ML20246F680
| ML20246F680 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 06/30/1989 |
| From: | Berkow H Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20246F683 | List: |
| References | |
| NUDOCS 8907130356 | |
| Download: ML20246F680 (13) | |
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UNITED STATES
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g NUCLEAR REGULATORY COMMISSION 5
,j WASHINGTON, D. C. 20555
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VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC C00 PEP.ATivE DOCKET NO. 50-339 NORTH ANNA POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. iou License No. NPF-7 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Virginia Electric and Power Company, et al., (tne licensee) dated June 17, 1987, complies with the j!'indards and requirements of the Atomic Energy Act of 1954, as amended sa Act), and the Comission's rules and regulations set forth in IJ 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 Comission; C.
There is reasonable assurance (1) 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the common l
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 Comission's regulations and all applicable requirements have been satisfied.
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8907130356 890630*
DR ADOCK 0500 9
. 2.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this licente amendment, and paragraph 2.C.(2) of Facility Operating License No. NFF-7 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Apr.endice.s A and B, as revised through Amendment No.100, are nereby incorporated in the license. The licensee shall operate the fac4lity in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of ir,suance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY C01941SSION h
/
H bert N. Berkow, Director Project Directorate 11-2 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: June 30, 1989
TO FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-339 Replace the following pages of the Appendix "A" Technical Specifications with-the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
.Page B2-1 3/4 1-5 3/4 1-6 3/4 3-10 B3/4 1-2 83/4 2-5 B3/4 2-6 a
)
l 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure fr'om nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through a correlation'. The.DNB~ correlation l
has been develop' d to predict the DNB flux'and the location of DNB for axially e
uniform and non-uniform heat flux distributions.
The local DNB. heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNQ of the limiting rod during Condition I i nd II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit.
In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95% probability that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.
As an additional criterion, meeting the DNBR limit also ensures that at least 99.9%
c:
.he' core avoids the onset of DNB when the plant is at the DNBR limit.
The curves of Figures 2.1-1, 2.1-2, and 2.1-3 show the loci of points of THERMAL POWER, Reactor Coolant System pressur and, average temperature for which the minimum DNBR is no less than the design limit DNBR, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.
NORTH ANNA - UNIT 2 B 2-1 Amendment No. 77, 100
l_
SAFETY LIMITS BASES The curves are based on an enthalpy hot, channel factor Ffg, of 1.49 and a j
reference cosine with a peak f 1.55 for axial power shape.
An allowance is included for an increase in at reduced power based on the expression:
l.
(H = 1.,49 { 1+0. 3 (1-P)]
where P is the fraction of RATED THERMAL POWER 1'
These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdtawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f(delta I)' function of the Overtemperature trip.
When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature AT trips will reduce the setpoints to provide protection.
consistent with core safety limits.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE i
The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
l The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure.
The Reactor Coolant System piping, l
valves and fittings, were initially designed to ANSI B 31.1 1967 Edition and ANSI B 31.7 1969 Edition (Table 5.2.1-1 of FSAR) which permito a maximum transient pressure of 120% (2985 psig) of c mponent design pressure.
The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code regoirements.
The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.
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l NORTH ANNA - UNIT 2 B 2-2 Amendment No. 8 f,71
)
REACTIV'ITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT I
LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be:
For the al rods withdrawn, beginning of core life condition
<0.6 x 10-} ak/k/'F below 70 percent RATED THERMAL POWER a.
~0.0 x 10-4 ok/k/*F at or above 70 percent RATED THERMAL POWER.
b.
Less negative than -5.0 x 10-4 Ak/k/'F for all the rods withdrawn, l
end of core life at RATED THERfiAL POWER.
APPLICABILITY:
Specification 3.1.1.4.a - MODES 1 and 2* only#
Specification 3.1.1.4.b - MODES 1, 2 and 3 only#
ACTION:
'~
a.
With the MTC more pos'tive than the limit of 3.1.1.4.a above, operations in MODES 1 and 2 may proceed provided:
1.
Control rod withdrawal limits are established and maintained sufficient to restore the MTC to within its limit,within 24 l
hours or be in HOT STANDBY within the.next.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. These I
withdrawal limits;shall be in addition to the insertion limits of Specification 3.1.3.6.
2.- The co!. trol rods are maintained within the withdrawal limits established above until subsequent measurement verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.
3.
Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition, b.
With the MTC more negative than the limit of 3.1.1.4.b above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- With Keff >1.0
- See Special Test Exception 3.10.3 NORTH ANNA - UNIT 2 3/4 1-5 Amendment No. 47,U, U,100 l
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REACTIVITY CONTROL SYSTEMS MODERITOR TEMPERATURE COEFFICIENT SURVEILLANCE REQUIREMENTS 4.1.1.4 Th6 MTC shall be determined to be within its limits during each fuel' cycle as follows:
a.
The MTC shall be measured and compared to the BOL Limit of
. Specification 3.1.1.4.a above, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel; loading.
l.'
b.
The MTC shall be measured at any THERMAL POWER and compared to -4.0 x 10-4 delta k/k/*F (all rods withdrawn, RATED THERMAL-l POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm.
In the event'this comparison indicated the MTC is more negative than -4.0 x 10-4 ' delta k/k/*F, l
the MTC shall be remeasured, and compared to the E0L MTC limit of specification 3.1.1.4.b, at 1 the remainder of the fuel cycle.99qt once per 14 EFPD during t'l r
(1) Once the equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER condition) is 60 ppm or less, further measurement of the MTC in accordance with 4.1.1.4.b may be suspended providing that the 4
measured MTC at an equilibrium boron concentration of <60 ppm is less negative than -4.7 x 10- ak/k/*F.
1 NORTH ANNA - UNIT 2 3/4 1-6 Amendment No. $J, 72,100
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3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTOOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions Lre controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
i SHUTDOWN MARGIN requirements vary throughout core life as a function of fueldepletion,P.CSboronconcentration,andRCST,yfn.
The most restrictive condition occurs at EOL, with T,yg at no load operat g temperature, and is associated with a postulated steam line break accident and resu1 ting uncontrolled RCS cooldown.
In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.77% delta k/k is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T,yg less than 200*F, the reactivity transients resulting from a postulated steam line break cooldown are minimal.
3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 3000 GPM, as provided by either one RCP or one RHR pump as required by Specification 3.4.1.1, provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System.
A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System 1
volume of 9957 cubic feet in approximately 30 minutes.
The reactivity change rate associated with boron reductions will therefore be within the capability for operator recognition and control.
The requirement that certain valves remain closed at all times except during planned boron dilution or makeup, activities provides assurance that an inadvertent boron dilution will not occur.
NORTH ANNA - UNIT 2 B 3/4 1-1
l.
3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.4' MODERATOR TEMPERATURE COEFFICIENT.(MTC)
The limitations on MTC are provided to ensure that the value of this f
coefficient remains within the limiting conditions assumed for this parameter in the FSAR accident and transient analyses.
The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparis6n.-
The most. negative MTC value was obtained by incrementally correcting"the MTC used in the FSAR analyses to nominal operating conditions.
These corrections involved adding the1 incremental change in the.MTC. associated with 6 core condition of Bank D inserted to an all rods withdrawn condition and an incremental change in MTC to account for measuremer) 6ncertainty at RATED THERMAL POWER
'10jitions. These corrections result in the lig'iting MTC value'of -5'.0 x con delta k/k/"F. The MTC ulue of -4.0 x 10- delta' k/k/ F represents a -
conservative value (with c,rrections for burnup and soluble boron) at a core condition of 300 ppm equilnri6m boro, concentration and js obtained by making these corrections to the'11miti.y MTC value of -5.0 x 10.,delfa k/k/ F.
~ Once the equilibrium boron concentration falls below about 60 ppm, dilution operations take an extended amount of time and reliable MTC measurements become more difficult to obtain due to the potential for fluctuating core conditions over the test interval.
For this reason, MTC measurements may be suspended provided the measured MTC value at an equilibrium full power boron concentration
<60 ppm is less negative than -4.7 x 10-4 delta k/k/'F. The difference between this value and the limiting MTC value of-5.0 x 10-4 delta k/k/*F conservatively bounds the maxiiiu credible change in MTC between the 60 ppm equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER conditions) and the licensed end of cycle, including the effect of boron concentration, burnup, and end-of-cycle coastdown.
The surveillance requirements for measurement of the MTC at the beginning and near the end of each fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.
3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541*F. This limitation is required to ensure 1) the moderator temperature coefficient is within its
{
analyzed temperature range, 2) the protective instrumentation is within its normal operating range, and 3) the P-12 interlock is above its setpoint, 4) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and
- 5) the reactor pressure vessel is above its minimum RT temperature.
NDT NORTH ANNA - UNIT 2 B 3/4 1-2 Amendment No./A,100
POWER DISTRIBUTION LIMITS BASES When F" is measured, 4% is the appropriate experimental error allowance for a full k$re map taken with the incore detection system.
The specified limit for F Normal operation will result inameasur$$cgntainsa4%errorallowance.
F less than or equal to 1.49.
The 4% allowance is based on the following considerations:
abnormal perturbations in ghe radial power shape, such as from a.
rod misalignment, effect F more directly than F,
aH q
i b.
although rod movement has a direct influence upon limiting F to withgn its limit, such control is not readily available th limit FaH, and c.
errors in prediction for control power shape detected during startup physics tests can be compensated for in F0gyrestricting axial flux distributions. This compe isation for 'F is less aH readily available.
Fuel rod bowing reduces the value of the DNB ratio. Credit is available to offset this reduction in the margin available between the safety analysis design DNBR value (1.46 for Virginia Electric and Power Company statistical methods) and the l'imiting design DNBR value (1.26.for Virginir Electric and Power Company statistica1' methods). A disc ~ussion of the rod bow penalty is presented in the FSAR.
The hot channel factor F M(Z) is neasured periodically and increased by l
n a cycle and height dependent power factor, N(Z), to provide assurance that the limit on the hot channel factor, Fg(Z), is met. N(Z) accounts for the non-equilibrium effects of normal operation transients and was determined from expected power control maneuvers over the full range of burnup condi-tions in the core. The N(Z) function for normal operation is provided in the Core Surveillance Report per Specification 6.9.1.7.
3/4.2.4 00ADRANT p0WER TILT RATIO I
The quadrant power tilt ratio limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis.
Radial power distribution measurements are made during startup testing and periodically during power operation.
The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts.
The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod.
In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the power g
by 3 percent for each percent of tilt in excess of 1.0.
NORTH ANNA - UNIT 2 B 3/4 2-5 Amendment No.15,55,M,77,100 i
l POWER DISTRIBUTION LIMITS BASES l
For_ purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of 4 symmetric thimbles. The two sets of 4 symmetric thimbles is a unique set of 8 detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8.
3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain -
a minimum DNBR greater than the design limit throughout each analyzed transient.
Measurement uncertainties must be accounted for in the DNB design margin.
l The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters thru instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.
NORTH ANNA - UNIT 2 B 3/4 2-6 Amendment No. 77, 6A /J,100 l
1