceeds 212*F during normal reactor power operation for a period of greater
-than 24 hours, an engineering ;
evaluation shall be performed justifying continued operatior, for ; the corresponding temperature increases. X U Amendment No. 42, 56, 88, 126
\
LIMITING CONDITIONS FOR OPERATION ' SURVEILLANCE REOUIREMENTS 3.6.D Safety Relief Valves (Con't)-
- - 4. Any safety relief valve whose discharge-pipe temperature exceeds
-212*F for 24 hours or more shall 1 -be removed at the next cold shutdown of 72 hours or more, tested in the as-found condition, and recalibrated as necessary.
prior,to reinstallation. ' Power e operation'shall not continue E beyond 90 days from the initial discovery of discharge pipe temperatures in excess of 212'F for more than 24 hours without prior-NRC approval of the engineering evaluation delineated-in 3.6.D.3.
- 5. :The limiting conditions of operation for the instrumentation that monitors tail pipe temperature are given in Table 3.2.F.
E. Jet Pumos E. , Jet Pumos
- 1. Whenever the reactor is in the Whenever there is recirculation startup or run modes, all jet flow with the reactor in the pumps shall be operable. If it is startup or run modes,' jet pump determined that a jet pump is operability shall be checked daily inoperable, an orderly shutdown by verifying that the following
-shall be initiated and the reactor conditions do not occur
-shall-be in'a Cold Shutdown simultaneously.
Condition within 24 hours.
- 1. .The two recirculation loops have a flow imbalance of 15% or more when the pumps are operated 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%.
- 3. The diffuser to lower plenum differential pressure reading on an individual jet pump
< varies from established jet pump delta P characteristics by more than 10%.
O ~Anendment No. 42, 56, 77, 93, 127
- v. ,
p LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.F ' Jet Pumn Flow Mismatch- 3.6.F -Jet Pumn Flow Mismatch: l '. Whenever both recirculation pumps Recirculation pump speeds shall be are in operation, pump speeds checked and logged at least once-shall-be maintained within 10% of. per day, each 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. If Specification 3.6.F.1 is eXc9eded immediate corrective action shall be taken. If recirt.ulation-pump speed mismatch is not corrected within 30 minutes, an orderly. shutdown shall be initiated and the reactor shall be in the Cold Shutdown condition within 24 hours unless the recirculation pump speed mismatch is brought within limits sooner.
G. Structural Intearity G. Structural Intearity
- 1. The structural integrity of the Inservice inspection of components primary system boundary shall br shall be performed in accordance with the PNPS Inservice Inspection maintained at the level requireu by the ASME Boiler and Pressure Program. The results obtained from-compliance with this program will-Vessel Code, Section XI " Rules for Inservice Inspection of Nuclear be evaluated at the completion of Power Plant Components," Articles each ten year interval. The IHA, IHB;:IHC..IND and IHF and conclusions of this evaluation will-mandatory appendices as required be reviewed with the NRC.
by 10CFR50, Section 50.55a(g), except where specific relief has been granted by the NRC pursuant to 10CFR50, Section 50.55a(g)(6)(i). H. Deleted H. Deleted O 127c Amendment No. 79, 93,
d 4 BASES: l hy 3.6.D and'4'.6 & Safety and' Relief Valves f
;The valve sizing analysis considered four, 10% capacity relief / safety valves E, and two 8% capacity safety valves. These are sized and set pressures are I" established in accordance with the following three requirements of Section III of-the ASME Code:
- 1. The lowest safety valve must be set to open at or below vessel design pressure and the highest safety valve br set at or bel e 105% of design pressure. J l- 2. The velves must limit the reactor pressure to no more than 110% of design = l pressure.
- 3. Protection systems directly- related to the valve sizing transient must not !
be credited with action (i.e., an indirect scram must be assumed). i A main' steam line-isolation with flux scram has been selected to be used as the safety valve sizing transient since-this transient results in the highest :)
. peak vessel pressure of any. transient when analyzed with an indirect scram. l l
The original FSAR analysis concluded that the peak pressure transient with I indir.act scram would be caused-by a loss of condenser vacuum (turbine trip 1 with failure of the bypass valves to open). However, later observations have J shown that the long lengths of steam lines to the turbine bufler the faster stop valve closure isolation and thereby reduce-the peak pressure caused by ,y this transient to a value below that produced by a main steam line isolation ? with flux scram. l ' Item 3 above indicates that no credit be.taken for the primary scram signal i
. generated by closure of the main steam isolation valves. Two other scram '
initiation signals would be generated, one due to high neutron flux and one due'to high reactor pressure.. Thus item 3 will be satisfied by assuming a scram due to high neutron flux. Relieving capacity of 401 (4 relief / safety valves) results in a peak pressure during the transient conditions used in the safety valve sizing analysis which is well below the pressure safety limit. The relief / safety valve settings satisfy the Cpde requirements that the lowest , safety valve set point be at or below the vessel design pressure range to I prevent unnecessary cycling caused by minor transients. The results of postulated transients where inherent relief / safety valve actuation is required
., arelgiven in Appendices R and Q of the Final Safety Analysis Report.
4 Experience in safety valve operation shows that a testing of at least 50% of the safety valves per refueling outage is adequate to detect failures or deterioration. The tolerance value of i 1% is in accordance with Section III i of the ASME Boiler and Pressure Vessel Code. An analysis has been performed ! which shows that with all safety valves set 1% higher, the reactor coolant i pressure safety limit of 1375 psig is not exceeded. Amendment No. 15, 56, 145
I,:
-1 BASES:
- , 3.6.D and 4.6.D
( /: Safety antt Relief Valves H The relief / safety valves have two functions; i.e., power relief or self-actuated by high pressure. Power relief is a solenoid actuated function (Automatic Pressure Relief) in which external instrumentation signals of coincident high drywell pressure and low-low witer level initiate the valves to open. This function is discussed in Specification 3.5.0. In addition, the valves can be operated manually. ; Pilgrim's experier.ce with 2 stage safety / relief valves has demonstrated that minimum. leakage exists when the tailpipe temperature is 215' Fahrenheit. Therefore, a reporting requirement triggered by a temperature of 212*F is conservative, and assures timely reporting before leakage reaches significant - proportions. l D L G i 4 l i c fN l Q i Amendment Nc. 42, 146 l
Y'
- LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS-
~ 3.11 REACTOR FUEL ASSEMBLY. 4.11 REACTOR FUEL ASSEMBLY Aeolicabilitv:- Anolicabilitv:
TheLLimiting Conditions for The surveillance requirements-
= 0peration associated with fuel apply to the parameters which rods apply to it.0=a paramat;rs monitor the fuel rod operating which monitor the fuel rod conditions.
operating conditions.. Obiective: Obiective: The Objective of Limiting The Objective of-the Surveillance Conditions for Operation is to Requirements is to specify the assure the performance of the type and frequency of fuel rods. surveillance to be applied to the fuei rods. Soecifications: Soecifications: A. Averaae Planar Linear Heat A. Averaae Planar Linear Heat Generation Rate (APLHGR) Generation Rate (APLHGR) During power operation with both The APLHG1 for each type of fuel recirculation pumps operating, as a func : ion of average planar the APLHGR for each type of fuel exposure thall be determined as a function of average planar daily during reactor operation at exposure shall not exceed the 225% ratec thermal power. applicable limiting value provided in-the CORE OPERATING-LIMITS REPORT. If at any-time during operation it is determined by normal surveillance-that the limiting value for APLHGR is being
. exceeded, action >shall be initiated within 15 minutes to restore operation to within the
. prescribed limits. If the APLHGR is not returned to within the prescribed limi':s within two (2) hours, the reactor shall.be brought to the Cold Shutdown condition withir 36 hours.
Surveillance and. corresponding action shall conti'nue until reactor operation is within the prescribed limits. O Amendment No. 75, 2d 27, 42, 59, 700, 205a
. . . - - ~. - - -
3. LIMIT'NG CONDITIONS FOR OPERATION SURVEILLANCE'REOUIREMENTS-Linear Heat Generation Rate (LHGR) I B. Linear Heat Generation Rate'(LHGR1 .B. During reactor power operation, The LHGR as'a function'of core. the-LHGR shall.not exceed the- height shall be checked daily limits provided in the CORE during reactor operation at 125% OPERATING LIMITS REPORT. . rated thermal' power., ; If at any time during operation it - is determined by normal - _ surveillance that the limiting , y :value for LHGR is being. exceeded, > action shall be initiated within 15 minutes to restore operation to ' within the prescribed limits. If the LHGR is not returned'to within the prescribed: limits within two L (2) hours, the reactor shall be- . .t L brought to the Cold Shutdown F condition within 36 hours.
. Surveillance and corresponding action shall continue until reactor operation is within'the prescribed limits.
C. ' Minimum Critical Power Ratio (MCPR) C. Minimum Critical Power Ratio-(MCPR)-
.f During power operation, MCPR shall MCPR shall be determined daily-
- i be greater than or equal to the- d;l"ing reactor power operation at limits provided in-the CORE i 25% rated thermal-power and OPERATING LIMITS REPORT. following any change in power '
level or distribution that-would If at any time during operation it cause operation with a limiting 3 is determined by normal control rod pattern as described surveillance that the limiting -in the bases for Specification value for MCPR is being e::ceeded, 3.3.B.S. L action.shall be initiated within > 15 minutes to restore cperation to I within the prescribed limits. If the steady state MCPR is not '* returned to within the prescribed limits within two (2) hours, the l reactor shall be brought'to the Cold Shutdown condition within 36 hours. Surveillance and corresponding action shall continue until reactor operation is within the. prescribed limits. Amendment No. 15, 27, 29, 42, 54, 105, 205b
e LIMITING CONDITIONS FOR OPERATION- SURVEILLANCE REOUIREMENTS-
.m -D. Power / Flow Relationshio Durina D.- Power / Flow Relationshin Durirg Power Oneration Power Operation e Thepower/flowrelationshipshd11. Compliance with the power /flowJ not exceed-the limiting-values relationship in Section 3.11.0 9 provided in'the CORE OPERATING shall be determined daily during.
LIMITS REPORT. ' reactor' operation. ; If at any time during power operation it is determined by normal surveillance that the ! limiting value for the power / flow relation >Mp is being exceeded, action shall be initiated within 15 minutes to restore operation 'o within the prescribed limits. if the pows;r/ flow relationship is not- ! returned to within the prescribed
. limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36-hours. Surieillance and corresponding action shall continue until reactor operation nis within the prescribed limits. ,
i p i l l l- ,-c> I. , 1 Amenda.ent No. 75, 54, 205c L-
_ .. ,m__ _ . _ . , s .- .. m.. ..m , .~. i ? 6 a BASES:
. 3.ll.A.- Averaae Planar Linear Heat- Generation Rite (APLHGR)
M This specification assures that the peak cladding temperature-following the postulated design basis loss-of-coolant' accident will
.not exceed the limit specified in 10CFR50, Appendix K.
.The' analytical method: Used to determine the APLHGR limiting values is described in the topical reports listed in Specification 6.9.A.4.
3.11.8 Linear Heat Generation Rate (LHGR) .b
& .This specification assures that the linear heat generation rate in any.
rod-is-less than the design linear heat generation rate. The analytical method used to determine the LHGR limiting value is <
<iescribed in the topical-reports listed in Specification 6.9.A.4.
3.11.C Minimum Critical Power Ratio (MCPR1 Ooeratina Limit MCPR
-For any abnormal operating transient analysis with the initial '
condition of the reactor 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 the instrument trip. settings given in Tables 3.1.1, 3.2. A' and 3.2.B'. ' The -analytical' method used to determine the Operating Limit MCPR values in the CORE OPERATING LIMITS REPORT is described in the topical ' reports' listed in Specification 6.9.A.4. By maintaining MCPR greater than or equal-to the Operating Limit MCPR, the Safety. Limit MCPR- -; L specified in-Specification 2.1.2 is maintained in the event of the most limiting ^ abnormal operating transient. 4 h l~ i. l-
~ Amendment No. 75, 24, 42, 54, 59, 205d
,-c, ,
4
' " BASES:
n, . ,4'* 3.11.D' Power / Flow Relatibnshin Durina Power Ooeration The power / flow curve is the locus of core thermal power as a function.
.of-flow;from which the occurrence of. abnormal operating transients
- ~ will yield results within defined plant safety limits. Each transient and postulated accident applicable to operation of the plant was.
t' analyzed along the power / flow line. .The analysis justifies the operatint envelope bounded by the power / flow curve as-long as other-operating limits are satisfied. Operation under the power / flow line is designed to enable the direct-ascension to full power within the de:Ign basis for the plant.'
- 4.11.C Minimum Critical Power Ratio (MCPR)
.At core '.hermal power levels less than or equal' to 25%, the reactor r will be operating at minimum recirculation pump speed and the mode ator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant. ,
experier;ce indicated that the resulting MCPR value is in excess of ' requirements by a considerable margin. Hith this low void content, any inadvtrtent core flow increase would only place operation in a more ccnservative mode-relative to MCPR. The daily requirement for calculating MCPR above 25% rated. thermal power-is sufficient since power distribution shifts are very 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 i L pg ensures-that MCPR will be known following a change in power or power Q shape (regardless of magnitude) that could place operation at a n- thermal limit.-
l L
-l I
l: y p L s l t l l Amendment No. 75, 27, 39, 42, 54, 705, 205e 1 d
. v . ,
3 5.0 MAJOR DESIGN FEATURES 51- SITE FEATURES-Pilgrim Nuclear Power Station is, located on the Mestern-_ Shore of Cape
-Cod-Bay in the Town of Plymouth Plymouth County, Massachusetts._ The site is located at approximately 41'51' north latitude and 70'35' west longitude on the Manomet Quadrangle, Massachusetts, Plymouth._
County 7.5 Minute Series (topographic) map issued by U.S. Geological Survey.: UTM coordinates are 19-46446N-3692E.- The reactor (center line) is located approximately 1800 feet from the nearest property boundary. 5.2 REACTOR CORE-The reactor vessel core design shall be as described in the CORE OPERATING LIMITS REPORT and shall be limited to those fuel assemblies which have been analyzed with NRC-approved codes and methods and have been shown to comply with all Safety Design Bases in the FSAR. The substitution of Zircaloy-4 or stainless _ steel filler rods or open water channels for fuel rods may be made in-fuel assemblies if justified by cycle-specific reload analyses using an NRC-approved methodology. Should more than 30 rods in the core, or 10 rods in any assembly, be replaced per refueling, a special report describing the number of rods replaced shall be included in the CORE OPERATING LIMITS REPORT and submitted to the NRC in accordance with Specification 6.9.A.4. 5.3 REACTOR VESSEL The reactor vessel shall be as described in Table 4.2.2 of the FSAR.
.The applicable design codes shall be as described in Table 4.2.1 of the FSAR.
.5.4 CONTAINMENT A. The principal design parameters for the primary containment shall be as given in Table 5.2.1 of the FSAR. The applicable design codes shall be as described in Section 12.2.2.8 of the-FSAR.
B. The secondary containment shall be as described in Section 5.3.2 of the FSAR. C. Penetrations to the primary containment and piping passing through such penetrations shall be designed.in accordance with standards set forth in Section 5.2.3.4 of the FSAR. O Amendment No. 29, 36, 42, 78, 98, 105, 206m
,o
~6.7 Deleted--
-6.8 ERDCEDURES A. Written procedures land admiwistrative policies shall' be established, implemented and maintained that meet or exceed the requirements and:
+
recommendations of Sections 5 1 and 5.3 of ANSI N18.7 1972 and Appendix "A" of USNRC Regulatory Guid >1.33' except as provided in 6.8.B and 6.8.C below. B. Each procedure of 6.8.A above, and changes thereto, shall be reviewed
, by the ORC and approved by the responsible department manager prior.to implementation.- These procedures shall be reviewed periodically as set forth in administrative procedures.
liQII: ORC review and approval of, procedures for vendors / contractors, who_have a QA Program approved by Boston Edison Company, is not required.for work performed at the vendor / contractor facility. C. Temporary changes to procedures of 6.8.A abovi may.be made provided:-
- 1. The intent of the original procedure is r.ot altered.
- 2. The change is approved by two members of the plant management staff, at least one of whom holds a Senior-Reactor-0perator's license on the unit affected.
- 3. The change is documented, subsequently-reviewed by the ORC within-7 days of implementation, and approved by the responsible O department manager.
D. Written-procedures to implement the Fire Protection Program shall be established, implemented and maintained. O Amendment No. 29,-30, 46, 74, 88, 722, 217
e 6.9 REPORTING REOUIREMENTS In' addition to the applicable reporting requirenents of Title 10, Code of Federal Regulations, the following identified reports shall, be submitted O to the Commission. A o Routine Reports
- 1. Startuo Report A: summary report of plant startup and power escalation test).'a shall be submitted.following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level,-(3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. The report shall address each of
' the-tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions or characteristics obtainert during the test program and a comparison of Lthese values with desig, predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other comitments shall be included-in this report.
Startup reports shall be submitted within (1)-90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation,-or (3) 9 O months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of comercial power operation), supplementary reports shall be submitted at least every three months until all three events have been completed.
-2. Monthlv Ooeratinc Reoort Routine reports of operating statistics, shutdown experience and forced reductions in power shall be submitted on a monthly basis to the Commission to arrive no later than the 15th of each month following the calendar month covered by the report.
The Monthly Operating Report shall include a narrative summary of operating experience that describes the operation of the facility, including safety-related maintenance, for the monthly report period. O Amendment No. 30, 68, 88, 703, 218
'" 6.9.A Routine Renorts (Continued) 3.- Occunational Exnosure Tabulation A tabulation of the number of station, utility and other personnel (including contractors) receiving. exposures greater than 100 mrem /yr and their associated man-rem exposure according to work.and job functions, e.g. reactor, operations and-surveillance inservice inspection, routine maintenance, special' maintenance (including a description), waste processing, and refueling shall be submitted on an annual basis. This tabulation supplements the requirements of 20.407 of 10 CFR 20. .The dose assignment to various duty functions may be estimates based or. pocket-dosim6ter, TLD, or film badge measurements. Small exposures totallini less than 20% of the individual totti dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources-shall be assigned to specific'mijor work functions'.
- 4. Core Ooeratina 1.imits ReDort Core operating limits-shall be established and documented in the CORE OPERATING = LIMITS REPORT before each reload cycle or any remaining <
part of a reload cycle. The analytical methods used to determine;the core operating. limits shall be those previously reviewed and approved-by.the NRC in NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel," (Applicable Amendment specified in the CORE . OPERATING LIMITS REPORT) and in NEDO-21696, " Loss of Coolant Analysis Report for Pilgrim Nuclear Power Station " dated August 1977, as
-amended. The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic -limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident. analysis limits) are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to-the Regional Administrator and Resident Inspector.
'6.9.8 Deleted O Amendment No. 30, (Next page is 223) 219}}