ML20072P897
| ML20072P897 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 11/15/1990 |
| From: | Quay T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20072P901 | List: |
| References | |
| NUDOCS 9011290210 | |
| Download: ML20072P897 (34) | |
Text
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UNITED sT ATEs
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l ENTERGY OPERATIONS. INC.
SYSTEM ENERGY RESOURCES. INC.
$_0VTH M1551SSIPP1 ELECTRIC POWER ASSOCI ATION MISSISSIPP1 POWER AND LIGHT COMPANY DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 73 License No. NPF 29 e
1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by the licensee dated June 8,1990, as revised August 15, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set fort'n in 10 CFR Chapter It The facility will cperate in conformity with the application, the B.
provisions af the Act, and the rules and regulations of the Comission1 There is reasonable assurance (i) that the activities authorized by C.
this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; The issuance of this amendment will not be inimical to the comon D.
defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 E.
of the Comission's regulations and all applicable requirements have been satisfied.
fFY $0$$ $((Q nc
2 2.
Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; andparagraph2.C.(2)ofFacilityOperatingLicenseNo.NPF.29ishereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 73, are hereby incorporated into this license. Entergy Operations, Inc, shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effecthe as of its date of issuance.
FOR THE NUCLEAR RE(ULATORY COMMIS$10N c\\
144 d
Theodore R. Quay, Acting Director Project Directorate IV.1 Division of Reactor Projects.111 IV, V, and Special Projects Office of Nuclear Reactor Regulation Attachment Changes tr Technical Specifie ns Date of 1
.ance: November 15, 1990
r ATTACHMENT TO LICENSE AMENDMENT NO. n FACILITY OPERATING LICENSE NO. NPF 29 DOCKET NO. 50 416 Replace the following pages of the Appendix A Technical Cpecifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
I REMOVE PAGEs INSERT PAGES 12 1-2 2-1 2-1 B 2-1 B21 B 2 la B 2-la B22 B 2-2 3/4 1-16 3/4 1-16 3/4 2-1 3/4 2 1 3/4 2-2 3/4 2-2 3/4 2-2a 3/4 2-2b 3/4 2-2c 3/4 2-2d 3/4 2-2e 3/4 2-3a 3/4 2-3b 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2 6 3/4 2-6 3/4 2-6a 3/4 2-7 3/4 2-7 3/4 2 7a 3/4 2-7a 3/4 2 7b b
3/4 2-7c 3/4 4 1 3/4 4-1 3/4 4-la 3/4 4-la 3/4 10 2 3/4 10-2 B 3/4 1-3 B 3/4 1-3 B 3/4 1-4 B 3/4 1-4 8 3/4 1-Aa B 3/4 1-4a B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2 4 B 3/4 2-4 B 3/4 2-4a B 3/4 2-5 B 3/4 2-5 B 3/4 2-6 B 3/4 2-6 B 3/4 2 7 B 3/4 2-7 B 3/4 2-7a B 3/4 4-1 B 3/4 4-1
?
DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel.
Normal movement of the SRMs, IRMs, LPRMs, TIPS, or special movable detectors is not considered to be CORE ALTERATION.
Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative
- position, CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the ANFB correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power, DOSE EQUIVALENT l-131 j
1.9 DOSE EQUIVALENT l-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, 1-133, 1-134, and 1-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in Table !!! of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."
DRYWELL INTEGRITY 1.10 DRYWELL INTEGRITY shall exist when:
a.
All drywell penetrations required to be closed during accident conditions are either:
1.
Capable of being closed by an OPERABLE drywell automatic isolation system, or 2.
Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.4-1 of Specification 3.6.4.
b.
The drywell equipment hatch is closed and sealed.
c.
The drywell airlock is in compliance with the requirements of Specification 3.6.2.3.
d.
The drywell leakage rates are within the limits of Specification 3.6.2.2.
e.
The suppression pool is in compliance with the requirements of Specification 3.6.3.1.
f.
The sealing mechanism associated with each drywell penetration; e.g., welds, bellows or 0 rings, is OPERABLE.
GRAND GULF-UNIT 1 1-2 Amendment No. 73 l
2.0 SAFETY LIMITS AND LIMITING SAFETY AYSTEM 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 reac vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.
APPLICABILITY: OPERATIONAL CONDITIONS I 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 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.
THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.09 during both two loop operation and single loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.
' APPLICABILITY: OPERATIONAL CONDITIONS I and 2.
ACTION:
With MCPR less than the above limits 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 SHUTOOWN 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 Specifi-cation 6.7.1.
REACTOR COOLANT SYSTEM PRESSURE
- I The reactor coolant system pressure, as measured in the reactor vessel I
2.1.3 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 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.
Amendment No.7l, GRAND GULF-UNIT 1 2-1
'.',2. 3 SAFETY LIMITS BASES
2.0 INTRODUCTION
The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs.
Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is cal:ulated to occur if the limit is not violated.
Because fuel damage is not directly observable, a step back approach is used to establish a Safety Limit for the MCPR.
MCPR greater than the applicable Safety Limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.
The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.
Although some 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 condi-tions and the Limiting Safety system Settings. While fis; ion product migration from cladding perforation is just as measurable " t;ist from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.
Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.
These conditions represent a significant departure from the condition intended by design for planned operation.
2.1.1 THERMAL p0WER. Low pressure or Low Flow The Advanced Nuclear Fuels Corporation (ANF) ANFB critical power correla-tion is applicable to the ANF core.
The applicable range of the ANFB correla-tion is for pressures above 585 psig and bundle mass flux greater than 0.25M1bs/
hr-fta.
For low pressure and low flow conditions, a THERMAL POWER safety limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig and below 10%
RATED CORE FLOW was justified for Grand Gulf cycle 1 operation based on ATLAS test data and the GEXL correlation.
The use of the GEXL correlation is 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 was established by other means. This was 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.
E GRAND GULF-UNIT 1 B 2-1 Amendment No. 73
?.1 SAFE 7Y LIM 175 BASES THERMAL POWER. Low Pressure or Low Flow (Continued)
Analyses show that with a bundle flow of 28 x IOS lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi.
1hus, the bundle flow with a 4.5 psi driving head will be greater than 28 x IOS lbs/hr.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indi-cate that the fuel assembly critical power at this flow is approximately 3.35 Ma't.
With the design peaking f actors, this cot responds 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.
j Overall, because of the design thermal-hydraulic compatibility of the ANF fuel designs with the cycle 1 fuel, this justification and the associated low pres-sure and low flow limits remain applicable for future cycles of cores contain-ing these fuel designs.
With regard to the low flow range, the core bypass region will be flooded l
~
at any flow rate greater than 10% RATED CORE FLOW. With the bypass region flooded, the associated elevation head is sufficient to assure a bundle mass flux of greater than 0.25 Mlbs/hr-ft2 for all fuel assemblies which can approach critical heat flux.
Therefore, the ANFB critical power correlation is appro-l priate for flows greater than 10% RATED CORE FLOW.
The low pressure range for cycle 1 was defined at 785 psig.
Since the ANFB correlation is applicable at any pressure greater than 585 psig, the cycle I low pressure boundary of 785 psig remains valid for the ANFB correlation.
GRAND GULF-UNIT I B 2-la Amendment No.73 l
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$AFETY LIMIT $
BASES 2.1.2 THERMAL POWER, Hich Pressure and High Flow The onset of transition boiling results in a decrease in heat transfer from the clad, elevated clad temperature, and the possibility of clad failure.
However, the existence of critical power, or boiling transition, is not a di-rectly observable parameter in an operating reactor.
Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution.
The mar-gin for each fuel assembly is characterized by the critical power ratio (CPR),
which is the ratio of the bundle power which would produce onset of transition boiling divided by the actual bundle power.
The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (MCPR).
The Safety Limit MCPR assures sufficient conservatism such that, in the event of a sustained steady state operation at the MCPR safety limit, at least 99,9's of the fuel rods in the core would be expected to avoid boiling transi-l tion.
The margin between calculated boiling transition (MCPR = 1.00) and the Safety Limit MCPR is based on a detailed statistical procedure whien considers the uncertainties in monitoring the core operating state and includes the effects associated with channel bow.
One specific uncertainty included in the safety limit is the uncertainty inherent in the ANFB critical power correlation.
ANF report XN NF 524(P), Rev. 2, " Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors," April 1989, including Supple-ment 1, describes the methodology used in determining the Safety Limit MCPR.
t The ANFB critical power correlation is based on a significant body of l
practical test data, providing a high degree of assurance that the critical l
power as evaluated by the correlation is within a small percentage of the actual critical power being estimated.
The assumed reactor conditions used in defining the safety limit introduce conservatism into the limit because bound-ing radial power factors and bounding flat local peaking distributions are l
l Still further con-used to estimate the number of rods in boiling transition.
servatism is induced by the tendency of the ANFB correlation to overpredict the l
l number of rods in boiling transition.
These conservatisms and the inherent accuracy of the ANFB correlation provide assurance that during sustained opera-l tion at the Safety Limit MCPR there would be essentially no transition boiling in the core.
l l
GRAND GULF-UNIT 1 B 2-2 Amendment No. 73
t REACTIVITY CONTROL SYSTEMS ROD PATTERN CONTROL SYSTEM 4
LIMITING CONDITION FOR OPERATION
)
3.1.4.2 The rod pattern control system (RPCS) shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2*#
ACTION a.
With the RPCS inoperable or with the requirements of ACTION b, below, not satisfied and with:
1.
THERMAL POWER less than or equal to the Low Power Setpoint, control rod movement shall not be permitted, except by a scram.
2.
THERMAL POWER greater than the Low Power Setpoint, control rod withdrawal shall not be permitted.
b.
OPERABLE control rod movement may continue by bypassing control rod (s) in the RPCS** provided that:
1.
With one control rod inoperable due to being immovable, as f
a result of excessive friction or mechanical interference, or known to be untrippable, this inoperable control rod may be bypassed in the rod action control system (RACS) provided that the SHUTDOWN MARGIN has been determined to be equal to or greater than required by Specification 3.1.1.
2.
With up to eight control rods inoperable for causus other than addressed in ACTION b.1, above, these inoperable tontrol rods 1
may be bypassed in the RACS provided that:
a)
The control rod (s) to be bypassed is inserted and the directional control valves are disarmed eithet:
1)
Electrically, or 2)
Hydraulically by closing the drive water and exhaust water isolation valves.
b)
All inoperable control rods are separated from all other inoperable control rods by at least two control cells in all directions.
c)
There are not more than 3 inoperable ca xrol rods in any RPCS group.
3.
Control rods may % bypassed in the Rod Action Control System (RACS) at anv time.
However, if THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER:
' See Special Test Exception 3.10.2
- Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RPCS prior to withdrawal of control rods for the purpose of bringing the reactor to criticality.
- Bypassing control rod (s) in the RPCS shall be performed under administrative control.
GRAND GULF-UNIT 1 3/4 1-16 Amendment No. 73
3/4.2 P0tfER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITINC CONDITION FOR OPERATION 3.2.1 During two loop cperation, all AVERAGE PLANAR LINEAR HEAT GENERATION l
RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figure 3.2.1-1.
l During single loop operation, the APLHGR for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limit shown in Figure 3.2.1-1 multiplied by 0.8.
APPLICABILITY:
OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
ACTION:
During two loop operation or single loop operation, with an APLHGR exceeding the limits of Figure 3.2.1-1 as corrected by the appropriate multiplication l
factor, 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 POWEL to less then 25% of RATED THERMAL POWER within the next 4 haurs.
SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the required limits:
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 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 j
operating with a LIMITING CONTROL ROD PATTERN for APLHGR, j
d.
The provisions of Specification 4.0.4 are not applicable.
GRAND GULF-UNIT 1 3/4 2-1 Amendment No. 73
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POSER DISTR 88UT20N LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR, MCPR, and MCPR, limits at indicated core flow THERMAL POWER, f
p and exposure, as shown in Figures 3.2.3 1, 3.2.3-2, and 3.2.3-3.
APPLICABILITY:
OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or l
equal to 25% of RATED THERMAL POWER.
ACTION:
With MCPR less than the applicable MCPR limits determined from Figures 3.2.3-1, 3.2.3-2, and 3.2.3-3, initiate corrective action within 15 minutes and restore l
MCPR 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 l
1ess than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.2.3 MCPR shall be determined to be equal to or greater than the applicable MCPR limits determined from Figures 3.2.3-1, 3.2.3-2, and 3.2.3-3:
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 l
least 15% of RATED THERMAL POWER, and 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 c.
operating with a LIMITING CONTROL ROD PATTERN for MCPR.
d.
The provisions of Specification 4.0.4 are not applicable.
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GRAND GULF-UNIT 1 3/4 2-4 Amendment No. 73 l
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POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE y
LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the limits shown in Figure 3.2.4-1 as multiplied by the smaller of either the flow-dependent LHGR factor (LHGRFAC ) of Figure 3.2.4-2, or the power-dependent LHGR factor f
(LHGRFAC ) of Figure 3.2.4-3.
p APPLICABILITY:
OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25*4 of RATED THERMAL POWER, s
ACTION:
With the LHGR of any fuel rod exceeding the limit of Figure 3.2.4-1, as cor-rected by the appropriate multiplication factor, initiate corrective action within 15 minutes and restore the LHGR to within the limit 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 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4
4.2.4 LHGR's shall be determined to be equal to or less than their allowable limits:
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 15% of RATED THERMAL POWER, 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 c.
operating on a LIMITING CONTROL R0D PATTERN for LHGR, and d.
The provisions of Specification 4.0.4 are not applicable.
GRAND GULF-UNIT 1 3/4 2-7 Amendment No. 73
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. GRAND GULF-UNIT 1 3/4 2-7c Amendment No. 73
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATIONLOOM LIMITING CONDITION FOR OPERATION 3.4.1.1 The reactor coolant recirculation system shall be in operation with either:
a.
Two recirculation loops operating with limits and setpoints per Specifications 2.2.1, 3.2.1, and 3.3.6, or l
b.
A single recirculation loop operatinn with:
1.
A volumetric loop flow rate less than 44,600 gpm, and 2.
The loop recirculation flow control in the manual mode, and 3.
Limits and setpoints per Specifications 2.2.1, 3.2.1, and 3.3.6.
l Operation is not permissible in Regions A, B or C as specified in Figure 3.4.1.1-1 except that operation in Region C is permissible during control rod withdrawals for startup.
APPLICABILITY:
OPERATIONAL CONDITIONS 1* and 2*.
ACTION:
a.
With no reactor coolant system recirculation loops in operation and the reactor mode switch in the run position, immediately place the reactor mode switch in the shutdown position, b.
With operation in Region A as specified in Figure 3.4.1.1-1, immediately place the reactor mode switch in the shutdown position, With operation in regions B or C as spe'cified in Figure S.4.1.1-1, c.
observe the indicated APRM, neutron flux noise level. With a sustained APRM neutron flux noise level greater than 10%
peak-to-peak of RATED THERMAL POWER, immediately place the reactor mode switch in the shutdown position.
d.
With operation in Region B as specified in Figure 3.4.1.1-1, immediately initiate action to either reduce THERMAL POWER by inserting control rods or increase core flow if one or more recirculation pumps are on fast speed by opening the flow control valve to within Region D of Figure 3.4.1.1-1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
With operation in Region C as specified in Fig'.:re 3.4.1.1-1, unless e.
operation in this region is for control rod withdrawals during startup, immediately initiate ection to either reduce THERMAL POWER or increase' core flow to within Region D of Figure 3.4.1.1-1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
f.
During single loop operation, with the volumetric loop flow rate greater than the above limit, immediately initiate corrective action to reduce flow to within the above limit within 30 minutes.
"See Special Test Exception 3.10.4.
GRAND GULF-UNIT 1 3/4 4-1 Amendment No. 73
~
LIMITING CONDITION FOR OPERATION (Continued) l-g.
During single loop operation, with the loop flow control not ir the I
manual mode,' place it in the manual mode within 15 minutes.
h.
During single loop operation, with temperature differences exceeding l
the limits of SURVEILLANCE REQUIREMENT 4.4.1.1.5, suspend the THERMAL POWER or recirculation loop flow increase, 1.
With a change in reactor operating conditions, from two recircula-tion loops operating to single loop operation, or restoration of two loop operation, the limits and setpoints of Specifications 2.2.1, l
3.2.1, and 3.3.6 shall be implemented within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or declare the associated equipment inoperable (or the limits to be "not satisfied"),
and take the ACTIONS required by the referenced specifications.
SURVEILLANCE REQUIREMENTS 1
L 4.4.1.1.1 At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor coolant recirculation system 1
shall b: verified to be in operation and not in Regions A, B or C as specified l
in Figure 3.4.1.11 except that operation in Region C is permissible during L
control rod withdrawals for startup, f
4.4.1.1.2 Each reactor coolant system recirculation loop flow control valve l
in an operating loop shall be demonstrated OPERABLE at least once per 18 months by:
Verifying that the control valve fails "as is" on loss of hydraulic a.
pressure at the hydraulic unit, and l
- b..
Verifying that the average rate of control valve movement is:
1.
Less than or equal to 11% of stroke per second opening, cnd 2.
Less than or equal to 11% of stroke per second closing.
4.4.1.1.3 During single loop operation, verify that the loop recirculation ~
L flow control in the operating loop is in the manual mode at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
4.4.1.1.4 During single loop operation, verify that the volumetric loop flow rate of the loop in operation is within the limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l 1
l 1
GRAND GULF-UNIT 1 3/4 4-la Amendment No.73
t SPECIAL TEST EXCEPTIONS 3/4.10.2 ROD PATTERN CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.10.2 The sequence constraints imposed on control rod groups by the rod pattern control system (RPCS) per Specification 3.1.4.2 may be suspended by means of the individual rod position bypass switches
- for the following tests:
a.
Shutdown margin demonstrations, Specification 4.1.1_.
b.
Control rod scram, Specification 4.1.3.2.
c.
Control rod friction measurements.
d.
Startup Test Program with the THERMAL POWER less than 10% of RATED THERMAL POWER, APPLICABILITY:
OPERATIONAL CONDITIONS I and 2.
ACTION:
With the requirements of the above specification not satisfied, verify that the RPCS is OPERABLE per-Specification 3.1.4.2.
SURVEILLANCE REQUIREMENTS 4.10.2 When the sequence constraints imposed on control rod groups by the RPCS are bypassed, verify; a.
Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to bypassing any sequence constraint 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 /> while any sequence constraint is bypassed, that movement of the control rods at less than or equal to 10% of RATED THERMAL POWER is limited to the established control rod sequence for the specified test, and Conformance with this specification and test procedures by a second L
b.
licensed operator or other technically qualified member of the unit I
technical staff.
l l
- Bypassing control rod (s) in the RPCS shall be performed under administrative l
control.
1 GRAND GULF-UNIT 1 3/4 10-2 Amendment No. 73
~
i REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued)
Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR. The overtravel position fea-ture provides the only positive means of determining that a rod is properly coupled and therefore this check must be performed' prior to achieving criti-cality after completing CORE ALTERATIONS that could have affected the control rod coupling integrity.
The subsequent check is performed as a backup to the initial demonstration.
In order to ensure that the control rod patterns can be followed and thenfore that other parameters are within their limits, the control rod position indication system must be OPERABLE.
The control rod housing support restricts the outward movement of a con-trol rod to less than 3 inches in the event of a housing failure.
The amount i
of rod reactivity which could be added by this small amount of rod withdrawai l
is less than a normal withdrawal increment and will not contribute to any dam-age to the primary
- coolant system.
The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
I The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system l
, components.
3/4.1.4 CONTROL ROD PROGRAM CONTROLS The rod withdrawal limiter system input power signal orginates from the first stage turbine pressure.
When operating with the steam bypass valves open, this tignal indicates a core power level which is less than the true
~
Consequently, near the low power setpoint and high power setpo 4 t core power.
of the rod pattern control system, the potential exists for nonconservati',e control red withdrawals.
Therefore, when operating at a sufficiently high power lesel, there is a small probability of violating fuel Safety Lir.its dur-ing a licensing basis rod withdrawal error transient. To ensure that fuel Safety Limits are not violated, this specification prohibits control rod with-drawal when a biased power signal exists and core power exceeds the specified level.
l Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident.
The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 10% of RATED THERMAL POWER, there 1. no possible rod worth l
which, if dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm. Thus requiring the rod pattern controller function to be OPERABLE when THERMAL POWER is less than or equal to 10% of l
RATED THERMAL POWER provides adequate control.
GRAND GULF-UNIT I B 3/4 1-3 Amendment No. 73
i REtCTIVITY CONTROL SYSTEMS e
BASES CONTR0t. D00 PROGRAM CONTROLS (Continued)
The RPCS piovides automatic supervision to assure that out-of-sequence rods will.iot be withdrawn or inserted.
A rod is out of sequence if it does not meet the criteria of the Banked Position Witholawal Sequence (Reference 1) l as described in the FSAR, The RPCS function is allowed to be bypassed in the Rod Action Control System (RACS) if necessary, for example, to insert an in-operable control rod, return an out-of-sequence control rod to the proper in-sequence position or move an in-sequence control rod to another in-sequence position. The requirement that a second qualified individual verify such bypassing and positioning of control rods ensures that the bases for RPCS limitations are not exceeded.
In addition, if THERMAL POWER is below the low power setpoint, additional restrictions are provided when bypassing control rods to ensure operation at all tirr.es within the basis of the control rod drop accident analysis, i
The baseline analysis of the rod drop accident is preser..ed in Section 15.4 of the FSAR and the techniques of the analysis are presented in Reference 1.
Analyses applicable to the current cycle are addressed in the appropriate cycle-specific documentation,
l The RPCS is also designed to automatically prevent fuel damage in the l
event of erroneous rod withdrawal from locations of high power density during higher power operation.
A dual channel system is provided that, above the low power setpoint, restricts the withdrawal distances of all non peripheral control rods.
This restriction is greatest at highest power levels.
3/4.1.5 STANDBY LIQUID CONTROL SYSTEM h
The standby liquid control system provides a backup capability for bring-i l
ing the reactor from full power to a cold, xenon-free shutdown, assuming that I
the withdrawn control rods remain fixed in the rated power pattern. To meet this objective it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core in approximately 90 to 120 minutes.
A minimum available quantity of 4530 gallons of sodium pentaborate solution coritaining a minimum of 5800 lbs. of sodium pentaborate is required to meet a shutdown requirement of 3%. There is an additional allowance of 165 ppm in the I
reactor core to account for imperfect mixing and leakage.
The time requirement was selected to override the reactivity insertion rate due to cooldown following the xenon poison peak and the required pumping rate is 41.2 gpm.
The minimum storage volume of the solution is established to allow for the portion below the pump suction that cannot be inserted.
The temperature requirement is neces-sary to ensure that the sodium pentaborate remains in solution.
1 l
1.
C.J. Paone, " Banked Position Withdrawal Sequence," GE Topical Report, l
NEDO-21231, January 1977.
1 GRAND GULF-UNIT 1 8 3/4 1-4 Amendment No. 73
s REACT!VITV CONTR01. SYSTEMS BASES STANDBY L10U10 CONTROL SYSTEM (Continued)
With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.
Relief valves are provided on the SLCS pump discharge piping to protect the SLCS pump and piping from overpressure conditions.
Testing of the relief valve setpoint and verifying that the relief valve does not open during steady state operation of the SLCS pumps demonstrates OPERABILITY of the relief valve.
The relief valves are ASME Class 2 valves and, as such, have a i 3% tolerance in the opening pressure from the set pressure, per the ASME Code (Section III -
Division 1 Subsection NC-7614.2(b), 1974 Edition).
Surveillance requirements are established on a frequency that assures a high reliability of the system.
Once the solution is established, boron con-centration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.
Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.
Compliance with the NRC ATWS Rule 10CFR50.62 has been demonstrated by means of tha equivatant control capacity concept using the plant specific minimum parameters.
This concept requires that each boiling water reactor must have a standby liquid control system with a minimum flow capacity and boron content equivalent in control capacity to 86 gpm for 13% weight sodium pentaborate solution (natural boron enrichment) used for the 251-inch diameter reactor vessel studied in NEDE-24222, Reference 2.
The described minimum sys-l
. tem parameters (82.4 gpm, 13.6% weight with natural boron enrichment) provides an equivalent control capacity to the 10CFR 50.62 requirement.
The techniques of the analysis are presented-in a licensing topical report NEDE-31096-P, l
Reference 3.
Only one subsystem is needed to fulfill the system design basis, and two subsystems are needed to fulfill ATWS rule requirements.
An SLCS subsystem consists of the storage tank, one divisional pump, explosive type valve, and associated controls, and other valves, piping, instrumentation, and controls necessary to prepare and inject neutron absorbing solution into the reactor.
2.
" Assessment of BWR Mitigation of ATWS, Volume II," NEDE-24222, December 1979.
3.
L. B. Claasen et al., " Anticipated Transients Without Scram, Response to l
NRC ATWS Rule 10CFR50.62," G. E. Licensing Topical Report prepared for the BWR Owners' Group, NEDE-31096-P, December 1985.
GRAND GULF-UNIT 1 B 3/4 1-4a Amendment No. 73
'. 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications 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 LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.
The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily 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 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits of Figure 3.2.1-1 are applicable to two loop operation.
For single-loop operation, a MAPLHGR limit corresponding to the product of the MAPLHGR, Figure 3.2.1-1, and 0.8 can be conservatively used to ensure that the PCT for single loop operation is bounded by the PCT for two loop operation.
The daily requirement for calculating APLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribu-tion shifts are very slow when there have not been significant power or control j
i l
l-l l
L GRAND GULF-UNIT 1 B 3/4 2-1 Amendment No. 73 1
3/4.2 POWER DISTRIBUTION LIMITS BASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued) rod changes.
The requirement to calculate APLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize.
The requirement for calculating APLHGR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that APLHGR will be known following a change in THERMAL POWER or power shape, that could place operation exceeding a thermal limit.
1he calculational procedure used to establish the APLHGR limits is based on a loss-of-coolant accident analysis.
The analysis was performed using calculational models which are consistent with the requirements of Appendix K to 10 CFR 50.
These models are described in reference 1.
l 3/4.2.2 [0ELETED)
GRAND GULF-UNIT 1 B 3/4 2-2 Amendment No.73 i
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 feel clad-ding integrity Safety Limit MCPR, and an analysis of abnormal operational tran-sients.
For any abnormal operating transient analysis evaluation with the initial condition of the reactor being t.t 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 abnormal operational 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 CPR.
When added to the Safety Limit MCPR, the required operating limit MCPR of Specification 3.2.3 is obtained.
The power-flow map of Figure B 3/4 2.3-1 defines the analytical basis for generation of the MCPR operating limits l
(References 2 and 3).
MCPR tperating limits are defined as functions of exposure (MCPR,), flow (MCPR ), ard power (MCPR ).
The limit to be used at a given operating state f
p is the higtest of these three limits.
The pt rpose of the MCPR, is to define operating limits for all anticipated exposures juring the Cycle.
The MCPR, limits are established for a set of exposure intervals. The limiting transients are analyzed at the limiting expo-sure for each interval.
The MCPR, operating limits are established based on the largest delta-CPR calculated at the limiting exposure and ensure that the MCPR safety limit will not be exceeded during the most limiting transient in each of the exposure i r.tcrval s.
and MCPR is to define operating limits at other The purpose of the MCPRf p
than rated core flow and power conditions for all :.gosures during the cycle.
l The MCPR s are established to protect the core from inadvertent core flow f
increases such that the 99.9% MCPR limit requirement can be assured. The ref-is a hypothesized erence core flow increase event used to establish the MCPRf slow flow runout to maximum, that does not result in a scram from neutron flux overshoot exceeding the APRM neutron flux-high level (Table 2.2.1-1 item 2).
Two flow rates have been considered. The maximum credible flow during a runout transient depends on whether the plant is in Loop Manual or Non Loop Manual operation. 'The result of a single failure or single operator error during Loop Manual operation is the runout of one loop because the two recirculation loops are under independent control.
Runout of both loops is possible during Non Loop Manual operation because a single controller GRAND GULF-UNIT 1 B 3/4 2-4 Amendment No. 73
P0tlER DISTR 1BUTION L1M1TS BASES MINIMUM CRITICAL POWER RATIO (Continued) regulates core flow. With this basis, the MCPR curves are generated from a f
series of steady state core thermal hydraulic calculations performed at several core power and flow conditions along the steepest flow control line.
In the actual calculations a conservative highly steep generic representation of the 10% steam flow rodline flow control line has been used.
Assumptions used in the original calculations of this generic flow control line were consistent with a slow flow ircrease transient duration of several minutes:
(a) the plant heat balance was assumed to be in equilibrium, and (b) core xenon concentration s
GRAND GULF-UNIT I B 3/4 2-4a Amendment No.73
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POWER D]STR1BUTICN LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) was assumed to be constant.
The generic flow control line is used to define several core power / flow states at which to perform steady-state core thermal-hydraulic evaluations.
Loop Manual and Non loop Manual modes of operation were analyzed.
Consistent with the single failure / single operator error criterion, one loop runout was postulated for Loop Hanual operation whereas two loop runout was postulated for Non Loop Manual operation.
The maximum core flow at loop
(;
runout was assumed to be 110% of rated flow.
Peaking factors were selected L
such that the MCPR for the bundle with the least margin of safety would not decrease below the MCPR Safety Limit.
l The MCPR is established to protect the core from plant transients other p
than core flow increase including the localized rod withdrawal error event.
Core power dependent setpoints are incorporated (incremental control rod with-drawal limits) in the Rod Withdrawal Limiter (RWL) System Specification (3.3.6).
These setpoints allow greater control rod withdrawal at lower core powers where m
core thermal margins are large.
However, the increased rod withdrawal requires higher initial MCPR's to assure the MCPR safety limit Specification (2.1.2) is not violated.
The analyses that establish the power dependent MCPR require-ments that support the RWL system are presented in Reference 4.
For core power l
below 40% of RATED THERMAL POWER, where the E0C-RPT and the reactor scrams on turbine stop valve closure and turbine control valve fa:t closure are bypassed, separate sets of MCPR limits are provided for high and low core flows to ac-countforthesignifi8antsensitivitytoinitialcoreflows.
For core power above 40% of RATED THERMAL POWER, bounding power-dependent MCPR limits were developed.
The abnormal operating transients analyzed for single loop operation are discussed in Reference 5 and the appropriate cycle-specific documents.
No change to the MCPR operating limit is required for single loop operation.
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 modera-tor void cont 6nt will be very small.
For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting McPR value is in excess of requirements by a considerable margin.
5 e
i-nt GRAND GULF-UNIT 1 B 3/4 2-6 Amendment No. 73 is i.i.-i.i.i.
POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)
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 requirement 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 requirement to calculate MCPR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a b
THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met af ter power distribution shifts while still a11otting time for the power distribution to stabilize.
The requirement for calculating MCPR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that MCPR will be known following a change in THERMAL POWER or power shape, that could place operation exceeding a thermal limit.
[
3/4.2.4 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.
7 The LHGR limits of Figure 3.2.4-1 are multiplied by the smaller of either the flow dependent LHGR factor (LHGRFAC ) or the power dependent LHGR factor f
=-
(LHGRFAC ) corresponding to the existing core flow and power state to ensure p
adherence to the fuel mechanical design bases during the limiting transient.
LHGRFAC 's are generated to protect the core from slow flow runout transients.
f Two curves are provided based on the maximum credible flow runout transient for either Loop Manual or Non-Loop Manual operation. The result of a single failure or single operator error during operation in Loop Manual is the runout of only one loop because both recirculation loops are under independent control.
Non-Loop Manual operational modes allow simultaneous runout of both loops because a single controller regulates core flow.
LHGRFAC 's are generated to p
protect the core from plant transients other than core flow increases.
The daily requirement for calculating LHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distri-bution shifts are very slow when there have not been significant power or control rod changes.
The requirement to calculate LHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still a11otting time for the power distribution to stabilize.
The requirement for calculating LHGR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that LHGR will be known following a change in THERMAL POWER or power shape that could place operation exceeding a thermal limit, e
Amendment No. 73 GRAND GULF-UNIT 1 B 3/4 2-7 P-
POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) i
References:
1.
XN-NF-80-19(A), Volume 2 " Exxon Nuclear Methodology for Bolling Water Reactors:
EXEM BWR ECCS Evaluation Model," Exxon Nuclear Company, September 1982.
2.
General Electric Company, " Maximum Extended Operating Domain Analysis,"
March 1986.
3.
AECM 86/0066, " Final Summary Startup Test Report 12," Letter, 0.D.
Kingsley, MP&L, to J. N. Grace, NRC, February 1986.
4.
XN-NF-825(P)(A), Supplement 2, "BWR/6 Generic Rod Withdrawal Analysis; MCPR for All Plant Operations Within the Extended Operation Domain,"
l 3
Exxoh Nuclear Company, October 1986.
l 5.
GGNS Reactor Performance Improvement Program, Single Loop Operation i
Analysis, General Electric Final Report, February 1986.
l l
l l
i i
l L
GRAND GULF-UNIT 1 B 3/4 2-7a Amendment No.73 l
' 3/4.4 REACTOR COLLANT SYSTEM
=
BASES 3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor core coolant recirculation loop inoperable has
^
been evaluated and found to remain within design limits and safety margins pro-vided certain limits and setpoints are modified.
The "GGNS Single Loop Opera-tion Analysis" identified the applicable fuel thermal limits and APRM setpoint modifications necessary to maintain the same margin of safety for single loop operation as is available during two loop operation.
Additionally, loop flow limitations are established to ensure vessel internal vibration remains within 1
limits. A flow control mode restriction is also incorporated to reduce valve wear as a result of automatic flow control attempts and to ensure valve swings i
into the cavitation region do not occur.
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-f-
accident, increase the blowdown area and reduce the capability of reficoding 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 per-
~
hm formance on a prescribed schedule for significant degradation.
During two loop operation, recirculation loop flow mismatch limits are in compliance with ECCS LOCA analysis design criteria. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.
In cases where the mismatch limits cannot be maintained, continued operation is per-mitted with one loop-in operation.
L The power / flow operating map is divided into four (4) regions.
Regions A and B are restricted from operations.
They include the operating area above the 80% rod-line and below 40% core flow.
Region C includes the operating area above the 80% rod-line and between 40% and 45% core flow.
Operation in Region C is allowed only for control rod withdrawals during startup for required fuel preconditioning.
Region D consists of the rest of the operating No core thermal-hydraulic stability related restrictions are applied to map.
Region D since the potential onset of core thermal-hydraulic instabilities is not predicted within Region D.
The definition of Regions A, B and C is based on BWR stability operational data and required operator actions.
Although a large margin to onset of insta-bility was observed in Regions A, B and C during GGNS stability tests for typical hi operating configuration, a conservative approach is adopted in the specification.
With no reactor coolant system recirculation loops in operation, and the reactor mode switch in the Run position, an immediate reactor shutdown is required.
Reactor shutdown is not required when recirculation pump motors are de-energized during recirculation pump speed transfers.
Upon entry to Region A an immediate reactor shutdown is required.
Upon entry to Region B or Region C, unless operation in Region C is for control rod withdrawals during startup, either a reduction of THERMAL POWER to below the 80% tod-line by control rod insertion or an increase in core flow to exit the region by opening the recirculation loop FCV is required.
Per the specification, the APRM neutron flux noise level should be observed while in Regions B and C.
In the unlikely event in which a sustained GRAND GULF-UNIT 1 B 3/4 4-1 Amendment No. 73
.