ML17255A801
| ML17255A801 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 05/01/1984 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17255A802 | List: |
| References | |
| NUDOCS 8405040226 | |
| Download: ML17255A801 (20) | |
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ROCHESTER GAS AND ELECTRIC CORPORATION DOCKET NO. 50-244 R.
E.
GINNA NUCLEAR POWER PLANT AMENDMENT TO PROVISl'ONAL OPERATING LICENSE Amendment No.
61 License No.
DPR-18 1.
The Nuclear Regulatory Commission
~the Commission) has found that:
A.
The application for amendment by Rochester Gas and Electric Corporation (the licensee) notarized December 20, 1983 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by 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 Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
8405040226 840501 PDR ADOCN 05000244 PDS
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C(2) of Provisional Operating License No.
DPR-18 is hereby amended to read as follows:
(2)
Technical S ecifications The Technical Specifications contained in Appendix A as revised through Amendment No. 61, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY CONMISSION Attachment; Changes to the Technical Specifications Date of Issuance:
gay. 1, 1984 en Ope 1 vl is N.
C ch ield, Chief ating Rea to s Branch 85 sion of Lic nsing
ATTACHMENT TO LICENSE AMENDMENT NO.
61 PROVISIONAL OPERATING LICENSE NO.
DPR-18 DOCKET NO. 50-244
'I g
ll I
I Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.
The revised pages contain the captioned amendment number and marginal lines which indicate the area of
",,'hanges.
REMOVE page 1-1 page 2.1-2 page 2.1-4 Figure 2.1-1 pages 2.3-2 through 2.3-3 page 3.1-4b pages 3.1-17 through 3.1-19 page 3.8-3 through 3.8-4 pages 3.10-3 through 3.10-4 Figures 3.10-2, 3.10-3 INSERT page 1-1 page 2.1-2 page 2.1-4 Figure 2.1-1 pages 2.3-2 through 2.3-3 page 3.1-4b pages 3.1-17 through 3.1-19 page 3.8-3 through 3.8-4 pages 3.10-3 through 3.10-4 Figures 3.10-2, 3.10-3
TECHNICAL SPECIFICATIONS DEFINITIONS The following terms are defined for uniform interpretation of the specifications.
Thermal Power The rate that the thermal energy generated by the fuel is accumulated by the coolant as it passes through the reactor vessel.
Reactor 0 eratin Modes Mode Reactivity ak k'oolant Temperature QF Refueling Cold Shutdown Hot Shutdown Operating
)0 140 avg T
< 200 T
R 540 avg T
w 580 avg
~f Any operation within the containment involving movement of fuel and/or control rods when the vessel head is unbolted.
Operable Capable of performing all intended functions in the intended manner.
Amendment No. Q, 61 Proposed
boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB) and pt this point there is a sharp reduction of the heat transfer coefficient which would result in high clad temperatures and the possibility of clad failure.
DNB is not, however, an observable parameter during reactor operation.
Therefore, the observable parameters, thermal power, reactor coolant temperature and pressure have been related to DNB through the W-3 and/or WRB-1 DNB correlation.
These DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.
The local DNB heat flux ratio, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indi-cative of the margin to DNB.
A minimum value of the DNB ratio, MDNBR, is specified so that during steady state operation, normal operational transients and anticipated transients, there is a
95%
probability at a
95% confidence level that DNB will not occur.'1)
The curves of Figure 2.1-1 represent the loci of points of thermal power, coolant system pressure and average temperature J
for which this minimum DNB value is satisfied.
The area of safe operation is below these lines.
2.1-2 Amendment No. 61, Proposed
Since it is possible to have somewhat greater enthalpy rise hot channel factors at part power than at full power due to the deeper control bank insertion which is permitted at part power, a
conservative allowance has been made in obtaining the curves in Figure 2.1-1 for an increase in F H with decreasing power levels.
Rod withdrawal block and load runback occurs before reactor trip set points are reached.
The Reactor Control and Protective System is designed to prevent any anticipated combination of transient conditions for reactor coolant system temperature, pressure and thermal power level that would result in there being less than a
95% probability at a
95%
confidence level that DNB would not occur.'3)
(1)
FSAR, Section 3.2.2 (2)
FSAR, Section 3.2.1 (3)
FSAR, Section 14.1.1 2.1-4 Amendment No.
61 March 30, 1976 Proposed
FIGURE 2.1-1 CORE ONB SAFETY LIMITS 2
LOOP OPERATION 668 6SB 645
- 648
- 635
- 638 625
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>sg~
2400 PSIA UNACCEPTABLE OPERATION o~ 628
~ 615 618
'85 leap~ Pgg~
595 598 585 588 57S ACCEPTABLE OPERATION 8
~ 1
.2
.3
~ 4
.5
.6
.7
.S
.9 1 ~
1 ~
1 1
2 POVER 1frnct1on of noe1noI )
Amendment No. g, g, 61 Proposed
d.
Overtemperature hT o
[ 1 2(
)
3
)
1 +
2S where hT
= indicated bT at rated power,
'F T
= average temperature,
'F T
= 573 5oF 1
P
= pressurizer
- pressure, psig P
= 2235 psig 1
Kl
= 1.20 K2
=.000900 K3
=.0209 tl
= 25 sec t2
=
5 sec and f (DI) is a function of the indicated differ-ence between top an'd bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests where qt and qb are the percent power in the top and bottom halves of the core respectively, and q
+ qb is the total core power in percent of rated power such that:
(i) for q q
less than
+21 percent, f (DI) =
0 2
~ 3~2 Amendment No.
, 6]
March 30, 1976 Proposed
(ii) for each percent that the magnitude of qt - qb is more positve than +21 percent, the bT trip set point shall be automatically reduced by an equivalent of 1.6 percent of rated power.
e.
Overpower hT
[K~ - K5(T-
)
K6,3S 1
]
(
)
where T
Tl K4 Ks K6 73 indicated dT at rated power,
'F average temperature,
'F indicated T avg at nominal conditions at rated power, 'F 1.077 0.0 for T<T 0.0011 for T>T 0.0262 for increasing T
0.0 for decreasing T
10 sec f(ar )
as defined in 2.3.1.2.d.
2
~ 3 3
Amendment No.
61 March 30, 1976 Proposed
3.1.1e5 Pressur'izer I
Whenever the reactor is at hot shutdown or critical the pres'surizer shall have at least 100 kw of heaters
~
li operable',a'nd a water level maintained between 12% and B7% of'level span.
If the pressurizer is inoperable due to !heagers or water level, restore the pressurizer to operable status within 6 hrs. or have the RHR N
"I system,"in',,operation within an additional 6 hrs.
Bases II
~
II
~
The plant is designed to operate with all reactor coolant loops all in operation and maintain the DNBR above the limit value during all normal 3.1-4b Change No.
Amendment No. g, P6, Q, 66, 61 Proposed
3.1.3 3.1.3.1 Minimum Conditions for Criticalit Except during low power physics tests, the reactor shall not be made critical at a temperature below 500'F, and if the moderate temperature coefficient is more positive than a.
5 pcm/'F (below 70 percent of rated thermal power) b.
0 pcm/'F (at or above 70 percent of rated thermal power) 3.1.3.2 In no case shall the reactor be made critical above and to the left of the criticality limit line shown on Figure 3.1-1 of these specifications.
3.1.3.3 Basis When the reactor coolant temperature is below the minimum temperature specified
- above, the reactor shall be subcritical'y an amount equal to or greater than the potential reactivity insertion due to depressurization.
Previous safety analyses have assumed that for Design Basis Events (DBE) initiated from the hot zero power or higher power conditi'on, the moderator temperature coefficient (MTC) was either zero or negative.
Beginning in Cycle 14, the safety analyses (1) (2) have assumed that a maximum MTC of +5 pcm/'F can exist up to 70%
power.
Analyses have shown that the design criteria can be satisfied for the DBE's with this assumption.'
At greater than (3) 70% power the MTC must be zero or negatively 3.1-17 Amendment No.
61 Proposed
The limitations on MTC are waived for low power physics tests to permit measurement of the MTC and other physics design parameters of interest.
During these tests special operating precautions will be taken.
Amendment No. '61 Proposed
The requirement that the reactor is not to be made critical above and to the left of the criticality limit provides increased assurance that the proper relationship between reactor coolant pressure and temperature will be maintained during system heatup and pressurization.
Heatup to this temperature will be accom-plished by operating the reactor coolant pumps.
If the specified shutdown margin is maintained, there is no I
possibility of an accidental criticality as a result of an increase in moderator temperature or a decrease of coolant pressure.
Reference (1)
FSAR Table 3.2.1-1 (2)
FSAR Figure 3.2.1-8 (3)
Safety Evaluation for R.
E. Ginna Transition to 14 x 14 Optimized Fuel Assemblies, Westinghouse Electric Corporation, November 1983.
- 3. 1-19 Amendment No. g 61 Proposed
to public health and safety.
Whenever changes are not being (1) made in core geometry one flux monitor is sufficient.
This permits maintenance of the instrumentation.
Continuous moni-toring of radiation levels and neutron flux provides immediate indication of an unsafe condition.
The residual heat pump is used to maintain a uniform boron concentration.
The shutdown margin as indicated will keep the core subcritical, even if all control rods were withdrawn from the core.
During refueling, the reactor refueling cavity is filled with approxi-mately 230,000 gallons of borated water'.
The boron concentration of this water at 2000 ppm boron is sufficient to maintain the reactor subcritical by'at least 5% hk/k in the cold condition with all rods inserted (best estimate of 10% subcritical),
and will also maintain the core subcritical even if no control rods were inserted into the reactor.
Periodic checks of refueling (2) water boron concentration insure the proper shutdown margin.
Communication requirements allow the control room operator to inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.
In addition to the above safeguards, interlocks are utilized during refueling to insure safe handling.
An excess weight interlock is
- 3. 8-3 Amendment No.
61 Proposed
provided on the lifting hoist to prevent movement of more than one fuel assembly at a time.
The spent fuel transfer mechanism can accommodate only one fuel assembly at a time.
In addition interlocks on the auxiliary building.crane will prevent the trolley from being moved over storage racks.containing spent fuel.
The operability requirements for residual heat removal loops will ensure adequate heat removal while in the refueling mode.
The requirement for 23 feet of water above the reactor, vessel flange while handling fuel and fuel components in containment is con-sistent with the assumptions of the fuel handling accident analysis.
References:
(1)
FSAR - Section 9.5.2 (2)
Reload Transition Safety Report, Cycle 14 (3)
FSAR - Section 9.3.1 3.8-4 Amendment No. P, g 61
average power tilt ratio shall be determined once a
day by at least one of the following means:
a.
Movable detectors b.
Core-exit thermocouples Power distribution limits are expressed as hot channel factors.
At all times, except during low power physics tests the hot channel factors must meet the following limits:
F (Z)
= (2.32/P)*K(Z)
Q F (Z)
= 4.64*K(Z)
F hH
= 1.66
[1 +.3(1-P)]
forP
>.5 for P
.5 for 0
$ P$ 1.00 where P is the fraction of rated power at which the core is operating, K(Z) is the function given by Figure 3alON3, and Z is the height in the core.
The measured F
shall be increased by gree percent to yield F
. If the measured F
or Fh exceeds the limitin%I value, with due a118wance For measurement error, the maximum allowable reactor power level and the Nuclear Overpower Trip set point shall be reduced on percent for each percent which Fh or F exceeds the limiting value, whichever is mor5 restrictive.
If the hot channel factors cannot be reduced below the limiting values within one day, the Overpower hT trip setpoint and the Overtemperature hT trip setpoint shall be similarly reduced.
Except for physics tests, if the quadrant to average power tilt ratio, exceeds 1.02 but is less than 1.12, then within two hours:
a.
Correct the situation, or b.
Determine by measurement the hot channel factors, and apply Specification 3.10.2.2, or c.
Limit power to 75% of rated power.
3.10-3 Amendment No. g, tLO 61 Proposed
If the quadrant to average power tilt ratio exceeds 1.02 but is less than 1.12 for a sustained period of more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> without known cause, or if such a
tilt recurs intermittently without known cause, the reactor power level shall be restricted so as not to exceed 50% of rated power.
If the cause of the tilt is determined, continued operation at a power level consistent with 3.10.2.2
- above, shall be permitted.
Except for physics test, if the quadrant to average power tilt ratio is 1.12 or greater, the reactor shall be put in the hot shutdown condition utilizing normal operating procedures.
Subsequent operation for the purpose of measuring and correcting the tilt is per-mitted provided the power level does not exceed 50% of rated power and the Nuclear Overpower Trip "set point is reduced by 50%".
Following any refueling and at least every effective full power month thereafter, flux maps, using the movable detector
- system, shall be made to confirm that the hot channel factor limits of Specification 3.3.0.2.2 are met.
The reference equilibrium indicated axial flux difference as a function of power level (called the target flux difference) shall be measured at least once per equivalent full power quarter.
The target flux difference must be updated at least each equivalent full power month using a measured value or by linear interpolation using the most recent measured value and the predicted value at the end of the cycle life.
Except during physics tests, control rod exercises, excore detector calibration, and except as modified by 3.10.2.9 through 3.10.2.12, the indicated axial flux difference shall be maintained within 25% of the target flux difference (defines the target band on axial flux difference).
Axial flux difference for power distribution control is defined as the average value for the four excore detectors.
If one excore detector is out of service, the remaining three shall be used to derive the average.
Except during physics tests, control rod exercises, or excore calibration, at a power level greater than 90 percent of rated power, if the indicated axial flux difference deviates from its target band.
The flux di ference shall be returned to the target band immediately or the reactor power shall be reduced to a level no greater than 90 percent of rated power.
3.10-4 Amendment No. +
61 Proposed
CL D D I M <<C CL UJ UJ M CL oo 0 D D LU P-o 0
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D (ap g) gqca~qoyag
- paacnbag u~op>>QS 3. 10-12
1.50~
FIGURE 3.10-3 1.2500 NORNLIZED AXIAL DEPENDENCE FACTOR FOR Fq VS ELEVATION 1.0000 0.7500
H CD H
4J " 0. 000 0.2500 TOTAL FO 2.320 CORE HEIGHT 0.000 6.000 10.800 12.000 K(Z) 1.000 1;000 0.940 0.64 7 0.0 CD CD CD CD Al CD CD CD
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~D CU CORE HE! GHT (FT) amendment No. fg, gp, 6i PROPOSED