ML17156A126
| ML17156A126 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 04/11/1985 |
| From: | PENNSYLVANIA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17156A125 | List: |
| References | |
| NUDOCS 8504180550 | |
| Download: ML17156A126 (60) | |
Text
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER Low Pressure or Low Flow
- 2. l. 1 THERMAL POWER shall not exceed 25K of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than.
10K of rated flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 25K of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than lOX of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requi.rements of Specification 6.7. l.
THERMAL POWER Hi h Pressure and Hi h Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10K of rated flow.
, APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION: I With MCPR less than 1.06 and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than lOX of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7. l.
REACTOR COOLANT SYSTEM PRESSURE
- 2. l. 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 SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7. 1.
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TABLE 2. 2 ~ 1- 1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS AD ALLOMABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES
- l. Intermediate Range Monitor, Neutron Fl ux-Hi gh < 120/125. divisions < 122/125 divisions of full scale of ful scale 1
- 2. Average Power Range Monitor:
- a. Neutron Fl u>(-Upscal e, Setdown < 15X of RATED THERMAL POWER < 20X o f RATED TIIERMAL POMER
- b. Flow Biased Simulated Thermal Power-Upscal e
- 1) Flow Biased < 0. 58 M+59 X, Wi th < 0. 58 M<62X, wi th a maximum of a maximum of
- 2) High Flow Clamped < 1 13 5X of RATED < 1 15. 5X o f RATED THERMAL POWER TIIERMAL POWER c, Neutron Flux-Upscale < 1 18X of RATED THERMAL POWER < 120X of RATED THERMAL POWER
- d. Inoperative'.
HA NA Reactor Vessel Steam Dome Pressure - High < 1037 ps i g < l057 psi g
- 4. Reactor Vessel Mater Level - Low, Level 3 > 13. 0 inches above > I 1 . 5 inches aiiove instrument zero* Tns trumen t zero
- 5. Main Steam Line Isolation Valve - Closure < 10X closed < 1 1X closed
- 6. Main Steam Line Radiation - High < 3. 0 x full power < .'3.6 x foll po background background t
- 7. Drywel 1 Pressure - Hi gh < 1. 72 psig < Il. 88 ps ig
- 8. Scram Discharge Vol ume "Water Level - High < 88 gal lons < '88 gallons Ib Turbine Stop Valve - Closure I
- 9. < 5 ~ 5X closed <,7X cl osed B
- 10. Turbine .Control Valve Fast Closure, Trip Oi 1 Pressure Low > 500 ps ig > 460 ps 1 1. Reactor Mode Swi tch Shutdown Posi on ti HA i'A
- 12. Manual Scram HA Se.-. Bases Figure 8 3/4 3- 1 .
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2.1 SAFETY LIMITS.
BASES
- 2. 0 INTRODUCTION 4
The fuel cladding, reactor pressure vessel and primary system piping
. are the principal barriers to the release of radioactive materials to the c4 environs. Safety Limits. are established to protect the integr'ity of these barriers during normal plant operations and anticipated transients. The integrity Safety Limit is set such that no fuel damage is calculated fuel'ladding
~c q to occur if the limit is not violated. Because fuel damage is not directly observable a ste -back approach is used to establish a Safet Limit such that g the MCPR is not less than . . MCPR greater than . represents a conser-vative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the'radioactive
,c 4) barrier is related materials from the environs. The integrity of this cladding to its relative freedom from perforations or cracking.
Al.though some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally V
0 cumulative and continuously measurable. Fuel cladding perforations, however, Q
can result from thermal stresses which occur from reactor operation signifi-
~
- l. cantly above design conditions and the Limiting Safety System Settings. While
+o
,~ fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than i,ncremental 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. l. 1 THERMAL POWER Low Pressure or Low Flow XM-5 The use of the SBft. correlation is not valid for all critical power calculations at pressures below 785 psig or 'core flows less than l(C of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation'ead, 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 10~ lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3..5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lbs/hr. Full scale ATLAS test data taken at pressures from 14. 7 psia to 800 psia indicate that the fuel assembly criticaT power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than SO% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25K of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
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hei &-e ~is T~ai.e Bases Table S2. 1.2-1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL" CLADDING SAFETY LIMIT" Standard Deviation ua tit I of,Point Feedwa r Flow /i.76 Feedwater Temperature 0. 76 Reactor Pres ure 0.5 Core Inlet Temperature / 0. 2.
/
Core Total Flow 2.5 Channel Flow Area '
3.0 Friction Factor Multiplier 10. 0 Channel Friction Factor Mu1 tipl i er 5.0 TIP Readings 6.3 R Factor 1.5
/
Critical Powe~ 3.6
\
e uncertainty analysis used to establish the core wide Saf ty Limit MCPR is b ed on the assumption of quadrant power symmetry for the rea tor core.
SUS(UEHANNA - UNIT 1 B 2-3
3/4. 2 POWER DISTR IBUTIO IMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION QQ g~e,l ~ ftgg/t&E BUAJDLE GXPoSUAG Nor 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type G~ Suet of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2. 1-1, 3.2. 1-2, and 3.2. 1-3.+
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or A f IIATEII IEEAIAL PAlfEA.
ACTION:
With an APLHGR exceeding the limits of Figure 3.2. 1-1, 3.2. 1-2, or 3. 2. 1-3, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> s.
SURVEILLANCE RE UIREMENTS 4.2.1 All APQGRs shall be verified to be equal to or less than the limits determined from Figures 3.2. 1-1, 3.2. 1-2, and 3, 2. 1-3:
I
- a. At least o nc e p er 24 h o urs, A
- b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15 of RATED THERMAL POWER, and c Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL R00 PATTERN for APLHGR.
- d. The provisions of Specification 4.0.4 are not applicable.
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- UNIT 1 I'US(UEHANNA 3/4 2-1 Amendment No. 29
POWER DISTRIBUTION LIMITS I
3/4. 2. 2 APRM SETPOINTS
.LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships:
Tri Set oint Allowable Value S < 0.58W + 59K)T, . S < (0.58W + 62K)T
< (0.58W + 50K)T SRB
< (0.58W + 53K)T SRB where: 'S and SRB are in percent of RATED THERMAL POWER, W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million lbs/hr,
~
T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER (GE fuel) divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY. I is
< always less than or equal to 1.0.
7 Esse encl)"- f.o APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25K of RATED THERMAL POWER.
ACTION:
With the APRM flow bi'ased simulated thermal power upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or S B, as above determined, initiate codrective action within 15 minutes and adjust 3 and/or S to be consistent with the Trip Setpoint value within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL PSER to less than 25Sl of RATED THERMAL. POHER within the next S hours.
SURVEILLANCE RE UIREMENTS 4.2.2 The FRTP and the MFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal power-upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:
- a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
- b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15K of RATED THERMAL POWER, and
- c. Initially and at 1'east once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with MFLPD greater than or equal to FRTP.
- d. The provisions of Specification 4.0.4 are not applicable:
With MFLPD greater than the FRTP during power ascension up to 90K of 'RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to 100K times MFLPD, provided that the adjusted APRM reading does not exceed 100K of RATED THERMAL POWER, the required gain. adjustment increment does not exceed 10K of RATED THERMAL POWER, and a notice of the adjustment is posted on the reactor control panel ..
SUS(UEHANNA - UNIT 1 3/4 2-5 . Amendment No. 29 QpCCt $ (COf'fence g~g I' g el'g f' 5<n)!4 /SPED]e '6pCAdVOve CgSPIWset~fp .
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~ 0 0 0 OD Q.$ 0A '4006 007 008 009 1 1.20 A: Mo}n Tvrbine Bypass 'inoperable; EOC RPT Operable 7'Qrve Curve 8: EOC RPT lnapembhr, Main Turbine Bypass Operable Curve C: EOC RPT and Mafn Turbine Bypass Operable
- RBM o PER TABLE 3.3.6-2 RGURE 3.2.3-'lb >7RiP FvlJcnod
(. OL, 2 SUSQUEHANNA UNT $
TABLE 3.3.6-2 CONTROL ROD BLOCK IHSTRUHEHTATION SETPOINI'S x TRIP F UNCT ION TRIP SETPOIHT ALLOWABLE VALUE
- l. ROD BLOCK HONITOR
~ ~
IH~ HA A
- c. Downscale > 5/125 divisions of full scale > 3/125 of divisions full scale
- z. APRH a.
b. F)ow Biased Heutron
. Flux " Upscale Inoperative
~ <
HA 0.50 M+%OX~ < HA 0.58 M + 53X"
- c. Down sea le > SX of RATED TIIERHAL POWER > 3X of RATED TNERHAL POMER
- d. Neutron Flux - Upscale Startup < 12X of RATED TIIERHAL POWER < )4X of RATED TIIERHAL POWER
- 3. SOURCE RANGE HONITORS
- a. Detector not full in HA HA
- b. Upsca)e < 2x 10 cps <4 x lo cps a.. Inoperative HA HA
- d. Oownsca'le 3 cps A+ cpsAA 4.. INTERMEDIATE RANGE HONITORS
- a. Detector not full ln HA NA
~ b. Upscale < 108/)25 divisions of full scale < 'llo/125 dtvislons of full'cale C. Inoperative HA HA
- d. Downscale > 5/125 dlvjslons of full scale > 3/)25 x)ivislons of full scale
- 5. SCRAH DISCHARGE VOLUHE ID CL a. Mater Level - lligh < 44 ga)lons < 44 gallons B
Ill 6. REACTOR COOLANT SYSTEH RECIRCULATION FLOW rt
- a. Upsca)e' 108/125 divisions of full scale ill/125 divisions of full scale O b. Inoperative HA HA
- c. Compara tor < IOX flow deviation 11X flow deviation The Average Power Range Honitor rod block fpnction ls varied ps a function of recilculatlon loo>> flow (M). The trip.setting of this function oust be eaintained ln accordance with Specification 3.2.2.
" initial loading al>>l startup the co>>>>t rate. cay > 0.5 cps.
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SURVEYS.L'LANCE REQUIREMENTS 4.4. 1. 1.1. Each pump discharge val.ve and bypass valve shall. be demonstrated OPERABLE. by cycling each valve through at least one complete cycle of full travel during each startup"" prior to THERMAL POSER exceeding 25 o of RATED THERMAL POMER.
++.t.t.(.2-to be
.E hP Pdi h g closed zt least once per PPP 31 days.
I, if t ~ PERAPLE, I IIP ifid
+.+. I. I-I 3 E h P MG t p t h .\ t I i d h I I ' h 11 h demonstrated OPERABLE whith overspeed setpoints less than or equal to 102.5 and 105K, respectively, of rated core flow, at least once per 18 months.
I. l./.4- 'y, E t hii h I 11'PRM' LPIIM' f1 a point within 5X RATED THERMAL POQER of the 100K rated rod line with total core flow between 35ro and 50.o of rated total core flow during startup testing fol lowing each refueling outage.
""Ef Pnot performed within the previous 31 days.
FIgure 3.4.1.1-1 THERMAL POWER LIMITATIONS 80 r~~ 70- " REGION GREATER THAN LIMIT CI ~ g ao--
~ ~
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6l 20- r I 0 l O ' ~ ~ ~ 10-r
'0
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Figure 3.4.1.1.1-1 THERMAL POWER LIMITATIONS 80 ~ ~ ~~ ~ ~ ~~~ J ~ ~ J J ~ ~ 1
'I
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~ 1 ~ ~
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,.REACTOR COOLANT SYSTEM RECIRCULATION LOOPS << SINGLE LOOP OPERATION LIMITING CONDITION FOR OPERATION 3.4. 1. l. 2 One reactor coolant recirculation loop shall be in operation with the pump speed ~ 80% of the rated pump speed, and
- a. the following revised specification limits shall be followed:
1.'pecification 2.1.2: the MCPR Safety Limit shall be increased to 1.07.
- 2. Table 2.2.1-1: the APRM Flow-Biased Scram Trip Setpoints shall be as follows:
Tri Set oint Allowable Value
~ 0.58W + 55% ~ 0.58W + 58%.
- 3. Specification 3.2.1: the MAPLHGR limits shall be as follows:
- a. GE fuel: the limits specified in Figures 3.2.1-1 and 3.2.1-2, multiplied by 0.81.
- b. Exxon fuel: the limits specified in Figure 3.2.1-2 multiplied by 0.81.
- 4. Specification 3.2.2: the APRM Setpoints shall be as follows:
Allowable Value . S ~ (0.58W + 55%)T S (0.58W + 58%)T SRB~ (0.58W + 46%)T S W (0.58W + 49%)T
- 5. Table 3.3.6-2: the RBM/APRM Control Rod Block Setpoints shall be as follows'.
RBM Upscale Tri Set oint Allowable Value
- l. ~ 0.66W
~ 0.66W
+ 35% ~0.66W + 38%
~ 0.66W 20 + 37% + 40%
5.a.l and'.a.2 shall be used in conjunction with the MCPR limits specified in Figures 3.2.3-la and 3.2.3-1b, respectively. Allowable Value
~0.58W + 46% 0.58W + 49%
- b. APRM and LPRM*** neutron flux noise levels shall be less than three times their established baseline levels when THERMAL POWER is greater than the limit specified in Figure 3/4.1.1.1-1.
APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*, except during two loop operation./3
ACTION: a ~ With no reactor coolant system recirculation loops in operation, take the ACTION required by Specification 3.4.1.1.1. With any of the limits specified in 3.4.1.1.2a not satisfied, take the ACTION required by the referenced Specification. C ~ With the APRM or LPRM*** neutron flux noise levels greater than or equal to three times their established baseline levels when THERMAL POWER is greater than the. limit specified in Figure 3/4.1.1.1-1, immediately, initiate corrective action and restore the noise levels to within the required limits within 2 hours by initiating an orderly reduction of THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1.1-1. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
- d. With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours.
SURVEILLANCE RE UIREMENTS 4.4.1.1.2.1 Upon entering single loop operation and at least once per 24 hours thereafter, verify that the pump speed in the operating loop is ~ 80% of the rated pump speed. 4.4.1.1.2.2 With THERMAL POWER greater than the limit specified. in Figure 3.4.1.1.1-1, determine the APRM and LPRM*** neutron flux noise levels within 1 houi. Continue to determine the noise levels at least once per 8 hours and within 30 minutes after the completion of a THERMAL POWER increase' 5% of RATED THERMAL POWER. 4.4.1.1.2.3 Within 15 minutes prior to either THERMAL POWER increase resulting from a control rod withdrawal or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is 30%**** of RATED THERMAL POWER or the recirculation loop flow in 'the operating recirculation loop is 4 50%*+** of rated loop flow:
- a. ~ 145'F between reactor vessel steam space coolant and bottom head drain line coolant, b.kMÃ50'F between the reactor coolant within the loop not in operation and the coolant in the reactor pressure vessel, and c.fM 50'F between the reactor coolant within the loop not in operation and operating loop.
4.4.1.1.2.4 a. Establish a baseline APRM and LPRM neutron flux noise value
.at a point within 5% RATED THERMAL POWER of the 100% rated
rod line with total core flow between 35% and 50% of rated total core flow during startup testing following each refueling outage, or
- b. In lieu of establishing a single loop operation baseline value, utilize the value established pursuant to Specification 4.4.1.1.1.4 if a baseline value is needed to meet the requirements of Specification 3.4.1.1.2.
4.4.1.1.2.5 The pump discharge valve and bypass valve in both loops shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each startup** prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER. 4.4.1.1.2.6 The pump discharge bypass valve in the OPERABLE loop, if not OPERABLE, shall be verified to be closed at least once per 31 days. 4.4.1.1.2.7 The pump MG set scoop tube electrical and mechanical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 102.5 and 105%, respectively, of rated core flow, at least once per 18 months. 4.4.1.1.2.8 The pump discharge valve and bypass valve in the inoperable loop, if not OPERABLE, shall be verified to be closed at least once per 31 days. 4.4.1.1.2.9 During single recirculation loop operation, all get pumps, including. those in the inoperable loop, shall be demonstrated OPERABLE at least once per 24 hours by verifying that no two of the following conditions occur:888
- a. The indicated recirculation loop flow in the operating loop differs by more than 10% from the established single recirculation pump speed-loop flow characteristics.
b; The indicated total core flow differs by more than 10% from the established total core flow value from single recirculation loop flow measurements.
- c. The indicated difference-to-lower plenum differential pressure of any individual )et pump differs from established single recirculation loop patterns by more than.10%.
4.4.1.1.2.10 The SURVEILLANCE REQUIREMENTS associated with the specifications referenced in 3.4.1.1.2a shall be followed. See Special Test Exception 3.10.4. If not performed within the previous 31 days. Detectors A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.
1 ~
**~* Initial value. Final value to be determined based on required change to this value shall be submitted to the startup'esting.
Any, Commission within 90 days of test completion. See Specification 3.4.l.l.l for two loop operation requirements. M This requirement does not apply when the loop not in operation is isolated from the reactor pressure vessel. 888 During startup testing following each refueling outage, data shall be recorded for the parameters listed to provide a basis for establishing the specified relationships. Comparisons of the actual data in accordance with the criteria listed shall commence upon the performance of subsequent required surveillances.
REAC'FOR COOLANT SYSTEM JET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE. APPLICASILITY: OPERATIONAL CONDITIONS I and ~wAe Lnt/ rain oJu<< l ass fA o)4(aalu loA < ACTION: With one or more jet pumps inoperable, be in at least HOT SHUTDOWN-within 12 .hours.- SURVEILLANCE RE UIREMENTS
- 4. 4. l. 2 Each of the above required jet pumps shall be demonstrated OPERABLE prior to THERMAL POWER exceeding 25K of RATED THERMAL POWER and at: least once per"24 hours by determining recirculation loop flow, total core flow and diffuser to-lower plenum differential pressure for each jet pump and verifying that no two of the following conditions occur when the recirculation pumps are operating at the same 'speed:
- a. The indicated recircu1ation loop flow differs by more than l(C from the established pump speed-loop flow characteristics.
- b. The indicated total core flow differs by more than 10K from the established total core flow value derived from recirculation loop flow measurements.
- c. The indicated diffuser"to-lower plenum differential pressure of any individual jet pump differs from'stablished patterns by more than lOX.
SUS(UEHANNA - UNIT 1 3/4 4-2 Amendment No. 29'
, ~
REACTOR COOLANT SYSTEM RECIRCULATION PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.3 Recirculation'ump speed mismatch shall be maintained within:
- a. 5Ã of each'other with core flow greater than or equal to 75K of rated core flow.
- b. 10K of each other with core flow less than 75K of rated core flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1" and 2"~g44o& rcciic~(~io loot s
~C lg 0$ 4f'4.t'444 ACTION:
With the recirculation pump speeds different by more than the specified limits, either:
'a. Restore the recirculation pump speeds to within the specified limit within 2 hours, or
- b. Declare the recirculatiori loop of the pump with the slower speed not in'operation and take the ACTION require by. Specification 3.4. l. 1. l.'f SURVEILLANCE RE UIREMENTS 4.4.1.3 Recirculation pump speed mismatch shall be verified to be within the limits at least once per 24 hours.
See Special Test Exception 3.10.4. SUSQUEHANNA - UNIT 1- r 3/4 4~3
SPECIAL TEST EXCEPTIONS '/4.'10. 4 RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION J 3.10.4 The requirements of Specifications 3.4.1.1 and 3.4.1.3 may be suspended for. up to 24 hours for the performance. of:
- a. PHYSICS TESTS, provided that THERMAL POWER does not exceed 5X of .RATED THERMAL POWER, or
- b. The Startup Test Program.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2, during. PHYSICS TESTS and the Startup Test Program. ACTION: With the above specified time limit exceeded, insert all control rods.
- b. With the above specified THERMA POWER limit exceeded during PHYSICS TESTS, immediately place the reactor mode switch in the Shutdown position.
SURVEILLANCE RE UIREMENTS 4.10;4.1 The time during which the above specified requirement has been suspended shall be verified to be less than 24 hours at least once per hour during PHYSICS TESTS and the Startup Test, Program.
- 4. 10.4.2 THERMAL POWER shall be determined to be less than 5X of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.
SUSQUEHANNA - UNIT 1 3/4 10-4 Amendment No. 24
J REACTIVITY CONTROL SYSTEMS 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 i,nsertion times are consistent with those used in the accident analysis, and (3) limit the potential effects of the rod drop accident. The ACTION statements permit variations from the basic but at the same time impose more restrictive criteria for continued . 4 'equirements 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. ~ 3 Damage within the conti ol rod drive mechanism could be a generic problem, therefore with a control rod immovable 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 SHUTDOWN MARGIN requirements.
~s
~
The number of control rods permitted to be inoperable could be more than the eight allowed by the specification,'ut'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. X~-Ae-8+-]i8 8~0. The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from ecoming less than . i t ly di during the This analysis shows that the negative reactivity routes resulting from the scram with the average response of all the drives as given in the specifications rovide th required protection and MCPR remains greater than . . The occurrence of scram times longer then those specified should be viewed as an indication of a ~pete&~ problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem. The scram discharge volume is required to be OPERABLE s'o that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required. Control ro'ds.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 has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.
'SUS(UEHANNA - UNIT 1" B 3/4 1'-2
POWER DISTRIBUTION LIMITS I s BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit HCPRs at steady state operating conditions. as specified in Specification 3.2.3 are derived from the established fuel cladding'ntegrity Safety Limit MCPR e+-ll-.86, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evalua-tion with the initial condition of the reactor being at th'e steady state operating limit, it is required that the resulting HCPR 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 abnormal operati'onal transient, the most limiting tran-. sients have been analyzed to determine which result in the largest reduction in CRITICAL.POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta MCPR. When added to the Safety Limit MCPR ~~8, the required minimum operating limit MCPR of Specification 3. 2.3 is obtained and presented in ~~ ~ 3. 2.3"l. xv-HP -8+ -li8 E)yon TD The evaluation of a given transient begins with the ystem initial
'parameters shown in . that are input to a -core dynamic behavior ,transient computer program.
I The outputs of this'program along with the initial MCPR form the input for further analyses of the. thermally limiting bundl e. 1 in MCPR caused by the transient. uepgr The purpose of the 0~f~~ of Figure 3.2.3-1 is to define operating limits at other than rated core flow conditions. At less than 100K of rated flow the required MCPR is e that the Safety Limit MCPR will not be violated during a flow increase u (JQA,T'g SUSQUEHANNA - UNIT 1 B 3/4 2-4
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3/4.4 REACTOR COOLANT SYSTEM BASES 3/4. 4.1 RECIRCULATION SYSTEM
~ ~
~ ~ ~
THERMAL POWER, core flow, and neutron flux noise level limitations are prescribed in accordance with the recommendations of General Electric Service Information Letter No. 380, Revision 1, "BWR Core Thermal Hydraulic Stability," dated February 10, 1984. An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in ca0e of a design"basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump .inoperable. Jet pump failure can be detected by monitoring jet pump performance on 'a prescribed schedule for significant degradation..
+wo (gyes 0lSCAfl+0 ~
Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA hnalysis design criteri The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA. a
'~~ 'f 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'ther prior to startup of .an idle loop. The loop temperature must also be within 50 F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper regions of the core, undue stress on the vessel would result if the temperature difference was greater than 1'45'F:
3/4.4.2 SAFETY/RELIEF VALVES The safety valve function of the safety/relief valves operate to prevent the r eadtor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 10 OPERABLE safety-relief valves is. required to limit reactor pressure to within ASME III allowable values for the worst case upset transient, Oemonstration,of the safety/relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Specification 4. 0. 5. SUSQUEHANNA - UNIT 1 B. 3/4 4-1 Amendment No.29
X. Operation with one reactor recirculation l'oop inoperable has and found acceptable, provided that the unit is operated in been'valuated accordance with Specificati'on 3.4.1.1.2. C For single loop operation, the MAPLHGR limits are multiplied by a factor of 0.81.'se of this factor in con]unction with the GE high enriched fuel MAPLHGR is conservative to use as a limit for the Exxon fuel. The multiplication factor is derived from LOCA analyses initiated from single loop operation; it maintains the same peak clad temperature margin to the licensing limit of 2200'F as in two loop operation. For single loop operation, the RBM and APRM setpoints, are ad)usted by a 7X decrease in recirculation drive flow to account for the active loop drive flow that bypasses the coze and goes up through the inactive loop )et pumps Surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive reactoz vessel internals vibration. Surveillance'on differential temperatures below the threshold limits on THERMAL POWER or recirculation loop flow'itigates undue thermal stress on vessel nozzles, recirculation pumps and the vessel bottom head during extended operation in the single loop mode. The threshold limits are those values which will sweep up the cold water from the vessel bottom head. Y. In the case where the mismatch limits cannot be maintained during the loop operation, continued operation is permitted in the single loop mode.
~l A 2.0 SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS
- 2. 1 SAFETY LIMITS THERMAL POWER Low Pressure or Low Flow 2.1.1 THERMAL POWFR shall not exceed 25K of RATED THERMAL POWER with the reactor vessel steam dome pressure l.ess than 785 psig or core flow less than lOX of rated flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: 1 With THERMAL POWER exceeding 25K of RATED THERMAL POWER and 'the reactor vessel steam dome pressure less than 785 osig or core flow less than 10 of rated flow, be in at least HOT SHUTDOWN within 2 hours and comply wi th the. requirements of Specification 6.7.1. THERMAL POWER. Hi h Pressure and !lioh F'ow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10X of rated '.ow. APPLICABILITY: GPERATIONAL CG D. IGNS 1 and 2. ACTION: With MCPR less than 1.06 and the reactor vessel steam dome pressure greater than 785 psig and,core flow greater than 10~ of rated flow, be in at least HOT SHUTDOWN 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 sot exceed 1325 psig.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4. ACTION: With the rea'ctor coolant system pressure, as measured in the reacto~ vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN 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. Scc Spec((A~ <~ 9 ff I f~ ~'4 'gf'(etc, loogt of&osl~ f+$ 44il c~cevf SUSQUEHANNA - UNIT 2 2-,1
TABLE 2 2 l-l REACTOR PROTECTION SYSTEM INSTRUIIENTATIOH SETPOINTS CA ALLOWABLE FUNCTIONAL UNIT TRIP SETPOIHT VALUES >> 1. Intermediate Range Monitor, Neutron Flux-High < 120/125 divisions < 122/125 divisions of full scale of ful) scale
- 2. Average Power Range Monitor:
c a. ~ Neutron Flux-Upscale, Setdown < 15X of RATED THERMAL POWER < 20'f RATED e THERMAL POWER
- b. Flow Biased Simulated Thermal Power-Upscale 4F wi tM
')
Flow Biased 0.58 M<59X, with < 0. 58 M~ 62X, a maximum of a maximum of
- 2) High Flow Clamped < 113.5X of RATED < 115.5X of'RATED TIIERtQL POMER THERMAL POWER
- c. Neutron Flux-Upscale < llQX of RATED THERMAL POMER < 120X of RATEO THERMAL POWER
- d. Inoperative HA
- 3. Reactor Vessel. Steam Dome Pressure - High < 1037 psig < 1057 psig 4 Reactor Vessel Mater Level - Low, Level 3 > 13.0 inches above > 11.5 inches above instrument zero" instrument zero
- 5. Main Steam Line Isolation Valve - Closure < 10X closed < llX closed
- 6. Main Steam Line Radiation - High < 3.0 x full power < 3. 6 x full power background background
- 7. Drywell Pressure - High < 1.72 psig < 1.88 psig
- 8. Scram Discharge Volume Mater Level " High
- a. Level Transmitter < 88 gallons. < 88 gal l ons
- b. Float Switch < 88 gallons < 88 gal 1 ons
- 9. Turbine Stop Valve - Closure. < 5 5X closed' < 7X closed
- 10. Turbine Control Valve Faqt Closure, Trip Oil Pressure - Low 500 psig > 460 psig ll. Reactor Mode Switch Shutdown Position HA HA
- 12. Manual Scram See Bases Figure B 3/4 3-1.
r I See Spec'L(ieee e 3A. <.I.2.e A,e er~g(e (eep 'eredr eve)>> n
2a 1 SAFETY. LIMITS BASES
- 2. 0 INTRODUCTION The fuel cladding, reactor pressure vessel and primary system piptng are the principal barriers'to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these b barriers during normal plant operations and anticipated transients. The fuel 4 V
cladding integrity Safety Limit is set such that no fuel damage is calculated to- occur if the limit is not, violated. Because fuel damage is not directly
~
4 Ol the MCPR is not ess than .. servable a ste -back approach is used to establish a Safet Limit such that MCPR greater than . represents a conser vative margin relative to the conditions required to maintain fuel cladding Vi iniegrity. 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. Al.hough 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 signifi-cantly above design conditions and the Limiting Safety System Settfhgs. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perfora ions signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a sianificant .departure from the condition intended by design for planned operation.
- 2. l. 1 THERMAL POWER Low Pressure or Low Flow.
The use of the GEXL correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10K of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing 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. 'nalyses show that with a bundle flow. of 28 x 10'bs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greate~ than 28 x 10'bs/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 50K of RATED THERMAL POWER. Thus, a THERMAL POWER 'limit of 25K of RATED THERMAL POWER for reactor pressure below 785 psig i's conservative. SUSQUEHANNA - UNIT 2 B 2-1
Bases Table B2. 1.2-1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT" Standard Deviation uanti "of Poi Feedwater low l.j7 Feedwater Tem)erature .76 Reactor Pressure~ 0. 5-Core Inlet Temperatu e 0.2 Core Total Flow 2.5 Channel Flow Area 3.0 Friction Factor Multiplier 10. 0 Channel Friction Factor Multiplier 5.0 TIP Readings 6.3 R Factor 1.5 Critical Power 3.6 he ncertalnty analysis used to establish the core wide Safety Lim't MCPR is ba d on the assumption of quadrant power symmetry for the reactor c e. SUSQUEHANNA - UNIT 2 B 2-3
3/4.2 'OWER DISTRIBUTION LIMITS 3/4 2 1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING.CONDITION FOR OPERATION ~
- 3. Z. 1 All AVERAGE PLANAR LINEAR. HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVFRAGE PLANAR EXPOSURE shall not exceed the limits in Figu'res 3.2.1-1, 3.2. 1-Z, and 3.2. 1-3.+ 'hown APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25K of RATED THERMAL POWER.
ACTION: With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, or 3.2.1"3, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours or reduce THERMAL POWER to less than 25~ of PATED THERMAL POW R w.thin the next 4 hours.
. SURVEILLANCE REQUIREMENTS'.
- 2. 1 All APLHGRs shall'be verified to .be equal to or less th'an the lim~ts determined from Figures 3.2. 1-1, 3.2. 1-2, and 3.2. 1-3:
- a. At least once per 24 hours,
- b. Within 12 hours after completion of a THERMAL POWER increase of at .
least 15 of RATED THERMAL POWER, and
- c. Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL ROD PATTERN for'PLHGR.
- d. The provisions of Specification 4.0.4 are not applicable.
8 See ~ col(<'e>i'e 9-'f.t t ~ ~ 4" SUS(UEHANNA - UNIT 2 3/4 2-1
~ Is fs POWER DISTRIBUTION LIHITS 3/4. 2. 2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRH flow biased simulated thermal power-upscale scram trip setpoint (5) and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships:
~T Allowable Value~
S < (0.58W + 59,)T S < (0.58W + 62 )T
< (0.58W'+ 50=)T SRB SRB (0'58W + 53K T where: S and SRB are in percent of RATED THERMAL POWER, W = Loop recirculation flow as a percentage of the loop recirculation f]ow which produces a rated core flow of 100 million lbs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER OENSITY. T is always less than or equal to 1.0.
AP. L:CABILITY: OPERATICs'lAL CONDITION 3., when THERI~IAL POWER is greater than or equal to 25% cf RATED THERMAL POWER. ACTION: Nith the APRN flow h asad simulated thermal power-ups=ale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conservatisve mohan the value shown in the Allowable Value column for S or So", as above determin'ed, initiate cor rective action within 15 mirutes and adjust S and/or SRB to be consistent with the Trip Setpoint value" within 2 hours or reduce iHERHAL POWER to less than 25K of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.2 The FRTP and the HFLPD shall be determined, the value of T calculated, and the most recent actual APRH flow biased simulated thermal power upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:
- a. At least once per 24 hours,
- b. Within 12 hours after completion of a THERMAL POWER increase of at least 15K of RATED THERMAL POWER, and
- c. Initially and at least once per .12 hours when the reactor is operating with HFLPD greater than or equal to FRTP.
- d. 7he provisions of Specification 4.0.4 are not applicable.
"With MFLPO greater than the FRTP during power ascension up to 90"s of RATED THERMAL POWER, rather than adjusting the APRM setpoints,.the APRM gain may be adjusted such that APRH readings are greater th'an or equal to 100~ times HFLPD, provided that the adjusted APRH reading does not exceed 100Ã of RATED THERMAL POWER, the required gain adjustment increment does not exceed 10~ of RATED THERMAL POWER, and a notice of the adjus ment is posted on the reactor control panels s
SUSQUEHANNA - UNIT 2 3/4 2"5 5p<cAa'csdf~ g.g.(.l 2,p. d'or sl~lc. /dhcpg dppcrcvbems- I cgmf'~4~
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TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUHENTATION SETPOINTS TRIP FUNCTION 1 RIP SETPOIHT ALLOWABLE VALUE
- l. ROD BLOCK HONITOR
<%En-r a.
b. 4peea~~ Inoperative q~~
~~~ HA HA C
'r M
- c. Downs ca 1'e > 5/125 divisions of full scale > 3/125 of divisions full scale
- 2. APRH
- a. Flow Biased Heutrop Flux - Upscale~ < 0.58 M+ SOZ. < 0.58 M + 53K".
- b. Inoperative HA HA
- c. Downscale > 5X of RATED THERHAL POWER > 3X of RATED TliERHAL POWER
- d. Neutron Flux - Upscale Startup < 12K of RATED TIIERHAL POWER < 14K of RATED -TtlERHAL POWER
- 3. SOURCE RANGE HONITDRS
- a. Detector not full in HA HA Col b. Upscale <2x 10 cps <4x 10 cps
- c. Inoperative HA HA
- d. Downscale, > 0.7 cps"" ) 0 5 cps't 4 IHTERHEOIATE RANGE HOHITORS
- a. Detector. not full in HA HA
- b. Upscale < 108/125 divisions of full scale < 110/125 divisions of full.scale
- c. Inoperative HA HA
- d. . Downscale > 5/125 divisions of full scale > 3/125 divisions of full scale
- 5. SCRAH DISCHARGE VOLUHE
- a. Mater Level - Nigh < 44 gallons < 44 gallons
- 6. REACTOR COOLANT SYSTEH RECIRCULATION FLOM
- a. Upscale ~ < 108/125 divisions of full scale < ill/l25 divisions of full scale
- b. Inoperative NA HA
- c. Comparator ~ < 10K flow devia'Lion < 11K flow deviation The Average Power. Range Honitor rod block function is varied as a function of recirculation loop flow (M). The trip setting of this function must be maintained in accordance with Specification 3.2.2.
""Provided signal-to-noise ratio is > 2. Otherwise, 3 cps as trip setpoint and 2.8 cps for allowable value.
n ~
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+0 Q'ea GpccAiadcon 3. A( I. 2- .odor sio3(c loof y~eraho>> rc~uirceooVs
3/<. < REACTOR COOLANT S EM 3/4. 4. 1 RECIRCULATION S EM 'R CIRCULATION LOOPS - 740 upas otemr~oA LIMITING CONDITION FOR OPERATION 3.4.1.1.I Two reactor coolant system recirculation loops shall be in operation. and:
'otal core flow shall be greater than or equal to 45 million lbs/hr, or
- b. THERMAL POWER shall be less than or 'equal to.the limii specified in Figure ~~1.. 9.4.l.l.l I APPI ICABILITY: OPERATIONAL CONDITIONS 1* and 2"~c~t ~i <<<!~ ~+io-.
~ 4<~ ~
ACTION: ~i p 'f <'4 t4. sigh .(iieet pfcroAi+~ Icgaii'oa ovfs of $ )gcafscopio~
+el,l, Z one c or coolant system recirculaticn loop not in operation, ol'ith
~
g ~ be in at least HOT 'SHUTDOWN.
- b. With no reac or coolant sys.em recirculation loops in operatiicn, imm diately initiate an than or equal to the limit specifiied in Figure measures to place the unit in at least STARTUP within)6 hours and in
~~,
orderly reduction of THERMAL POWER to less and initiate HOT SHUTDOWN within the next 6 hours. 3.V.< I. l- I
~
- c. With two reac or coolant system recirculation loops in c; aiion and than the limit specified in Figure ~
total core flow less than 45 million lbs/hr and THERMAL Pic..-. nreater Sent.I (~ I i~~ C~
- 1. Reduce. THLRMAL POWER to less han ar equal to the 1',mit specified in Figure or
- 2. Increase core flow to greater than 45 million lbs/hr, or
- 3. Determine the APRM and LPRtP"~ neutron flux noise levels 'wi thi n l hour, and:
a) Ii .the APRM and LPRM""" neutron flux noise levels are less than three times their. established baseline levels', continue to determine the noise levels at least once per 8 hours and within 30 minu es after the completion of a THERMAL POWER increase of at least 5X of RATED THERMAL POWER, or b) If the APRM cr LPRM~"~ neutron flux noise levels are greater
'than or eaual :o three times their established baseIine levels,'mmediately initiate corrective ac ion and restore the noise levels to within the required limits within 2 hcurs by increasing core flow to greater than 45 million lbs/h'r, and/cr by initiating an orderly reduction of THERiMAL POWEP, to less than or equal to the limit specified in Figure
- 5. 'I.
"See Soecial Test Exception 3.~0. 4.
"""..Detec:ors A and C of. one LPRM string per core oc ant plus detectors A anci C of one LPRM st".ing in the center of the core should be'monitored.
~p<<'<'~i'n > 5 l.l < pie sl p!c (os bppw ~ fc(coal'~tops (U(c;i <'A<i)i" Ua! T ' "/4 Amendment No. 2
SURVEILLANCE REQUIREMENTS 4;4.1.1.1.! Each pump discharge valve and bypass valve shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each 's artup~~ prior to THERMAL POWER exceeding 25- of RATED THERMAL POWER. +.+. t .l. t. 2 E I p PPI P E Pyp 1,1'IPEEAEIE, PIIP to be closed at least once per 31 days. +-+"I I t.3 Each pump MG set scoop tube electrical and mechanical stop shall be demonstrated OPERABLE with overspeed, setpoints less than or equal to 102;5 and 105K, respectively, of rated core flow,'t least once per 18 months. ++.t.t,<.+ Establish a baseline APRM and LPRPP neutron flux noise value at a point within SX RATED THERMAL POWER of the 100." rated rod line with total core flow between 35~'nd 50Ã of"rated total core flow during startup testing following each refueling outage.
~"If not performed. within the. pr vious 31 days.
CA m Figure 3.4.1.1-1 THERMAL POWER LIMITATIONS 80
" ~ ~
~ r REGION GREATER THAN LIMIT.
P 70- .
* ".'. """ "c"."--."I"""" ~ ~ ~ ~ ~ ~ ~ ~ ts ~ ~
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REACTOR COOLANT SYSTEM RECIRCULATION LOOPS SINGLE LOOP OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1.2 One reactor coolant recirculation loop shall be in operation with the pump speed ~ 90% of the rated 'pump speed, and a.. the following revised specification limits shall be followed:
- 1. Specification 2.1.2: the MCPR Safety Limit shall be increased to 1.07.
2.. Table 2.2.1-1: the APRM Flow-Biased Scram Trip Setpoints. shall be as follows: Tri Set oint Allowable Value 0.58W + 55% ~0.58W + 58%.
- 3. Specification 3.2.1: the MAPLHGR limits shall be the limits specified in Figures 3.2.1-1, 3.2.1-2, and 3.2.1-3, multiplied by 0.81.
- 4. Specification 3.2.2: the APRM Setpoints shall be as follows:
Allowable Value S ~ (0.58W + 55%)T S ~ (0.58W + 58%)T SR +(0.58W + 46%)T S~>> (0.58W + 49%)T
- 5. Table 3.3.6-2: the RBM/APRM Control Rod Block Setpoints shall be as follows:
- a. RBM Upscale Tri Set oint Allowable Value
- 1. W 0. 66W + 35% 0.66W + 38%
2~ ~0.66W + 37% ~0.66W + 40% 5.a.l and 5.a.2 shall be used in conjunction with the MCPR limits specified in Figures 3.2.3-1a and 3.2.3-lb, respectively. Allowable Value
~0.58W + 46% ~0.58W + 49%
- b. APRM and LPRM+** neutron flux noise levels shall be less than three times their established baseline levels when THERMAL POWER is greater than the limit specified in Figure 3/4.1.1.1-1.
APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*, except during two loop operation./f ACTION:
- a. With no reactor coolant system recirculation loops in operation, take the ACTION required by Specif'ication 3.4.1.1.1.
- b. With any of the limits specified in 3.4.1.1.2a not satisfied, take, the ACTION required by the referenced Specification.
C ~ With the APRM or LPRM*** neutron flux noise levels greater than or equal to three times their established baseline levels when THERMAL POWER is greater than the limit specified in Figure 3/4.1.1.1-1, immediately initiate corrective action and restore the noise levels to within the required limits within 2 hours by initiating an orderly reduction'of THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1.1-1. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours.
- d. With one or more get pumps inoperable, be in 'at least HOT SHUTDOWN within 12 hours.
SURVEILLANCE RE UIREMENTS 4.4.1.1.2.1 Upon entering single loop operation and at least once per 24 hours thereafter, verify that the pump speed in the operating loop is ~90% of the rated pump speed. 4.4.1.1.2.2 With THERMAL POWER greater than the limit specified in Figure 3.4.1.1.1-1, determine the APRM and LPRM*+* neutron flux noise levels within 1 hour. Continue to determine the noise levels at least once per 8 hours and within 30 minutes after the completion of a THERMAL POWER increase ~5% of RATED 'THERMAL POWER. 4.4.1.1.2.3 Within 15 minutes prior to either THERMAL POWER increase resulting from a control rod withdrawal or recirculation loop flow increase, verify that the following differential temperature requirements are met if THERMAL POWER is ~30%**** of RATED THERMAL POWER or the recirculation loop flow in the operating recirculation loop is ~ 50%**** of rated loop flow:
- a. ~145'F between reactor vessel steam space coolant'nd bottom head drain line coolant, b./38 ~50'F between the reactor coolant within .the loop not in operation and the coolant in the reactor pressure vessel, and c.f/8 ~50'F between the reactor coolant within the loop not in operation and operating loop.
4.4.1.1.2.4. a. Establish a baseline APRM and LPRM neutron flux noise value at a point within 5% RATED THERMAL 'POWER of the 100% rated rod line with total core flow between 35% and 50% of rated total core flow during startup testing following each refueling outage, or
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- b. In lieu of establishing a single loop operation baseline value, utilize the value established pursuant to Specification 4.4.1.1.1.4 if a baseline value is needed to meet the requirements of Specification 3.4.1.1.2.
4.4. 1. 1. 2.5 The pump discharge valve and bypass valve in both loops shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each startup** prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER. 4.4.1.1.2.6 The pump discharge bypass valve in the OPERABLE loop, if not OPERABLE, shall be verified to be closed at least once per 31 days. 4.4.1.1.2.7 The pump 'MG set scoop tube electrical and mechanical stop'hall be demonstrated OPERABLE with overspeed setpoints less than or equal to 102.5 and 105%, respectively, of rated core, flow, at least once per 18 months. 4.4.1.1.2.8 The pump discharge valve and bypass valve in the inoperable loop, if not OPERABLE, shall be verified to be closed at least once per 31 days. 4.4.1.1.2.9 During single recirculation loop operation, all jet pumps, including those in the .inoperable loop, shall be demonstrated OPERABLE at least once per 24 hours by verifying that no two of the following conditions occur:!ll$ ff
- a. The indicated recirculation loop flow in the operating loop differs by more than 10% fr'om the established single recirculation pump speed-loop flow characteristics.
- b. The indicated total core flow differs by more than 10% from the established total core flow value from single recirculation loop flow measurements.
- c. The indicated difference-to-lower plenum differential pressure of any individual jet pump differs from established single recirculation loop patterns by more than 10%;
4.4.1.1.2.10 The SURVEILLANCE REQUIREMENTS associated with the specifications referenced in 3.4.1.1.2a shall be followed.'ee Special Test Exception 3.10.4. If not performed within the previous 31 days. D'etectors A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored. Initial value. Final value to be determined based on startup
. testing. Any required change to this value shall be submitted to the Commission within 90 days of test completion.
I See Specification 3.4.1.1.1 for two loop operation requirements. 88 This requirement does not apply when the loop not in operation is isolated from the reactor pressure vessel. During startup testing following each refueling outage, data shall be recorded for the parameters listed to provide a basis for establishing the specified relationships. Comparisons of the actual data in accordance with the criteria listed shall commence upon the performance of subsequent required surveillances.
REACTOR COOLANT SYSTEM
\
JET PUMPS LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.: APPLICABILITY: OPERATIONAL CONDITIONS 1 and c.' Zc ~~ h~W rEc rM~~ l ~>>' ACTEQM: Mith one or more jet pumps in'operable, be in at least HQT SHUTDOWN within 12 hours. SURVEILLANCE REOUIREMEHTS
- 4. 4. 1.2 fach of the above required jet pumps sha'I'I be demonstrated OPERASLE prior to THERMAL POMER exceeding 25 of RATED THERMAL POWER and at least once per 24 hours" by determining recirculation loop flow, total core flow and diffuser-to"lower plenum differential pressure for each jet pump and. verifying that no two of the following condi ions occur when the recirculation pumps are operating at the same speed:
- a. The indicated recirculation loop flow differs by more than 10~ rom the established pump speed-loop flow characteristics.
- b. The indicated total core flow differs by more tshan 10 from the established total core flow value derived from recirculation loop flow measurements.
- c. The indicated diffuser to-lower plenum differential pressure of any individual jet pump differs from established patterns by more than 10'~.
"During the startup test program, data shall be recorded for the parameters listed to provide a basis for'stablishing the specified relationships.
Comparisons'f the actual data in accordance with the criteria listed shall ~ commence upon the conclusion of the startup test program. 4C SP CC ((Cafe ~ 4 '4, (, l,k 9 +y'(mSglt (dPCPP P(PI~ S~ f'(eg 44(f~g ~ C'i C'0 'I I'k( '7
REACTOR COOLANT SYSTEM RECIRCULATION PNPS LIMITING CONDITION FOR OPERATION 3.4.1.3 Recirculation pump speed mismatch shall be maintained within..
- a. 5X of each other with core flow greater than or equal to 75K of rated core flow.
- b. 1GX of each other with core flow 1'ess than 75K of rated core flow.
APPLICA8ILITY: OPERATIONAL CONDITIONS 1" and 2"~>Le QgjI yacc
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ACTION: With the recirculation pump speeds different by more than the specified limits, either: aO Restore the recirculation pump speeds to within the specified limit within 2 hours, or 0 1
- b. Declare the recircu1ation 1'oop of the pump with the slower s'peed not in operation and take the ACTION required by Specification 3.4. l. 1.t.
SURVEILLANCE RE UIREMENTS 4.4.1.3 Recirculation pump speed mismatch shall be verified to be within the limits at least once per 24 hours.
*See Special Test Exception 3.10.4.
SUS/UEHANNA - UNIT 2 3)4 4-3
I ' SPECIAL TEST EXCEPTIONS 3/4.10.4 RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.10.4 The requirements of Specifications 3.4. l. l.!and 3.4.1.3 may be suspended for up to 24 hours for the performance of:
- a. PHYSICS TESTS,'rovided that THERMAL POWER does not exceed 5X of RATED THERMAL POWER,, or
- b. The Staitup Test Program.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2, during PHYSICS TESTS and the Startup Test Program. ACTION: ae With the. above specified time limit exceeded, insert all control rods.
- b. With the above specified THERMAL POWER limit exceeded during PHYSICS TESTS, immediately place the reactor mode switch in the Shutdown position.
SURYEILLANCE RE UIREMENTS
- 4. 10.4. 1 The time during which the above specified requirement has been suspended shall be verified to be less than 24 hours at least once per hour during PHYSICS TESTS and the Startup Test Program.
- 4. 10.4.2 THERMAL POWER shall be determined to be less than 5X of RATED THERMAL POWER at least once per'hour during PHYSICS TESTS.
E Amendment No.
REACTIVITY CONTROL SYSTEMS BASES 3/4.1. 3 CONTROL ROOS The specification of this section'nsure that (1) the minimum SHUTOOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3)- limit the potential effects of the.rod drop accident. The ACTION statements permit variations from the basic requirements but at the same time impose more restrictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and sc~am 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. Oamage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period hich is reasonable to determine the cause of the inoperability ahd at the same time prevent operation with a large number of inoperable control rods. C Control rods that are inoperable for other reasons are permitted'o be taken out of service provided that those in the nonfully"inserted position are consistent with the SHUTOOWN MARGIN requirements. ~ ~ The number of control rois permitted to be inoperable could be more than the eight allowed by he specifica ion, but the occurrence of eight inoperable .' rods could be indicative, of a generic problem and the reactor must be shutdown ~ for'nvestigation and resolution of the problem. The cont", ol rod system is designed to bring the reactor subcritical at a rate fast enough o prevent the MCPR from becoming less than . uring e limiting power transient analyzed in ection 15.1 of the FSAR. This analysis shows that the negative reactivity rates resulting from the scram with the average r'esponse of all the drives as given in the specifications rovide the required protection and MCPR remains greater than . . e occurrence of scram times longer then those specified should be viewed as an indication of a p bl ih di <<h f h is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem. 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 will 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 i'esult in less reactivity insertion on a scram than has been analyzed even though control rods, with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the. accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor. SUSQUEHANNA - UNIT 2 B 3/4 1-2
POWER DISTRIBUTION LIMITS BASES 3/4. 2. 3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR ~44&, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evalua-tion with the initial condition of the reactor being at the steady state opera~ing limit, it is required that the resulting HCPR 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 abnormal operational transient, the most limiting tran" sients have been analyzed to determine which result in the larges reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion,. and coolant
". mperature decrease. >> en = aC 0 he Safety i HCPR M~
The limiting transient yields the largest delta NCPR. the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in Figure 3.2.3-l. The evaluation of a given transient begins with the system initial
-=-". ~waters shown in FSAR Table 15.0-2 that are input to a GE-core dynamic behavior -". =nsient computer prof".am. The code ised to evaluate pressurization events is des'cribed in NEDO-24154 alod the program used in nonpressurization even s '.s. described in NEDO-10802 . The outputs of his program alono with the initial h1CPR form the input for'urther analyses of the thermally limiting i
bundle with the single channel transient thermal hydraulic TA code described in NEDE-25149 . The principal result of this'valuation is the reduction in HCPR caused by the transient. The purpose of the Kf facto~ of Figure 3. 2. 3-1 is to define operating limits at other than rated core flow conditions. At -less than lOOX of rated flow the required HCPR is the product of the HCPR and the Kf facto~. The Kf factors assure that the Safety Limit HCPR will not be violated during a flow increase transient resulting from a motor generator speed control failure. The Kf factors may be applied to both manual and automatic flow control modes.
'he Kf factor values shown in Figure 3.2.3-1 were devloped generically
. and are applicable to all BWR/2, BWR/3 and BWR/4 reactors.'he Kf factors were derived using the flow control line corresponding to RATED THERMAL POWER at
'rated core flow.
For the manual flow control mode, the Kf factors were calculated such that for the maximum flow rate, as limited by the pump scoop tube set point and the corresponding THERMAL POWER 'along the rated flow control line, the limiting bundle's relative power was adjusted until the HCPR changes with different core flows. The ratio of the HCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the Kf. 1 SUSQUEHANNA - UNIT 2 B 3/4 2-4,
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( H 3/4. 4 REACTOR COOLANT SYSTEM BASES 3/4.4. 1 RECIRCULATION SYSTEM THERMAL POWER, core flow, and neutron flux noise level limitations, are prescribed in accordance with the recommendations of General Electric Service Information Letter No. 380, Revision 1, "BWR Core Thermal Hydraulic Stability," dated February 10, 1984. An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable,'ut it does, in case of a design basis accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule .for significant degradation. goi q~~ i~p ~(~~t(o~- Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteri The limits will ensure an'dequate core flow . coastdown from either'recirculation loop following a LOCA. >
~ (gsCgx- g 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 loop. The loop temperature must also be wi'thin 50 F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper regions of the core, undue stress on the vessel would result temperature difference was greater than 145'F. if the 3/4.4. 2 SAFETY/RELIEF VALVES The safety valve function of the safety/relief valves 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 10 OPERABLE safety/ relief valves is required to limit reactor pressure'to within ASME III allowable values for the worst 0ase upset transient. Demonstration of the safety/relief. valve lift settings will occur 'only during shutdown and will be performed in accordance with the provisions of Specification 4.0. 5. C SUS(UEHANNA - UNIT 2 B 3/4 4-1 Amendment No. 2
X. Operation with one reactor recirculation loop inoperable has been evaluated and found acceptable, provided that the unit is operated in
'ccordance with Specif ication 3.4. 1 ..1. 2.
For single loop operation, the MAPLHGR limits are multiplied by a factor of 0.81. The multiplication factor is derived from LOCA analyses initiated from single loop operation; it maintains the same peak clad temperature margin to the licensing limit of 2200'F as in two loop operation. For single loop operation,, the RBM and APRM setpoints are ad)usted by a 7/ decrease in recirculation drive flow to account for the active loop drive flow that bypasses the core and goes up through the inactive loop )et pumps Surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive reactor vessel internals vibration. Surveillance on differential temperatures below the threshold limits on THERMAL POWER or recirculation loop flow mitigates undue thermal stress on vessel nozzles, recirculation pumps and the vessel bottom head during extended operation in the single loop mode. The threshold limits are those values which. will sweep up the cold water from the vessel bottom head. Y. In the case where the mismatch limits cannot be mqintained during the loop operation, continued operation is permitted in the single loop mode.
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