PY-CEI-NRR-1353, Proposed Tech Specs 2.1.2,2.2.1,3.3.1,3.3.6,3.4.1.1,3.4.1.2 & 3.4.1.3
| ML20082B526 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 06/28/1991 |
| From: | CENTERIOR ENERGY |
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| Shared Package | |
| ML20082B519 | List: |
| References | |
| PY-CEI-NRR-1353, NUDOCS 9107150265 | |
| Download: ML20082B526 (55) | |
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{{#Wiki_filter:. - _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ - - _ _ _ - _ e . A*lTACitMr.NT 2
SUMMARY
LISTING OF Tile PROPOSED TECIINICAL SPECIFICATION AND BASES CilANGES i 910715026S 910628 PDR ADOC1'. 05000440 ! P PDR I \
e . Attachment 2 PY-CEI/NRR-1353 L Page 1 of 4 PROPOSED TECIINICAL SPECIFICATION CIIANGES TO IMPLEMENT SINGLE RECIRCULATION LOOP OPERATION Pago Specificat3_on Title (number) Description or Action 2-1 SAFETY LIMITSI TilERHAL POVER, Add a p'trase that for single recirculation
, liigh Pressure and liigh Flow loop operation the SLHCPR is 1.08. Change (2.1.2) lHCPR Safety Limit] action statement to refer to the above Safety Limits rather than specify the value.
2-3 LIMITING SAFETY SYSTEM Add a footnote to the LCO indicating that SETTINGS REACTOR PROTECTION the APRH flow biased instrumentation SYSTEH INSTRUMENTATION setpoints may be functionally SETPOINTS (2.2.1) implemented by adjusting the APRM gains.
- 2-4 RPS INSTRUMENTATION Separate the Flov Blased Simulated SETPOINTS, TABLE 2.2.1-1 Thermal Pover-liigh (Item 2.b)
(2.2.1) function into two parts, one for two loop operation (TLO) and the other for single loop operation SLO. Add new Item 2.b setpoint and allovable value equations to reflect differences for single loop flow conditions. Relabel some existing footnotes. 2-5 LPS INSTRUMENTATION Add a footnote (a) to the end of the table ! SETPOINTS, TABLE 2.2.1-1 to explain that the % drive flow, V in SLO is ! (2.2.1) different than that during TLO. Add a ! footnote (b) to explain that the APRH l gains may be temporarily adjusted in lieu l of adjusting the APRM flow biased equations to implement SLO. List all relabeled foot-notes at the end of the table. B 2-1 BASES: INTRODUCTION (2.0) Change from specifying the exact value of the HCPR Safety Limit to referr'.ig to Specification 2.1.2 since the value changes with the mode of operation. B 2-2, BASES: TilERHAL POVER. Add a statement that Reference 2 applies B 2-3, liigh Pressure and liigh for both two and single loop operation. B 2-4 Flov (2.1.2) Delete blank pages B 2-3 and 2-4 and renumber page B 2-2 to refer to the next page B 2-5. B 2-7 BASES: RPS INSTRUMENTATION Add a statement explaining the difference SETPOINTS, ITEM 2. APRH in drive flow relationships for SLO and TLO. Add a statement explaining that temporary APRM gain adjustments may be made in lieu
- of adjusting the APRM flov biased equations.
Attachment 2 PY-CEl/NRR-1353 L Page 2 of 4 3/4 3-0 REACTOR PROTECTION Expand note (d) to explain that temporary SYSTEM INSTRUMERTATION APRH gain adjustments may be made in lieu SURVEILLANCE REQUIREMENTS, of adjusting the APRH flov biased equations TABLE 4.3.1.1-1 (3.3.1) and that the gain adjustment is not included in determining the absolute difference. 3/4 3-55 CONTROL ROD BLOCK Add a footnote to the LCO that the APRH INSTRUMENTATION flov biased instrumentation setpoints (3.3.6) may be tunctionally implemented by temporarily adjusting the APRH gains. 3/4 3-58 CONTROL ROD BLOCK Separate the Flov Blased Neutron Flux-INSTRUMENTATION SETPOINTS, Upscale (Item 2.a) function into two parts, TABLE 3.3.6-2 (3.3.6) one for TLO and the otner for SLO. Add new Item 2.a setpoint and allovable value equations to reflect the differences for SLO flow conditions. Add a footnote (b) to the bottom of the page to explain that the % drive flow, V, in SLO is different than that during TLO. Add a footnote (c) to explain that the APRH gains may be temporarily adjusted in lieu of adjusting the APRM flov biased equations to implement SLO. List all relabeled footnotes at end of the table in order. Add an overflow page 3/4 3-58a to accommodate the changes. 3/4 4-1 RECIRCULATION LOOPS Revise LCO by combining core flov and through (3.4.1.1) thermal power requirements specifying the 3/4 4-3 allovable region of operation on Pigure 3.4.1.1-1 into one paragraph. Split LCO into two separate parts specifying requirements for TLO and SLO. Delete shutdown Action a requirement with only one loop in operation. Add new Action Statements a through e specifying specific actions to take during SLO. Relabel existing Actions b and c. Add footnote specifying temporary APRM gain adjustments may be made in lieu of adjusting the APRH flov biased equations. Add new Surveillance Requirements 4.4.1.1.3 and 4.4.1.1.4 specifying pover/flov and equipment requirements that must be demonstrated during SLO and temperature restrictions that must be met prior to pover/ flow increases.
e , ; Attachment 2 PY-CEI/NRR-1353 L Page 3 of 4 f 1 3/4 4-1 RECIRCULATION LOOPS Renumber all pages and add overflow pages through (3.4.1.1) 3/4 4-2a and 2b to contain the added info. 3/4 4-3 (continued) (see retyped version of this Specification). 3/4 4-4 JET PUHPS Revise Surveillance Requirement 4.4.1.2 to (3.4.1.2) be applicable for SLO. Correct inconsist-ency regarding 4.0.4 exception. Clarify that the surveillance is only applicable for the jet pumps in an operating recirculation loop. Remove redundant reference to Specification 3.4.1.3. Clarify recirculation loop jet pump flow, recirculation loop drive flov and jet pump flov for consistency between specifications. Revise jet pump diffuser-to-lover plenam dp acceptance - limit and add a new acceptance limit for jet-pump flow. Replace the existing ' footnote to clarify that the characteristic curves are taken at several data points during post-refuel testing in TLO or upon entering SLO, only then is there an established curve to compare. future surveillances to and explicitly meet the survel) lance requirement. 3/4 4-5 RECIRCULATION-LOOP Change recirculation loop flov to recirculation FLOV (3.4.1.3) loop jet pump flov for clarity throughout the spec. Change rated recirculation flow to rated core flow within the LCO for clarity. Add to Applicability section that this specification only applies during TLO. Revise Action b to shutdovn one of the recirculation loops and enter SLO, when unable to meet mismatch limits, but ontinued operations are desired. Add a new paragraph retaining shutdown actions if it is not - desired to enter SLO. Add a 4.0.4 exception to permit startup of-the shutdovn recirculation loop. B 3/4 1-2 -CONTROL RODS Change from referring to a specific (3/4.1.3) value of 1.07 for the MCPR Safety Limit to specifying it by name to generalize reference. t u-=t e- , e r ,-r,-c-ee-- er--t--evy,r.e-,re-,.u ,,,.t,e--.wv+ry- r -y ,-- .m e w w . ww .e e,w e _ - ~.,--w,y.--e- . - _ _ _ _m--, ++v- - r ww- = -vr-vs-ww -,e-w- vv-<-
o O Attachment 2 l PY-CEl/NRR-1353 L Page 4 of 4 l l B 3/4 2-1, AVERAGE PLANAR LINEAR Clarify description of HAPTAC limits. l B 3/4 2-2 HEAT GENERATION RATE Revise to specify that the HAPLHGR limits I and (3/4.2.1) specified in the COLR are modified by l B 3/4 2-3 multiplication by a single loop HAPLilGR ' reduction factor during SLO. Explain differences betveen the SLO and two loop , ECCS analyses. Correct statements to r indicate that HAPLilGR values are nov ; presented in the COLR. Use blar.. page ' B 3/4 2-4 as an overflow page if needed. B 3/4 2-4 HINIHUH CRITICAL POVER Generalize reference to HCPR Safety Limit. and RATIO (3/4.2.2) Revise reference to pover/ flow map to B 3/4 2-5 explain that it applies to TLO and indicate that the pover/ flow operating restrictions for SLO are described in Appendix ISF. Clarify description of pover/flov dependent and operating HCPR limits to indicate how they apply to SLO. t b 3/4 3-4 CONTROL ROD BLOCK Add statement to clarify that the control INSTRUMENTATION rod block inst umentation is adjusted ior (3/4.3.6) SLO. Also add a statement allowing temporary APRH gain adjustments in lieu of
, adjusting the SLO equations. Add an overflow page B 3/4 3-4a for this new info. 1 B 3/4 4-1 Pr. CIRCULATION SYSTEM Add an explanation of the changes necessary and. (3/4.4.1) to' implement SLO, i.e., power /flov and B 3/4 4-2 equipment operation restrictions, setpoint and limit adjustments, and temperature limitations to prevent thermal stratification. Clarify recirculation system flov terms and jet pump operability criteria both during SLO and TLo and revise criteria in accordance with the requirements of NUREG/CR-3052. Clarify that the acceptance criteria (established characteristic curves) are not meaningful during post-refuel testing or upon entry into SLO until data has been taken to establish new characteristic curves.
Add a statement-that mismatch limits do not apply during SLO. Clarify that Figure 3.4.1.1-1 graphically represents the 80% rod line which varies slightly cycle to cycle. Add an overflow page B 3/4 4-2a to contain this new info. Hove 3/4.4.2 Safety / Relief Valves discussion to end of that page.
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8 e l l i ATTACHMENT 3 1 MARKED UP TECHNICAL SPECII'ICATION
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AND BASES PAGES ) l i l t 1 l l-- - - - - - - - - - - . , . , _ . , , , _ ,
e . l Attachment 3 PY-CEl /NRR-13 53 L Page I of 38 4
- 2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours and comply with the requirements of Specification 6.7.1. THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 with the reactor vessel steam dome pressure greater than M5 psig l an ore flow atsttit_than_10%_.of_IAted flow. . Gl.es .tuin sin 3 c ruirculati.n l ect ortratienD 1 APPLICABILITYrOFERATIONAFCONDITIOHs 1 and z. ACTION: (I k .k,n s.iety LM With MCPR less than-irWAand the reactor vessel steam dome pressure l greater than 785 psig and core flow greater than 10% of rated flow, be in at let.st HOT SHUTDOVN within 2 hours and comply with the requirements of Specification 6.7.1. REACTOR COOLANT SYSTEM PRESSURE 2.1. 3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4. ACTION: With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTOOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours and comply with the requirements of Specification 6.7.1. PERRY - UNIT 1 2-1 W ndmee % ':
-. .- _ _- - -._- _ .- . - - - - . _ . . - . _ . _ . . - . . . - - - - . - - _ . ~ . - - . . .
l Attach:3nt 3 PY-CEI/NRR-1353 L i Page 1 of 1# i 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS ( ~ 2.2 LIMITING SAFETY SYSTEM SETTINGS i 1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETP0INTS 2.2.1 -The reactor protection systes instrumentation setpoints shall be set consistent with.the Trip Setpoint values shown in Table 2.2.1-1. t APPLICABILITY: As shown in Table 3.3.1-1. l ACTION: With a reactor protection system instrumentation setpoint less conservative than the value shown ,he Allowable Values column of Table 2.2.1-1, declare the channel inoperabl .d apply the applicable ACTION statement requirement ,
-of Specification 3.3.1 ntil the channel is restored to OPERA 8LE status with !
itssetpointadjusted< nsistent with the Trip Setpoint value. I AaA '
.?-
l t f Insert A
- The APRM flow biased instrumentation need not be declared inoperable upon {
entering single recirculation ~1oop operation provided the setpoints are-adjusted within 8 hours per Specification 3.4.1.1. s. PERRY JNIT 1 2-3
r ,0 , TABLE 2.2.1-1 REACTOR PROTECTION SYSTE INSTRUMENTATION SETPOINTS E ALLOWABLE Q TRIP SETPOINT VALUE5
, FUNCTIONAL UNIT
$ 1. Intermediate Range Monitor I
- a. Neutron Flux-High < 120/125 divisions < 122/125 divisions
*' Ef full scale if full scale
- b. Inoperative e NA
,Re p la.cc -wIth-- Insert B -
- 2. Average Power Range Monitor:
Neutron Flux-High Setdown < 15% of RATED < 20% of RATED a. THERMAL POWER THERMAL POWER b./ Elcw Biased Simulated Thermil Power-Hig4 7 , i (/ ,/1)/ i Flow Bia' sed / / /
/ / <M.66 W64%.Aith
/
/ <'O.664+67%dith
,/I maximum of ' /
e
/ ,/ / f
/
p/ ,// / <J111.0% maxim'un of' of RATED / < ,113.0Lof RATED ,/ ( /, - ,2) High F1pw'Clampe'd ,7 ' V <
' ' / / / THERMAL 40WW THFRMOWER/ -
l Neutron Flux-High < 118.0% of RATED < 120.0% of RATED
- c. THERMAL POWER THERMAL POWER y
WA NA
; d. Inoperative
- 3. Reactor Vessel Steam Dome Pressure - High 1 1064.7 psig i 1079.7 psip
> 177.7 inches abev q > 177.1 inches a Reactor Vessel Water Level - Low, Level 3 l 4.
Topofactivefu(}Kjy Top of active fu N l '
- 5. Reactor Vessel Water Level-High, Level 8 < 219.5 inches ak i< 220.1 inches a-top of active fu @ Top of active fu%$t-
- 6. Main Steam Line Isolation Valve - Closure i 8% closed i 12% closed
< 3.0 x full power < 3.6 x full power
- 7. Main Steam Line Radiation - High Eackground Sackgroend /
/ 15 5 1.68 psig 1 1.88 psig ; A;
- 8. Drywell Pressure - High )
v use "See Bases Figure B 3/4 3-1 *
!C $$
i Le e .:.+<+ i+ke+brf t- N 2' N {cu w.,a r..t a. ta a + - - i
. . _ _ _ . _ _ .. .. . . . _ ._ .- _ . . _ _ _ _ _ _ _ . - . _ _ . . _ _ _ _ _ _ ~ . _ _ _ _ __
o . Attachment 3 PY-CEI/NRR-1353 L Page j _ of R nsert B
- b. Flow Diased Simulated Thermal Pover-liigh
- 1) During two recirculation loop operations
- a. Flov Blased < 0.66V + 64% I "), < 0.66V + 67% I "), ,
Uith a maximum of with a maximum of
- b. liigh Flow Clamped < 111.0% of RATED < 113.0% of RATED THERHAL POVER filERHAL POVER -
- 2) During single recirculation loop operations
- a. Flov Biased < 0.66V + 42.7%(a)(b) < 0.66V + 45.7%(a)(b)
- b. liigh Flow Clamped Not Required Not Required OPERABLE OPERABLE 1
_ - _ , . - , ~_-. , . - - - . - , _ . . _ . . , , , . , , .......,_....<....,._,_,,.._.,,,-.,..-.m,,,,,,,.,,--.., . , - , , - . , . . , . _ , - , _ . . . , , , . , ~ . , .,
TAB J 7.2.1-1 (Continued) . 5 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS "n All0WABLE 1 TRIP SETPOINT VALUES c_ FUNCTIONAL UNIT 5
- Scram Discharge Volume Water level - High 9.
~ a. Level Transmitter -< 37.9 inchesM 4 -< 38.87 inchet l (626' 6.6" elevation) (626' 7.56" elevation)
- b. Float Switches < 626' 9.87" elevation < 626' 11.5" elevation CllN013A CllN0138 7 626' 10.25" elevation 7 626* 11.5" elevation CllH013C ~7 626' 10.87" elevation 7 626' 11.5" elevation C11N013D 1 626' 11.18" elevation [626'11.5" elevation
- 10. Turbine Stop Valve - Closure 1 5% closed i 7% closed l ?3N l
Turbine Control Valve Fast Closure. Valve gag
- 11. > 465 psig g g.
Trip 011 Pressure - tow > 530 psig his T e 12. Reactor Mode Switch Shutdown Position NA NA gg NA NA
- 13. Manual Scram (A w
- c*
l
?
N (a) The Average Power Bange Monitor flcv biased scram function varies as a [' AdicE Ne r _tes (a) a. A G)# function of recirculation loop drive flov V. During single loop operation U p reldeles t ..+ ..te s a n d is adjusted to account for the difference in indicated drive flov as described in the Bases.
' d erderth Iist.-
i (b) To functionally implement this protective function during entry into single loop operation, APRM gain adjustments may be made in lieu of adjusting the M APRM Flow Blased Simulated The mal Pover-High Trip Setpoint and Allowable ns I r a Period not to exceed 72 hours provided that the APRM h"tevelzerois 623' 4.69" elevatioln>. Value gains e9uati to a value at least 21.3% of RATED THERMAL POWER higher are adjusted than the actual ttiermal power. (c) See Bases Figure B 3/4 3-1. (d) Level zero is 623' 4.69" elevation. .
, e l
( Attachment 3 1 l'Y-CEl/NRR-1353 L rage G of 3R
, 2.1 SAFETY LIMITS BASES th vaine.i 3 m in Spedthd[7[M
- 2. 0 INTRODUCTION Mis
. + lhdt
, , (4 the corre spe.U.'j rw. A)
,ma -
The fuel clad ing, reacto prenar. esivt antqWTiMFy sistes piping are the principal b(arriers to t ( release of radioactive materials to the environs. Safety Limits are estaffiQhed to protect the integrity of these barriers during nomal')lant operationsgd anticipated transients. The fuel cladding integrity Safety'lliaQt is set suchMha no fuel damage is calculated to occur if the limit is not vi lated. Because fue Mamage is not directly observable, a step back approac is used to establish A S the HCPR is not less than -144 HCPRgreaterthan.WL.kafetyLimit,suchthat 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 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 and the Limiting Safety System Settings. 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 whit.h still greater thermal stresses may cause gross rather tnan incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation. 2.1.1 THERMAL POWER. Low Pressure or Low Flow The use of the General Electric critical power correlations (Reference 1) are not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by estab-lishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 103 lbs/hr, bundle pressure droo 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 2B x 103 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative. PERRY - UNIT 1 B 2-1 Amendment 'lo. 20
. _ _ _ _ _ _ _ _ _ _ - - _ - - - ~
Attachmont 3 PY-CI:l /NRR-1353 L SAFETY LIMITS page 1 of g BASES 2.1.2 THERMAL POWER, High Pressure and High Fly The fuel cladding integrity Safe'.y Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the themal 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 th'It a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is ca'culated to c: cur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state ind 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 assemoly for which more than 99.9*, of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties. The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis GEIAB (Reference 1), which is a statistical model that comoines all of the uncertainties in opei'ating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is detemined using a GE critical power correlation. This correla-tion it, valid over the range of conditions used in the tests of the data used l to develop the correlation. Details of the fuel cladding integrity safety limit calculation are given in Reference 2. Reference 2 provides the uncertainties used in the determination of the Safety Limit MCP4and of the nominal values of the parameters used in the Safety Limi1JCPR statistical analysis. for L th 4we le.p oR sigle r e d r u l.t ,*. 3 f.ep op mtley "3enet al Electric BWR Themal Analysis Bases (GETAB) Data. Correlation i and Design Application," NED0-10958- A.
- 2. " General Electric Standard Application for Reactor Fuel, GESTAR-II,"
NEDE-240ll-P-A (latest approved revision). NL')* 8 PERRY - UNIT 1 B 2- 2 *' Amendment No. 20
O^ o I Attachment 3 ! PY-CEl/NRR-1353 L D e le.f c. l'aso L . of _3_ L Pye.
/
?
i This page inten ona left blank 4 PERRY - UNIT 1- B 2-3 Anendment tio. 20
o e Attachment 3 PY-ct.1/NRR-1353 L
. m
- Page _1 _ of lj_ ,
De lete f*j" C- This page t ntionally lef t blank l-l 1 I k PE Y - UNIT 1 B 2-4 Amendment No. 20
Attachment 3 PY-CEI/NRR-13$3 L LIMITING SAFETY SYSTEM SETTINGS '# --- Oo BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) Average Power Ranae Monitor (Continued) 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shatdown before the power could exceed the Safety Limit. The 15% neutron flux trip remains active until the mode switch is placed in > the Run position. The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the sys- > tem and therefore the monitors respond directly and quickly to changes due to transient operation for thG case of the Neutron Flux-High setpoint; ; i.e, for a power increase, the THERMAL POWER of the fuel will b1 less than that indicated by the neutron flux due to the time constants of the heat trans-fer associated with the fuel. For the Flow Biased Simulated Thermal Power-High setpoint, a time constant of 610.6 seconds is introduced into the flow biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1. I The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unneces-sary shutdown. 4-- In s e r t C (Cutisme para.)
- 3. Reacter Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear '
system process barrier resulting in the release of fission products. A pres-sure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip set-ting is slightly higher than the operating pressure to pemit normal operation without. spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power / flow conditions when the turbine control valve fast closure and turbine stop vdve closure trips are bypassed. For a load rejection or turbine trip under these conditions, the transient analysis indicated an adequate margin to the thermal hydriulic limit. PERRY - UNIT 1 8 2-7
a e Attachment 3 PY-{El /NRR-135 3 L Pag _j{,of g
.rYvy -insert C T tj for single tecitculation loop opetation, the reduced APRM setpoints are based on a delta V value of 8%. The delta V value corrects for the difference in indicated drive flov (in percentage of drive flow which produces rated cote flov) between two loop and single loop operation at the same core flov. The decrease in setpoint from HEOD to SLO conditions is derived by multiplying the slope of the flow biased setpoint equations by 8%, which results in a reduction in setpoint of L.3%, and then decreasing an additional 16% to account for the difference between the HEOD and standard pover-flov map todlines for a total decrease in setpoint of 21.3% for SLO. APRH gain adjustments of at least 21.3% (to be consistent with the above basis) n..f be made to functionally implement the single loop equation setpoints (vhile these equations 4,re being chanced over) f or a limi+;cd time period.
e k V A
- .=. -
A,- e q
]
4 TABLE 4.3.1.1-1 (Continued) g REACTOR PROTECTION SYSTEM INSTRUMENTATIGN SURVEILLANCE REQUIREMENTS I E CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH j g FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED
- 10. Turbirm Stop Valve - Closure NA M R 1 l 11. Turbine Control Valve fast Closure Valve Trip System Oli Pressure - Low NA M A 1 L 12. Reactor Mode Switch i Shutdown Position ,
NA R NA 1.2,3,4,5 f 13. Manual Scram NA M, MA 1,2.3,4,5
- l. (a) Neutron detectors say be excluded from CHANNEL CALIBRATION. -
l (b) The IRM and SRM channels shall be determined in sverlap for at least 1/2 decades during each startup af ter entering OPERATIONAL CONDITION 2 ana ihe IRM and APRM channels shall be determined to overlap for R at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days. , (c) Within 24 hours prior to startup, if not performed within the previous 7 days.
- Y (d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values j' calculated by a heat balance during OPERATIONAt' CONDITION 1 when THERMAL POWER > 2SE of RATED THERMAL i POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER
! (e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a ! calibrated flow signal. i (f) The LPRMs shall be calibrated at least once per 1000 WO/T using the TIP system. (g) Calibrate trip unit setpoint at least once per 31 days. ! (h) Verify measured core flow (total core flow) to be greater than or equal to established core flow at the
- existing loop flow (APRM % flow).
(i) This calibration shall consist of verifying the 610.6 second simulated thermal power time constant. e (j) This function is not required to be OPERABLE when the reactor pressure vessel head is removed . per Specification 3.10.1. mm> (k) With any control rod withdrawn. Not applicable to control ' rods removed per , i Specification'3.9.10.1 or 3.9.10.2. *{" m" 1 { (1) This function is not required to be OPERA 8tE whsn Drywell Integrity is not required. (m) The CHANNEL CALIBRATION shall exclude the flow reference transmitters, these transmitters shell be
-y M5a j calibrated at least'once per 18 months. o7"
' meu To functionally implement this protective function during entry into single U loop operation, APRM chant:el gain adjustments may be made in . lieu of $ "* j adjusting the APRM Flow Biased Simulated Thermal Power-High Trip Setpoint and I # ! Allowable Value equations for a period not to exceed 72 hours, provided the l criteria in Note b to Table 2.2.1-1 are met. Any APRM channel gain
- i adjustments made in compliance with Specifications 2.2.1 and 3.3.1 shall not (heincludedindet'erminingtheabsolutedifference.
o o i Attachment 3 INSTRUMENTATION PY-CEI/NRR-1353 L g. Page l3 of 38 3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION LIMITING CONCITION FOR OPERATION 3.3.6. The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.6-2. APPLICABILITY: As shown in Table 3.3.6-1. ACTION: AgA
- a. With a control rod block instrumentation channe trib'setpointless conservative than the value sho%r in the All le Values column of Table 3.3.6-2, declare the channel inoperabl til the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
- b. With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip function requirement take the ACTION required by Table 3.3.6-1.
SURVEILLANCE REQUIREMENTS 4.3.6 Each of the above required control rod block trip systems and instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1. Insert
% _ -s
- The APRM flov biased instrumentation need not be declared inoperable upon entering single recirculation loop operation provided the setpoints are adjusted within 8 hours per Specification 3.4.1.1.
PERRY - UNIT 1 3/4 3-55
A TABLE 3.6-2 . 2 j C0dTROL ROD BLOCK INSTRUMENTATION SETPOINTS j TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE
/
t !
- 1. R0D PATTERN CONTROL SYSTEM i E a. Low Power Setpoint M C^1 20 + 15. - 0% of RATED THERMAL P i G b. RWL - High Power Setpoint 20 + 0, 70 + 15,- 15%
- 0%'of RATEDTHE of RATED THERMALL POWOWE 70M (*)- 15% of RATED THERMAL P
+ 0,
- 2. APRM Rer!*ge wiQ Instef D) i 8% D A </1 0 R9)AI. [
- b. Inoperative NA NA 2
- c. Downscale -> 4% of RATED THERMAL POWER -> 3% of RATED THERMAL POWER
- d. Neutron Flux - Upscale !
Startup < 12% of RATED THERMAL POWER < 14% of RATED THERMAL POWER
- 3. SOURCE RANGE MONITORS i w a. Detector not full in NA NA I
1 b. Upscale < 1 x 105 cps < 1.6 x 105 cps w c. Inoperative NA NA g d. Downscale 1 0.7 cp U 10.5cpbAh j 4. IMERMEDIATC RANGE MDNITORS i a. Detector not full in NA NA { b. Upscale < 108/125 division of full scale < 110/125 division of full scale . l c. Inoperative NA NA
- d. Downscale 1 5/125 division of full scale 1 3/125 division of full scale ;
j 5. SCRAM DISCHARGE VOLUME ,,, l ! a. Water Level -'High < 16.6 inche (th < 17.48 inche @b E* {% mR j 1 - (624' 3.3" elevation) - (624' 4.17" elevation) - :r i 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW L_ M
- = :2 r
- a. Upscale i 111% of rated flow 1 114% of rated flow o7"
~~w l
- 7. REACTOR MODE SWITCH SHUTDOWN- 3 i POSITION NA j y nsert E NA %"
f *The Average Power Range Monitor rod block function is varied as a. function of recirculation loop flow (W). ;
**The actual setpoints are the corresponding values of the turbine first stage pressure for these power levels.
*** Level zero is 622.' 10.69" elevation; level transmitter readout. ;
#Provided signal to noise ratio 1 2. -
m_ _-m ! These fWt etes are eiew M* ** f**t ".tes te b . ten, ef new a_Adea overfg. ,g,y,3j,3_gg
- i l cochin eA jn I,,s ert E. -
C "'"J e4 f..ts,.te !='a els .d e ee A.,, A lig; fy 4 l _ - _ _ _ = - - _ w _ _
. t Attachment 3 PY-CEI/NRR-1353 L Page_jl~of_38._
Insert D
- a. Flow Blased Neutron Flux - Upscale
- 1) During two recirculation loop operations
- a. Flov Biased < 0.66V + $8%(b) , with < 0.66V + 61%(b) , with a maximum of a maximum of
- b. High Flow Clamped j 108.0% of RATED $ 110.0% of RATED THERMAL POVER THERMAL POVER
- 2) During single recirculation loop operations
- a. Flow Biased < 0.66V 4 36.7%(b)(c) < 0.66V + 39.7%(b)(c)
- b. High Flow Clamped Not Required Not Required OPERABLE OPERABLE tE
%_ ~ ~ _.
(a) The actual setpoints are the corresponding values of the turbine first stage , pressure for these power levels. (b) The Average Pover Range Monitor flov hiased control rod block function varies as a function of recirculation loop drive flow V. During single loop operation V is adjusted to account for the difference in indicated drive flov I as described for Specification 2.2.1 in the Bases. (c) To functionally implement this protective function during entry into single loop operation, APRH gain adjustments may be made in lieu of adjusting the APRM Flow Biased Neutron Flux - Upscale Trip Setpoint and Allovable Value equations for a period not to exceed 72 hours provided that the APRM gains l are adjusted to a value at least 21.3% of RATED THERMAL POVER higher than the - actual thermal power. (d) Provided the signal to noise ratio > 2. (e) Level zero is 622' 10.69" elevation level transmitter readout. l 1 l
. e l
Attachment 3 PY-CE1/NRR-1353 !. 3/4.4 REACTOR COOLANT SYSTEM Page j 6_ of 4_ w- - - u b 3/4.4.1 RECIRCULATION SYSTEM A rh[ b.rsien of t)is spSt RECIRCULATlW LOOPS I5 f. ' ' .
^5 A * "I **"* 1 0 I'"
LlHITING CONDITION FOR OPERATION Tbt 3.4.1.1 b reactor coolant system recirculation loop (s) shall be in operation withX th
)( [otal core flow greater than or equal to 45% of rated core flow, or g
)( CTHERMAL POWER less than or equal to the limit specified in Figure M 3.4.1.1-13 u A eithus Inse.rt F APM'CABILITY: OPERATIONAL CONDITIONS la and 2*.
ACTION: R c.r lue. a/ *th ute retctor coolant *ystep'recirculationjoop tot in TpTiatTon, p g.9 g / e mmpdiateIy al iglim)t sp,etifjed in/igurd 3.%.1-/wlydn Ofioups anditia)
""'I b j nit)6'tfth te y next 12 t4urs/ /
sures te plge tp unpVinj( lepst Hgl SHtKDOWhit,hfn,/ ' p
- f. /. With no reactor coolant system recirculation loops in operation, l' g immediately initiate action to reduce THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1 within 2 hours and 3ascrt H initiate measures to place the unit in at least STARTUP within 6 hours
- - - = = - and in HOT SHUTDOWN within the next 6 hours.
one. er 3.p'. With4two reactor coolant system recirculation loopt in operation and AM
'M total core flow less than 45% of rated core flow and THERMAL POWER greater than the limit specified in Figure 3.4.1.1-1:
I ns e.r t 7., ##
- 1. Determine the APRM and LPRMM noise levels (Surveillance 4.4.1.1.2):
a) At least once per 8 hours, and b) Within 30 minutes after the completion of a THERMAL POWER increase of at least 5% of RATED THERMAL POWER.
- 2. With the APRM or LPRd neutron flux noise levels greater than three times their established baseline noise levels, imediately initiate corrective action to restore the noise levels to within the required limits within 2 hours by increasing core flow to greater than 45% of rated core flow or by reducing THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1.
*SeeSpecialTestException3.10M
#15 MDetector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center o' ?: e core should be monitored.
PERRY - UNIT 1 3/4 4-1
i e Attachment 3 PY-CEl/NRR-1353 L P
~ ~' f 1 ~~
Ins m V
- a. Two recirculation loops operating, os
- b. A single recirculation loop operating with tN t ' 'ving limits and conditions:
- 1. a) The Hl!11 HUH CRITICAL POVER RATIO (HCI'R) Saf et, ,, 78-adjusted fot single recirculation loop operati or Specification 2.1.2, and b) The Maximum Average Planar Linear Heat Generation Rate (HC 'llGR) limits adjusted for single tecirculat o i loop 01 > nation per the CORE OPERATItJG LlHITS REPORT in accordance with Specification 3.2.1, and c) The Average Power Range Monitor ( APRH) Scram and Rod Illock Trip Setpoint and Allovable Value equations adjusted to those valygs applicable for single recirculation loop operation per Specifications 2.2.1 and 3.3.6, and
- 2. A volumetric recirculation loop drive flow less than or equal to 48,500 gpm, and
- 3. The recirculation flow control system in the Loop Hanual (Position Control) mode, and
- 4. TilERHAL POVER less than or equal to 2500 Megawatts-thermal.
i
. .- i h
i Attachment 3 PY-CEI/NRR-1353 L
- Page l8 of 3R nsertG]
- a. Upon initial entry into single loop operation, adjug}ments to the i limity,and setpoints of Specifications 2.1.2, 2.2.1 , 3.2.1, and 3.3.6 shall be implemented within 8 hours, or declare the associated equipment inoperable- (or declare the associated limits to be "not satisfiedd), and take the ACTIONS required by the applicable specifications.
** To functionally implement these protective functions during entry into single loop operation, APRM gain adjustments may be made in lieu of ndjusting the APRM Scram and Rod Block Flov Biased Setpoints for an interim period of 72 hours.
( MLKt [y L)
- b. During slugle loop operation, with the volumetric recirculation loop drive flow greater than the above limit, immediately initiate corrective action to reduce flow to less than or. equal to the above limit within 1 hour.
- c. During single loep operation, with the recarculation flow control system not in the Loop Manual mode, immediately initiate corrective action to place the recirculation flow control system in the Loop Manual node within 1 hour.
- d. During single loop operation, with THERMAL POVER greater than the above limit, immediately initiate corrective action to reduce THERMAL POWER to less than or equal to the above limit within 1
-hour.
- e. During single loop operation, with either:
- 1. THERMAL POVER $ 30% of RATED THERMAL POVER and temperature differences exceeding the limits in S rveillance Requirement 4.4.1.1.4, or
- 2. recirculation loop jet pump flow in the operating loop f 50%# '
of rated (two loop) core flow and temperature differences ( exceeding the limits in Surveillance - ement 4.4.1.1.4, suspend THERMAL POVER and racirculation loop flow increases. [nsert
# A conservative initial value. A lover recirculation loop jet pump flow value may be determined during SLO and submitted for approval, based upon the threshold flow v'..ich vill sweep the cold water from the vessel bottom head preventing stratification.
= - _ - _ _ __--___ _______ -_
Attachment 3 REACTOR COOLANT SYSTEM PY-CEI/NPR-1353 L p: Page 1 of X SURVEILLANCE REQUIREMENTS 4.4.1.1.1 Each reactor coolant system recirculation loop flow control valve thall be demonstrated OPERASLE at least once per 18 months by:
- a. Verifying that the control valve f ails "as is" on loss of hydraulic pressure at the hydraulic control unit, and
- b. Verifying that the average rate of control valve movement is:
- 1. Less than or equal to 11% of stroke per second opening and 4
- 2. Less than or equal to 11% of str ke per second closing. y
% 9' 4.4.1.1.2 Establish a baseline APRM and LP X utron flux noise value wi in n the regions for which monitoring is required pecification 3.4.1.1, ACTIO C O j within 2 hours of entering the region for which ronitoring is required unless baselining has previously been performed in the region since the last refueling outage.
g' .
)tectorlevelsAandCofoneLPRMstringpercoreoctantplusdetectorsA and C of one LPRM string in the center of the core should be monitored.
PERRY - UNIT 'l 1/' ^-2
s . Attachment 3 m n PY-CE1/NRR-1353 L Added New Page of Surveillances Page 2 0__ of j_ REACTOR COOLANT SYSTEM S1!REH16ELREQUIEDlERIMCnntlnp e d ) 4.4.1.1.3 Initially, within 1 hour upon entry into single recirculation loop operation and once per 12 hours thereafter, verify that: w
- a. The volumetric recirculation loop drive flov of the operating loop 7 is less than or equal to the limit stated in Specificacion 3.4.1.1.b.2, and
- b. The recirculation flow control system for the operating loop is in the Loop Manual (Position Control) mode, and
- c. THERMAL POVER is less that or equal to the limit stated in Specification 3.4.1.1.b.4.
4.4.1.1.4 Vith one reactor coolant system recirculation loop not in operation, and either THERMAL POVER less than or equal to 30% of RATED TilERMAL POVER or the regirculation loop jet pump flow in the operating loop less than or equal to 50% of rated (two loop) core flow, verify within 15 miautes prior to an increase in THERMAL POVER or recirculation loop jet pump flov that the following differential temperature requirements are met.
- a. $ 100 F between reactor vessel steam space cooiant and bottom head drain line coolant, and
- b. $ 50 F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and
- c. $ 50 F between the reactor coolant within the loop not in operation and the operating loop.
The differential temperature requirement of 4.4.1.1.4.a does not apply when the reacto) pressure vessel is below 25 psig. The differential temperature requirements of 4.4.1.1.4 b ar c do not apply when there is no flow through the loop not in operation dur to either one or both the loop suction / discharge valves (s) being closed.
# A conservative initial value. A lower recirculation loop jet pump flow value may be determined during SLO and submitted for approval, based upon the threshold flow which vill sweep tne cold water from the vessel bottom head preventing stratification.
PERRY - UNIT 1 3/4 4-2b
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PERRY - UNIT 1 3/4 4-3 5' - - - - . _ ~ - _ - _ - _ _ - _ _ _ - _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - - _ . - - -
l ( Attachment 3 PY-CEl/NRR-1353 L gACTOR COOLANT SYSTEM Page __A2 of 38 keh JET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION:
'iith one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours.
M SURVEILLANCE REOUIREMENTS in an operatine re circulatten 1..f 4.4.1.2 Each " "-[" r:;;ir:d jet pumpyAshall'be demonstrated OPERABLE _ ;M:r te-T"C?"A ?^ gen ==:dic,, "A .T l',TO TllEP"",L PWE.1, ;cd at least once g.. er 24 hour) hLgreater than 25% of RATED THERMAL POWER, by determining L vngann recirculation loopPflow, i: S ce% ow fl and diffuser-to-lower plenum differen-rowsR tial pressure for each jet pump and verifying that no two of the following 85 conditions o urnt= b:0 r: ir :kti= i=; indi::ted f'r r: ' n =g' i =:: - I t mith-!;=' N. qti n L '.1.h-0r it.wO L;ye Gecircut.fl. l..r je t f@
- a. The indicated recirculation loopaflow differs by more than 10% from the established
- flow control valve position-J+ep flow characteristics.
redrculatien l.. drive
- b. The indicated -tet:?r nt:Vflow c) jet puni[iffers by more than 10% from the established * " --- A low f value derived from recirculation loop drive.
flow measurements. et r Medr.ation f.d JehM 6r jet > famp f t,w)
- c. The indicaYed 4 difYuser-to-lower plenum differential pressure 40f any i individual jet pump differs from established" patterns by more than j -1:0E Ao L (or lo h 6 e flow),
i I
)y The provisions of Specification 4.0.4 are not applicable provided
-; that this surveillance is performed within 24 hours after exceeding 25% of RATED THEPJ8.AL POWER.
Data shall be recorded following each refueling outage or upon first entry into single loop operation during an operating cycle, in order to Al[L establish the specified relationships for that cycle / mode of operation. 1 - Comparisons of the actual data shall commence upon the establishment of the specified relationships for that cycle / mode of operation.
--* T be det:-ined during th:-:t rte; t::: pr:gr m l
l PERRY - JNIT 1 3/4 4-4
l l l Attachment 3 REACTOR COOLANT SYSTEM PY-CEI/NRR-1353 L [- Page _23 of Sg n RECIRCULATION LOOP FLOW LIMITING CONDITION FOR OPERATION jet pumt 3.4.1.3 Recirculation loopVflow mismatch shall be maintained within: cere
- a. 5% of rated .necircehthr flow with core flow greater than or equal to 70% of rai.ed core flow, c e r e.
- b. 10% of rated M-"!atbe-flow with core flow less than 70% of rated core flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*x A uria) t w e redral. tan ACTION:
"I *I '"*
jetpar With recirculation loop gflows different by more than the specified limits, either: je. tramp
- a. Restore the recirculation loop 4 flows to within the specified limit within 2 hours, or S hat 4.ww ene ,f the redreulatan lo.
- b. - 0;;h c th; r;;irhhtha h:; with ".:rshu:r 'b et S Opret4en-and take the ACTION required by Specification 3.4.1.1.
Otherwise, immediately initiate action to reduce THERMAL POVER to less than or AAA equal to the limit specified in Figure 3.4.1.1-1 vithin 2 hours and initiate measures to place the unit in at least HOT SHUTD0VN vithin the next 12 hours. SURVEILLANCE REQUIREMENTS jet >wmg 4.4.1.3 Recirculation loco low mismatch shall be verified to be within the limits at least once per 24 ours.
"See special Test Exception 3.10.4 The provisions of Specification 4.0.4 are not applicable provided that this surveillance is performed within 12 hours after starting an idle /
( recirculation loop. j PERRY - . NIT 1 3/4 4-5
Attachment 3 PY-CEI/NRR-1353 L
;, REACTIVITY CONTROL SYSTEMS Page M of X BASES 3/4.1.3 CONTROL RODS The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained. (2) the control rod insertion times are consistent with those used in the safety analyses, and (3) limit the potential effects of the rod drop accident. The ACTION statements pertnit variations from the' basic requirements but at the same time impose morf restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis. ~
Damage within tha ene.;rol rod drive mechanism could be a generic problem, therefore with a control rod immovabie because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods. Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are
, consistent with the SHUT 00VN MARGIN requirements.
( The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem. g gg , The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than , during the limiting power transient analyzed in Chapter 15 of the USAR. This analysis shows that the negative reactivity rates resulting frun the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than g4-OP. The occurrence of scram times longer than those specified should be viewed as an indication of a l systematic proble": with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the re ctor for long periods of time with a potentially serious problem. gg 3g py gg,p ud. The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and vill isolate the reactor coolant system from the containment when required. Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than (D has been analyzed even though control rods with inoperable accumulators may L still be inserted with normal drive water pressure. Operability of the accumu'stor ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor. l PERRY - UNIT 1 B 3/4 1-2 Amendment No. 20
Attachment 3 PY-CE1/NRR-1353 L 3/4.2 POWER DISTRIBUTION LIMITS "" Q, . BASES The specificatit of this section assure that the peak cladding temper-ature following the postulated design basis loss-of-coolant accident will not exceed the 2200*F limit specified in 10 CFR 50.46, 3/4.2.1 AVERAGE PLANAR LP: EAR HEAT GENERATION RATE This specification assures that the peak cladding temperature (PCT) following the postulated design basis Loss-of-Coolant Accident (LOCA) will not exceed the limits specified in 10 CFR 50.46 and that the fuei cesign analysis limits specified in GESTAR-II (Reference 1) will not be exceeded. The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is crimarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod wnicn is equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure depen-dent steady state gap conductance and rod-to-rod local peaking factor. The Tecnnical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor. ( The MAPLHGR limits specified in the CORE CPERATING LIMITS REPORT are { multiplied by the smaller of either the flow dependent MAPLHGR factor (MAPFAC ) 7 or the power dependent MAPLHGR factor (MAPFAC p
) corresponding to existing core flow and power state to assure the adherence to fuel mechanical design bases during the most limiting transient.$MAPFACf 's are determined using the three-g noe k dimensional BWR simulator code to analyze slow flow runout transients, g,.y*f APFAC p 's are generated using the same data base as the MCPR to protect the p core from plant transients other than core flow increases.
The Technical Specification MAPLHGR value is the most limiting composite (ncW Q f the fuel mecnanical design analysis MAPLHGR and the ECCS MAPLHGR. pr.y,f Fuel Mechanical Design Analysis: NRC approved methocs (specified in Reference 1) are used to demonstrate that all fuel rods in a lattice, ocerating at the bounding power history, meet the fuel design limits s;;ecified in Reference 1. This bounding power history is used as the bas is for the fuel design analysis MAPLHGR value. LOCA Analysis: A LOCA analysis is performed in accordance with 10 CFR Part 50 Appendix K to demonstrate that the MAPLHGR values comply with tne ECCS limits specified in 10 CFR 50.46. The analysis is performed for the most limiting break size, break location, and single failure combination for the plant. PERRY - UNIT 1 B 3/4 2-1 Amendment No. /Gh33
Attachment 3 PY-CEI/NRR-1353 L Pagej26_of 38 Insert J _u A variety of PNPP specific Feedvater Controller Failure and Load Rejection with Bypass Failure transient events (vith and without feedvater temperature reduction) together with a large data base of transient event results for other operating plants were used by General Electric to establish the MAPFAC and MAPFAvg limits, with suitable conservatism for operation in the Maximum Exte8ded Operating Vor.ain with and without partial feedvater heating (Reference 2). Insert K
==
For single recirculation loop operation the MAPLUGR limits contained in the CORE
- OPERATING LIMITS REPORT are multiplied by a single loop operation MAPLHGR reduction factor determined each cycle as part of the reload safety analyses. The single loop operation MAPLHGR reduction factor is derived from the LOCA analyses initiated from single loop operational conditions to account for the earlier boiling transition at the limiting fuel assembly node compared to the standard (two loop) LOCA evaluations.
Attachment 3 PY-CEI/NRR-1353 L Page J7 of 3B . h, _ POWER O!STRIBUTION LIMITS BASES A_VERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued) (c.oge ore AATf A)c LiHITS 8,E PeRD Only the most limiting MAPLHGR values are shown in the(Technical Soecifi kcatio1D fioures for multi R1 attice l fuel. When hand calculations are required. tnese Gecnnical Specification 1MAPLHGR figure values for that fuel type are used for allplattices MceAE OPEin R that ATs%bundlM__8M, Lf MIT3 For some GE fuel bundle designs MAPLHGR depends only on bundle type and burnup. Other GE fuel bundles have MAPLHGRs that vary axially depending upon the specific combination of enriched uranium and gadolinia that comprises a fuel bundle cross section at a particular axial node. Each particular comoination of enriched uranium and gadolinia, for these fuel bundle types. . is called a lattice type by GE. These particular fuel bundle types have MAPLHGRs that vary by lattice type (axially) as well as with fuel burnup. Approved MAPLHGR values (li:'iiting values of APLHGR) as a function of fuel and lattice types, and as a function of the averace olanar exposure are provided in the CORE OPERATING LIMITS REPORT. l , ( I l I l l i PERRY - UNIT 1 B 3/4 2-2 Amendment No. Aq33 i-
Attachment 3 PY-CEI/NAR-1353 L Page j L oi 38 R -
, m ,-
Overflow o n to thh p7e 3 if neeke& for earrya ve c of +at f rem previeos oP Mt.XY seveva f p., gg, This page intentionally left blank
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l l PERRY - UNIT I B 3/4 2-3 Amendment No. 20 l ..
i Attachment 3 PY-CEl/NRR-1353 L POWERDISTRIBUTIONJIMITS BASES 3/4.2.2 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at st dy state operating conditions as specified in Specification 3.2.2 are_deri d from the established fuel cladding integrity Safety Limit MCPR 6f 1.073and an analysis of the limiting operational transients. For any abnonnal operating transient analysis evalua-tion with the initial condition of the reactor being at the steady state operating limit. it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2. To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnonnal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduc-tion in CRITICAL POWER RATIO (CPR). The type of transients evaluated are documented in the USAR and Reference 1. The limiting transient yields the largest delta CPR. When added to the Safety Limit MCPR, the recuired operating limit MCPR of Specification 3.2.2 is obtained.fThe power-fica map of Figure S 3/4 2.2-1 defines the analytical basis f or generat1on of the MCPR operating limits. J g3 y Insut L Insert L The evaluation of a given transient begins with the sy WtMT!fa~l ( g,st,,,,) pr.3 ryh arameters shown in USAR Chapter 15 and/or Reference 1, that are input l to a GE-core dynamic behavior transient computer program. The codes used to evaluate these events are described in Reference 1. The purpose of the MCPR 7 and :1CP'i p is to define operating limits at other than rated core flow and power conditions. At less than 100% of rated flow and power the required MCPR is the larger value of the MCPR 7 and MCPR, at the existing core flow and power state. The MCPR 7 s are established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit seguirement can be assured. Insert H The MCPR figure contained in the CORE OPERATING LIMITS REPORT also b l l, f 5 I4eflects the 9equired MCPR values resulting from the analysis perfonned to ustify operation with the feedwater temperature ranging from 420*F to 320*F at 100". RATED THERMAL POWER steady state conditions, and also beyond the end of cycle with the feedwater temperature ranging from 420*F and 250*F. The MCPR 7 s were calculated such that for the maximum core flow rate and the corresponding THtRMAL POWER along a conservative steep generic power flow I control line, the limiting bundle's relative power was adjusted until the MCPR was sligntly above the Safety Limit. Using this relative bundle power, the i MCPRs were calculated at different points along this conservative steep power ( flow control line corresponding to different core flows. at a given point of core flow is defined as MCPR . The calculated MCPR 7 I PERRY - UNIT 1 3 3/4 2-4 Amendment No. /A/, 33
Attachment 3 PY-CE1/NRR-1353 L Page ,$d of ,}3_, The Maximum Extended Operating Domain (MEOD) power-flow map of Figure B 3/4-2.2-1 portrays the allovable operating region for two recirculation loop operation. A similar power-flow map for single recirculation loop operation is described in USAR Appendix 15F. The analytical basis for generation of the MCPR operating limits is described below. Insert M
~ :-
A variety of PNPP specific Feedvater Controller Failure and Load Rejection with Bypass Failure transient events (with and without feedvater temperature reduction) together with a large data base of transient event results for other operating plants were used by General Electric to establish the MCPR and MCPR limits, with suitableconservatismforoperationintheMEODwithandvkthoutparYialfeedvater heating (Reference 2). For single recirculation loop operation the MCPR Safety Limit is increased by 0.01 as described in Section 2.0 of the BASES. No increase in the rated Operating Limit MCPR and no changes in the flow and power dependent MCPR limits are required for single recirculation loop operation because the limiting operational transients analyzed indicated that there is more than enough MCPR margin to compensate for the increase in the MCPR Safety Limit (Reference 3). l 1 I
O O Attachment 3 PY-CEI/NRR-1353 L I POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) The MCPR p s are established to protect the core from plant transients other than core flow increases, including the localized event such as rod withdrawal error. The MCPRp s were calculated based upon the most limiting transient at the given core power level. For core power less than or equal to 40% of RATED THERMAL POWER, where the EOC-RPT and the reactor scrams on turbine stop valve closure and turbine control valve fast closure are bypassed, separate sets of MRPR p limits are provided for high and low core flows to account for the sig-nificant sensitivity to initial core flows. For core power ebove 40% of RATED THERMAL POWER, bounding power dependent MCPR limits were developed. At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indi-cates that the resulting MCPR value is in excess of requirements by a considerable margin. During initial start up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump
; speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. The daily requirener.t for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The require-ment for calculating MCPR when a limiting control red pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.
3/4.2.3 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.
References:
- 1. GESTAR II, General Electric Standard Application for Reactor Fuel, AEDE-240ll-P-A, (latest approved revision).
- 2. T. C. Lee, " Technical Specification Operating Limits with the Elimination of the APRM Trip Setdovn Requirements - Perry Nuclear Power Plants." April 1985 (NEDM-309e3).
- 3. USAR Appendix 15F, PNPP Single Loop Operation Analysis.
3 !Y - UNIT 1 B 3/4 2-5 Amencment No. 20
Attachment 3 PY-CE'L/NRR-1353 L INSTRUMENTATION Pap A of 38 BASES 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel. Operation with a trip set less conservative tbn its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the differ-ence between each Trip Setpoint and the Allowable Value is an allowance for instrument drif t specifically allocated for each trip in the safety analyses. 3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/4.2 Power Distribution Limits. The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the differ-ente between each Trip Setpoint and the Allowable Value is an allowance for c,di uclnstrument drift specifically allocated for each trio in the safety analyses. N " b *d5 bSh3 k " be 3/4.3.7 HONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION HONITORING INSTRUMENTATION
'[ [ * ,
The OPERABILITY of the radiation monitoring instrumentation ensures that; Sed 6 J,2,1 (1) the radiation levels are continually measured in the areas served by the It e m ,g , j individual channels; (2) the alars or automatic action is initiated when the / 1 radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with 10 CFR Part 50, Appendix A, General Design Criteria 19, 41, 60, 61, 63 and 64. 3.4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for the unit. This instrumentation is consistent with the recommendations of Regulatory Guide 1.12 " Instrumentation for Earthquakes", April 1974. 3/4.3.7.3 METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the meteorological monitoring instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to eval-uate the need for initiating protective measures to protect the health and safety of the public. This instrumentation is consistent with the recommenda-tions of Regulatory Guide 1.23 "Onsite Meteorological Programs," February,1972. PERRY - UNIT 1 B 3/4 3-4
+ e Attachment 3 *~
PY-CEI/NRR-1353 L b bCe f WIfb Page R of 3R g.g g g. 3/4.4 RE R COOLANT SYSTEM pqcs, BASES 3/4.4.1 RECIR TION SYSTEM Operation wi ne reactor core coolant recirculation loop inope le is prohibited until an aluation of the performance of the ECCS durin e loop opetation has been p ormed, evaluated and determined to be accep e. An inoperable jet p is not, in itself, a sufficient rea to declare a recirculation loop ino able, but it does, in case of a des' basis-accident, increase the bl own area and reduce the capabili f reflooding the core; thus, the requir nt for shutoown of the facilit th a jet pump inoperable. Jet pump failu can be detected by monitorin et pump perfor-mance on a prescribed schedu for significant degradati Recirculation loop flow mismatch limits are compliance with ECCS L analysis design criteria. The limits will ensu an adequate core flo castdown from either recirculation loop following a L In order to prevent undue stre on the vess ozzles and bottom head region, the recirculation loop tempe ures shal within 50*F of each other prior to startup of an idle loop. Th oop tem ature (nust also be within 50 F of the reactor pressure vessel co t te rature to prevent thermal shock to the recirculation pump and reci il on nozzles. Since the coolant in the bottom of the vessel is at a lower te ture than the coolant in the upper ( regions of the core, undue stress on the v el would result if the temperature difference was greater than 100 F. The objective of GE BWR plant an el de n is to provide stable opera-tion with margin over the normal ope ing doma However, at the high power / low flow corner of the operating d n, a small obability of limit cycle neutron flux oscillations exists ending on com tions of operating condi-tions (e.g., rod pattern, power pe). To provide surance that neutron flux limit cycle oscillations are d ted and suppresse PRM and LPRM neutron flux noise levels should be m tored while operating this region. Stability tests at op ing BWRs were reviewed to ermine a generic region of the power / flow in which surveillance of neu n flux noise levels should be perfor A conservative decay ratio of 6 was cncsen as the bases for determin' the generic region for surveillan to account for the plant to plant va bility of decay ratio with core and designs. This generic region been determined to correspond to a cor low of less than or equal to 4 of rated core flow and a thermal power gre r than that specified in Fig 3.4.1.1-1 (Reference 1). Plant spe ic calculations can be performed to determine an a icable region for m oring neutron flux noise levels. In this case the d e of conservatis an be reduced since plant to plant variability would be eliminate In this case, adequate margin will be assured by monitorin he region w . has a decay ratio greater than or equal to 0.8. PERRY - UNIT 1 B 3/4 4-1
"c.. ,
i g g. g Attachment 3 PY-CEI/NRR-1353 L 3/4.4 REACTOR COOLANT SYSTEM w& n age 3 t of R M1 Nf b dm. w 3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor _ core coolant recirculation loop out of service has been' evaluated (Reference 3) and found to remain within design limits and_ safety margins provided certain limits and setpoints are modified. Single loop operation is permitted at power levels up to 2500 Megavatts-thermal (MVt) (slightly less than 70% of RATED TliERNAL POVER), if the MCPR Tuel Cladding Integrity Safety Limit is increased as noted by Specification 2.1.2, the APRM flov biased scram and control rod block setpoints (or APRM gains) are adjusted as noted in Tables 2.2.1-1, 4.3.1.1-1 and 3.3.6-2, respectively, and the MAPLHGR limits are decreased in accordance with the value specified in the CORE OPERATING LIMITS REPORT. A-time period of 8 hours is allowed to make these adjustments following establishment of single loop operation since the need for single loop eperation of ten cannot be anticipated. Additionally, a limitation on the volumetric drive flov of the operating recirculation-loop (during SLO) is-imposed to exclude the possibility of excessive _ core internals vibration. Recirculation loop drive flov is the discharge flow-from the recirculation pumps. Recirculation loop jet pump flow is the summation of all the individual jet pump flows for a particular recirculation loop. Total core flow for two recirculation loop operation is the sum of the recirculation loop jet pump flows for both loops. Total core flow for single recirculation loop operation is the recirculation loop jet pump flow for the operating loop minus the reverse flow through the non-operating loop. Rated core flow as used in the Recirculation System Specifications corresponds to the rated (100%) core flow value for two recirculation loop operation (104 Ulb/hr). To prevent potential control system oscillations from occurring in the recirculation flow control system (and to eliminate'the need for -- flov control system failure analyses), the operating mode of the recirculation flow control system is restricted to the Loop Manual (Position Control) mode for single loop operation.
-The surveillance on differential temperatures belov 30% THERNAL POVER or 50% of rated (two loop) core flow is to prevent undue thermal stress on vessel nozzles, recirculation pump, and vessel bottom head during a power or flov increase during extended operation in the single recirculation loop mode, similar'to the restrictions desc-ibed below for idle recirculation loop startup. The current 50% of rated core flov value has'been determined by analysis and operating experience at e other GE BVRs to provide margin to the onset of stratification. To provide operational flexibility a threshold core flov (belov vhich thermal stratification might occur) may be determined during testing in SLO.and a new limit established with appropriate margin to this value.
PERRY - UNIT 1 B 3/4 4-1
A Akk neW or r ew'deg Attachment 3 REACTOR COOLANT SYSTEM PY-CEI/NRR-1353 L YEd b daw. Page 15' of Jg EM %# 3/4.4.1 RECIRCULATION SYSTEM (Continued) An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design-basis accident, increase the blovdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump tailure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation. During single loop operation the jet pump operability surveillance is only performed for the jet pumps on the operating recirculation loop, as the loads on the jet pumps on the inactive loop are expected to be very lov due to the lov flov in the reverse direction through them. This has been demonstrated through operating experience at other GE BVRs. The jet pumps in the non-operating recirculation loop during SLO are considered operable based on this lov expected loading, acceptable surveillance results obtained during two recirculation loop operation prior to entering SLO, or by visual inspection of the jet pumps during refueling / shutdowns. Upon startup of an idle recirculation loop when THERHAL POWER is greater than 25% of RATED THERHAL POVER, the specified jet pump surveillances are required to be performed for the previously idle loop vithin the 24 hour surveillance interval specified in Surveillance Requirement 4.4.1.2. Significant degradation is indicated if more than one of the three specified surveillance requirements performed confirms unacceptable deviations from established patterns or relationships. The surveillances, including the associated acceptance criteria, are in accordance with General Electric Service Ir. formation Letter No. 330, the recommendations of which are considered acceptable for verifying jet pump operability according to NUREG/CR-3052, " Closeout of IE Bulletin 80-07: BVR Jet Pump Assembly Failure." Performance of the specified surveillance requirements, however, is not required when thermal power is less than 25% RATED THERHAL POVER because flow oscillations and jet pump noise precludes collection of repeatable data during lov flow conditions, approaching the threshold response of the associated flov instrumentation. Jet pump operability is considered acceptable prior to startup of the plant due to acceptable results obtained during the past cycle, or by visual inspection of the jet pumps. Initial data collection vill be performed to establish the specified relationships during post-refuel performance testing for a new operating cycle, or upon first entry into single 1;op operation during an operating cycle. This satisfies the Surveillance Requirements of Specification 3.4.1.2, since taking the data establishes the relationships for that cycle or mode of operation, t.nd there are no valid prior relationships to compare against. Jet PERRY - UNIT 1 B 3/4 4-la
- a m ew -
0"
. Attachment 3 Te w N ile H PY-CEl/NRR-1353 L REACTOR COOLANT SYSTEM b t l> cleg ) Page_g.,of,1{
gg 7 m/ 3/4.4.1 RECIRCULATION SYSTEM (Continued) pump operability following a change from tvo to single loop operation or vice versa is considered acceptable for the duration needed to establish the r.av specified relationships based on having met the relationships for the former mode of operation. During two recirculation loop operation, recirculation loop jet pump flov mismatch limits are specified to ensure compliance with the two loop ECCS LOCA analysis design criteria. The limits vill ensure an adequate core flow coastdown from either recirculation loop following a LOCA. In the case where the flow mismatch limits cannot be maintained during two recirculation loop operation, continued operation is permitted in the single recirculation loop mode. A 4.0.4 exception is provided to permit the startup of an idle recirculation loop when in single loop operation (to return to two loop operation). In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50 F of each other prior to startup of an idle recirculation loop. The loop temperature must else be within 50 F cf the reacter pressure ves:el coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Sudden equalization of a tamperature difference greater than 100 F between the reactor vessel bottom head coolant and the coolant in the reactor vessel by increasing core flow rate could cause undue stress in the reactor vessel bottom head. The objective of GE BVR plant and fuel design is to provide stable operation with margin over the normal operating domain. However, at the high power / low flow corner of the operating domain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating conditions (e.g., rod patterr, power shape). To provide assurance that neutron flux limit cycle oscillations are detected and suppressed, APRM and LPRH neutron flux noise levels should be monitored while operating in this region. PERRY - UNIT 1 B 3/4 4-lb
- e Attachment 3 MW U PY-CEI/NRR-1353 L REACTOR COOLANT SYSTEM EM '
3/4.4.1 RECIRCULATION SYSTEM (Continued) Stability tests at operating BVRs were reviewed to determine a generic region of the pover/ flow map in which surveillance of neutron flux noise levels should be performed. A conservative decay ratio of 0.6 was chosen as the bases for determining the generic region for surveillance to account for the plant to plant variability of decay ratio with core and fuel designs. This generic region has been determined to correspond to a core flov of less than or equal to 45% of rated core flow and a thermal power greater than the 80% rodline in accordance with SIL-380 (Reference 1) and Supplement I to NRC Bulletin 88-07 (Reference 4). A generic 80% rodline is represented in Figure 3.4.1.1-1 (the actual 80% rodline varies slightly cycle to cycle with changes in fuel type, core nydraulic resistance etc.). Plant specific calculations can be performed to determine an applicable region for monitoring neutron flux noise levels. In this case the degree of conservatism ca.n be reduced since plant to plant variability would be eliminated. In this case, adequate margin vill be assured by moaltoring the regicn which has a decay ratio greater than or equal to 0.8. mA-wm Hove tu + e. , page g gjy y_, bad +. this pqe to rrued over[ low o n Ya P'V B 3 /y- y - 3 Of n eeA c4), W PERRY - UNIT 1 B 3/4 4 -1c
. i Attachment 3 PY -CEI/NRR-1353 L Page 38 of J,,,
REACTOR COOLANT SYSTEM f - BASES RECIRCULATION SYSTEM (Continued) Neutron flux noise limits are also established to ensure early detection of limit cycle neutron flux oscillations. BWR cores typically operate with neutron flux noise caused by randos boiling ano flow noise. Typical neutron flux noise levels of 1-12% of rated power (peak-to peak) have been reported for the range of low to high recirculation loop flow during both single and dual recirculation loop operation, Neutron flux noise levels which signifi-cantly bound these values are considered in the thermal / mechanical design of GE BWR fuel and are found to be of negligible consequence (Reference 2). In addition, stability tests at operating BWRs have demonstrated that when stabil-ity related neutron flux limit cycle oscillations occur they result in peak-to-peak neutron flux limit cycles of 5-10 times the typical values. Therefore, actions taken to reduce neutron flux noise levels exceeding three (3) times the typical value are sufficient to ensure early detection of limit cycle neutron flux oscillations. l Typically, neutron flux noise levels show a gradual increase in absolute , magnitude as core flow is increased (constant control rod pattern) with two I reactor recirculation loops in operation. Therefore, the baseline neutron flux noise level obtained at a specific core flow can be applied over a range of core flows. To maintain a reasonable variation between the low flow and high flow ends of the flow range, the range over which a specific baseline is ( applied should not exceed 20% of rated core flow with two recirculation loops l In operation. Data from tests and operating plants indicate that a range of 20% of rated core flow will result in approximately a 50% increase in neutron flux noise level during operation with two recirculation loops. Baseline data should be taken near the maximum rod line at which the majority of operation will occur. However, baseline data taken at lower rod lines (i.e., lower power) will result in a conservative value since the neutron flux noise level is proportional to the power level at a given core flow. References (1) "BWR Core Thermal-Hydraulic Stability" Service Information Letter 380, Revision 1 February 1984. (2) G. A. Watford, " Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria," December 1982 (NEDE 22277-P). 3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves (SRV) operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 13 OPERABLE l safety-relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient. Any combination of l 6 SRVs operating in the relief mode and 7 SRVs operating in the safety mode s ijLgceptable. - . (3) U S A R Apf e ,, d ,*x Js F, PNPP S i e,y le L.or o f e r.fien /in.l y s;5, 'N s ('O N8C Sull< tin No. 88-07 Surplemoo+ l lPa we r Osdlldiern in Beilio, PERRY - UNIT i B 3/4 4-2 mm gegg, gg
ATTACHMDir 3A RETTPED SPECIFICATIONS RECIRCULATION LOOPS - 3.4.1.1 JET PUMPS - 3.4.1.2
< e 3/4.4 REACTOR COOLANT SYSTEM Attachment 3A PY-CEI/NRR-1353 L 3/4.4.1 RECIRCULATION SYSTEM Page / of 6 RECIRCULATION LOOPS LIHITING CnJDITION FOR OPEiii10N 3.4.1.1 The reactor coolant system recirculation loop (s) shall be in operation with the total core flov greater than or equal to 45% of rated core flow, or THERMAL POVER less than or equal to the limit specified in Figure 3.4.1.1-1 and either:
- a. Two recirculation loops operating, or
- b. A single recirculation loop operating with the following limits and conditions:
- 1. a) The MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit adjusted for single recirculation loop operation per Specification 2.1.2, and b) The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits adjusted for single recirculation loop operation per the CORE OPERATING LIMITS REPORT in accordance with Specification 3.2.1, and
.c) The Average Power Range Monitor (APRH) Scram and Rod Block Trip Setpoint and Allovable Value equations adjusted to those valygs applicable for single recirculation loop operation per Specifications 2.2.1 and 3.3.6, and 1
j 2. A volumetric recirculation loop drive flow less than or equal to 48,500 gpm, and
- 3. The recirculation flow control system in the Loop Manual (Position Control) mode, and I.
j 4. THERMAL POWER less than or equal to 2500 Megavatts-thermal. APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2 . t ACTION:
- a. Upon initial entry into single loop operation, adjug}ments to the limity,and setpoints of Specifications 2.1.2, 2.2.1 , 3.2.1, and 3.3.6 shall be implemented within 6 hours, or declare the associated equipment inoperable (or declare the associated limits to be "not satisfied"), and take the ACTIONS required by the applicable specifications.
- See Special Test Exception 3.10.4.
** To functionally implement these protective functions during entry into single loop operation, APRM gain adjustments may be made in lieu of adjusting the APRM Scram and Rod Block Flov Biased Setpoints for an interim period of 72 hours.
PERRY - UNIT 1 3/4 4-1 l l
-_ ._, . _ _ . ~ ,_ ____- _ . . _ . . . . . _ . .
Attachment 3A
~
REACTOR COOLANT SYSTEM f gf LIMITING CQFDITION_E0ll.0fERATION (C2ntinut.d)
- b. During single loop operation, with the volumetric recirculation loop drive-flov greater than the above limit, immediately-initiate corrective action to reduce flov to less than or equal to the above limit within 1 hour.
- c. During single loop operation, with the recirculation flow control system not in the Loop Manual mode, immediately initiate corrective action-to place the recirculation flow control system in the Loop Manual mode within 1 hour.
- d. During single loop operation, with THERMAL POVER greater than the above limit, immediately initiate corrective action to reduce
. THERMAL POVER to less than or equal to the above limit within 1 hour.
- e. During single loop operation, with either:
- 1. THERMAL POVER < 30% of RATED THERMAL POVER and temperature differences exceeding the limits in Surveillance Requirement 4.4.1.1.4, or I
- 2. recirculation loop jet pump flow in the operating loop < 50%
of rated (two loop) core flov and temperature differences exceeding the limits in Surveillance Requirement 4.4.1.1.4, suspend THERMAL POVER and recirculation loop flov increases.
- f. Vith no_ reactor coolant. system recirculation loops in operation, immediately initiate action to reduce THERMAL POVER to less than or equal to the limit specified in Figure 3.4.1.1-1 within 2 hours and initiate measures to place the unit in at least STARTUP vithin 6 hours and in HOT SHUTDOVN within the next 6 hours, i
# A conservative initial value. A lover recirculation loop jet pump flov value may be determined during SLO and submitted for approval, based upon l the threshold flow which vill sweep the cold water from the vessel bottom l head preventing stratification.
l l PERRY - UNIT 1 3/4 4-2 l L i l _ . _ . . . , , - . _ . _ . , , ~ -
+ .
Attachmsnt 3A PY-CEI/NRR-1353 L REACTOR COLLANT SYSTEH Page 3 of f LIHlIIH(LCQ@jJLON__FOR OPEP 4 TlpN (continited)
- g. O th one or two reactor coolant system recirculation loops in operation, and total core flov less than 45% of rated core flow and THERHAL POVER greater than the limit specified in Figure 3.4.1.1-1:
- 1. Determine the'APRH and LPRHN noise levels (Surveillance 4.4.1.1.2):
a) At least once per 8 hours, and b) Vithin 30 minutes after the completion of a THERHAL POVER increase of at least 5% of RATED THERHAL POVER.
- 2. Vith the APRH or LPRH N neutron flux noise levels greater.than three times their established baseline noise levels, immediately initiate corrective action to restore the noise levels to within the required limits within 2 hours by increasing core flov to greater than 45% of rated core flov or by reducing THERHAL POVER to less than or equal to the limit specified in Figure 3.4.1.1-1.
jiURVEILLANCE _ _ REQUIREMENTS 4.4.1.1.1 Each reactor coolant system recirculation loop flow control valve shall be demonstrated OPERABLE at least once per 18 months by:
- a. Verifying that the control valve fails "as is" on loss of hydraulic pressure at the hydraulic control unit, and
- b. Verifying that the average rate of control valve movement is:
- 1. Less than or equal to 11% of stroke per second opening, and
- 2. Less than or equal to 11% of stroke per second closing.
4.4.1.1.2 Establish a baseline APRM and LPRMN' neutron flux noise value
. within the regions for which monitoring is required (Specification 3.4.1.1, ACTION g) within 2 hours of entering the region for which monitoring is required unless baselining has previously been performed in the region since the last refueling outage, l.
l-H Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRH string in the center of the core should be monitored. PERRY - UNIT 1 3/4 4-2a
l a Attachment 3A PY-CEl/NRR-1353 L REACTOR COOLANT-SYSTEM Page W of G i EURVEILLANCE REQUIREMENTS (Continued) 4.4.1.1.3 Initially, within 1 hour upon entry into single recirculation. loop-operation and once per 12 hours thereafter, verify that
- a. The volumetric recirculation loop drive flow of the operating loop is less than or equal to the limit stated in Specification 3.4.1.1.b.2, and
- b. The recirculation flow control system for the operating loop is in i the Loop Hanual (Position Centrol) mode, and l i
- c. .THERHAL POVER-is less than or equal to the limit stated in ;
Specification 3.4.1.1.b.4. 4.4.1.1.4 Vith one reactor coolant system recirculation loop not in operation, and either THERHAL POVER less than or equal to 30% of RATED THERHAL POVER or the re pump flow in the operating loop less than oc equal to 50%girculation looploop) of rated (two jet core flow, verify within 15 minutes prior to an increase in THERHAL POVER or recirculation loop jet pump flow that the following differential temperature requirements are mets
- a. < 100 F between reactor vessel steam space coolant and bottom head drain line coolant, and
- b. f 50 F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and
- c. f 50 F between the reactor coolant within the loop not in operation and the operating loop.
The differential temperature requirement of 4.4.1.1.4.a does not apply when the reactor pressiire vessel is belov 25 psig. The differential temperature requirements of 4.4.1.1.4.b and c do not apply when there is no flow through the loop not in operation due to either one or both the loop suction / discharge valve (s) being closed, it A conservative initial value. A lower recirculation loop jet pump flow value may be determined during SLO and submitted for approval, based upon l the threshold flow which vill sweep the cold water from the vessel bottom l head preventing stratification. PERRY - UNIT 1 3/4 4-2b
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REACTOR COOLANT SYSTEM Attachment 3A PY-CEI/NRR-1353 L JET PUMPS of 6_, Fece 6 MLIlFG COMil10F FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE. APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2 ACTION Vith one or more jet pumps inoperable, be in at least DOT SHUTDOVN within 12 hours. SUFMILLANQ1LM0VlEENIS 4.4.1.2 Each jet pump in an operating recirculation loop shall be demonstrated OPERABLE at least once per 24 hours when THERMAL POVER is greater than 25% of RATED THER!:AL POVER, by deteraining recirculation loop drive flow, recirculation loop jet pump flov and dif fuser-to-lover plenum dif ferential pressure (or flow), n or each jet pump and verifying that ao two of the following conditior.s occurt
- a. The indicated recircylation loop drive flow differs by more than 10%
from the established flow control valve position-drive flow characteristics,
- b. The indicated recirculation Igop(s) jet pump flow dif fers by more than 10% from the established recirculation loop (s) jet pump flov value derived from recirculation loop drive flow measurements,
- c. T w indicated jet pump dif f user-to-lover plenum dif ferential pressure (Og jet pump flov) of any irdividual jet pump dif fers f rom established patterns by more than 20% (or 10% for flov).
The provisians of Specification 4.0.- are not applicable provided that this curveillance is performed within 24 hours after exceeding 25% of RATED THERMAL POWER.
- Data shall be recorded following each refueling outage or upon first entry into single loop operation during an operating cycle, in order to establish the specified relationships for that cycle / mode of operation.
Comparisons of the actual data shall commence upon the establishment of the specified relationships for that cycle / mode of operation. PERRY - UNIT 1 3/4 4-4
ATTACllMENT 4 CilANGES TO T!!B CORE OPERATING L1!!!TS REPORT TO IMP 12 MENT SING 12 RECIRCULATION 100P OPf3tATION
o e
, i Attachment 4 OM18: PDB-P0001 l PY-CE1/NRR-1353 L y,,,, y P a p __ [_,, o f t__ g,y , , y( g l INTRODUCTION AND REFERENCES )
INTRODUCTION , This Core Operating Limits Report for PNPP Unit 1 Cycle 3 is prepared in accordanc,e with the requirements of FNPP Technical Specification 6.9.1.9. The core operating limits presented vere developed using NRC-spproved methods (keference 2). Results from the reload analyses for the General Electric fuel in PNFP Unit 1 for Cycle 3 are documented in References 3, . 4 and 5. The cycle-specific core operating limits for the following FNPP Unit 1 Technical Specifications are included in this reports i 1. Average Planar Linear Beat Generation Rate (APLBCR) Limits for each fuel / lattice type, including the power and flow dependent MAPPAC , curves, (Technical sneg1Lig.a.11pn 3/4.2.1) Wh ne sin 3fe le*P M APLHGR redvs.h'en fa.iep
- 2. Minimus Crit cal rever Ratto operating Limit incliiiIing the power and flow dependent MCPR curves. (Technical Specification 3/4.2.2)
- 3. Linear Beat Generation Rate (L8GR) Limit for each fuel type.
(Technical Specification 3/4.2.3) 3 mir R NC S l l 1. J.R. Rail (USNRC) to M.D. Lyster (CEI), Amendment No. 33 to Pacility l Operating License No. NPP-58, September 13, 1990.
- 2. " General Electric Standard Application for Reactor Fuel-GESTAR II,"
NEDE-24011-P-A-9 and NEDE-24011-P-A-9-US (US Supplement).
- 3. " Supplemental Reload Licensing submittal for the Perry Nuclear Power Plant Unit 1, Reload 2, Cycle 3," CE Document 23A6492 Rev. 0 (September 1990).
- 4. ' Supplement 1 to the Supplemental Reload Licensing submittal for the Perry Nuclear Power Plant Unit 1, Reload 1, Cycle 2," GE Document 23A5948AA Rev. 0 (October 1988).
- 5. ":lupplement 1 to the Supplemental Reload Licensing Submittal for the furry Nuclear Power Plant Unit 1, Reisad 2, C/cle 3 " GE Document 23A6492AA Rev. 0 (September 1990).
- 6. Perry Nuclear Power Plant Updated Safety Analysis Report, Unit 1, Appendix 158-Reload Safety Analysis.
CTCLE 3 CORE OPERATING PERRY INIT 1 LIMITS KRPORT
O e 1 Attachment 4 OM18: l'Dil 170001 PY-CEI/NRR-1353 L l' age : 4 Page 4 of P Itev. I 1 PNFP No 1097 1.1 1.0 -
, e
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0.6 ; F l O 5 -- 0 20 40 60 80 100 120 CORE FLOW (% RATED), F FLOW DEPENDENT MAPLIIGH FAC10lt (MAPFAC g ) FIGURE 3.2.11 CYCLE 3 CORE OPEllATING 18EltlW UNIT 1 LIMITS REPORT
O o Attachment 4 OM18: PDil F0001 PY-CE1/NRR-1353 L Page : 5 Page 1 of j Iley. I 1 PNPP Na M95 1.1 __ _, 21 i
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, MAPFAC p- 1.0 + 0.0052 (P 100) ;
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. snm s .
MAPPACp- 0.6 + 0.002 (P 40) :
- 25 % $ P g 40% ; Core flow F > 50%
l 0.5 l O 20 40 60 80 100 120 l CORE TIIERMAL POWER (% RATED), P POWER DEPENDENT MAPLIIGR FAC1DR (MAPFAC p) FIGURE 3.2.12 CYCLE 3 CollE OPERATING l PERRY UNIT 1 LIMITS REPORT
O r Attachment 4 PY-CEI/NRR-1353 L O M is: PDil-F0001 Page 't of '/- Page : 14 rNrr Na 9147 Rev. : 0 2.3 - i TilERMAL POWER 25%~ < P ~< 40% ---- 2.2 - f CORE FLOW > 50% ::-- w i
' 4aK 2.1 I
L TilERMAL POWER 25%~ < P ~< 40% 2.0 - - -;
- i. m. CORE FLOW- < 50%
'L w
1.9 a PERMISSIIILE REGION OF OPERATION ---- - c: 1.7 L Ti!ERMAL POWER 40% < P < 70% :: N
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N , __l _ TilERMAL POWER P > 70% _ 1.4 : BEFORE AND AT END OF CYCLE: ,
- All core average exposures and v ,'
- 0 f 6 Of 100>F and [
1.3 -; Core Flow $ 105% \ x
~
AFTER END OF CYCLE: T Core average exposure > EOCE e. l g ,3 _: '
- and Of 6kl70*F and
; Core flow $ 105% _
1.1 -; f . Dur.ng steady state operation with only
- one recirculation I ump operating, normal feedwater heating shallin maintained. 1 1.0 1
y e p i j 0 20 4G 60 80 100 120 CORE THER51AL POWER (% RATED), P POWr.% DEPENDENT MCPR FAC'IOR (MCPR p) FIGURF 3.2.2 2 CYCLE 3 CORE OP9 RATING PERRY - UNIT 1 LIMITS 'EPORT}}