ML20077K547
| ML20077K547 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/31/1982 |
| From: | STONE & WEBSTER, INC. |
| To: | |
| Shared Package | |
| ML20077K532 | List: |
| References | |
| NUDOCS 8301130104 | |
| Download: ML20077K547 (45) | |
Text
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ENCLOSURE 2 NORTH ANNA UNITS 1 & 2 587.8'F REACTOR COOLANT SYSTEM
- STONE & WEBSTER / BOP
. SAFETY EVALUATION
SUMMARY
t STONE & WEBSTER ENGINEERING CORPORATION OCTOBER 1982
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TABLE OF CONTENTS A. Objective B. Conclusions C. Accident Analysis and Environmental Qualification D. Transient Analysis E. Pipe Stress and Supports F. Major Equipment Supports and Pipe Ruture Restraints G. Structures -
H. BOP Systems and NSSS Interfaces I. Heat Balance Calculations J. Control Systems and Instrumentation K. Electrical Systems
- L. Review of Technical Specifications e
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l A. OBJECTIVE The objective of this report is to provide a technical basis for determining that the proposed 7.5'F increase in reactor coolant system Tavg does not involve an unreviewed safety question in accordance with the requirements of 10CFR50.59. This review is limited to systems within Stone & Webster Eng-ineering Corporation's original scope of work. The NSSS and Turbine-Gener-ator review has been performed by Westinghouse and is documented as Enclosure 1.
Our evaluatica used the following parameters which bound or are equiv.alent to the propos,ed uprated conditions:
Main Steam Pressure 100% Power 930 psia Main Steam Temp. No-Load 547*F Main Steam Pressure No-Load 1020 psia RCS Tavg 587.8'F 0
Steam Flos 10 lb/hr Total 12.25 Reactor Power MWg 2775.
B. CONCLUSION The proposed change in reactor coolant system average temperature has been.
reviewed and evaluated with respect to the following:
- 1. Accident Analysis and Environmental Qualification
- 2. Pipe Stress and Supports
- 3. Major Equipment Supports and Pipe Rupture Restraints
- 4. Structures
- 6. Heat Balance Calculations
- 7. Control Systems and Instrumentation
- 8. Electrical Systems
- 9. Review of Technical Specifications 2
Based on the results of our review it has been concluded that the proposed 7.5*F increase in Tavg does not represent an unreviewed safety question as defined in 10CFR50.59. The summary of the analyses related to the above are attached.
- 1. It has been determined that the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the Safety Analysis Report has not been increased.
- 2. The possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report has not been created.
- 3. The margin of safety as defined in the basis for any Technical Speci-fication has not been reduced.
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C. ACCIDENT ANALYSIS AND ENVIRONMENTAL QUALIFICATIONS
- 1. Containment Loss of Coolant Accident (LOCA)
An analysis has been performed to investigate the effect of the 7.5*F uprate on containment integrity and Net Positive Suction Head Avail-able (NPSHA) for the Recirculation Spray (RS) and Low Head Safety In-jection (LHSI) pumps. The following are the results:
' . 1. Containment Integrity Table 1 shows the effect on containment peak pressure, subatmos-pheric peak pressure and depr'essurization time due to the uprate.
The pressures increase slightly but are still within the North Anna acceptance criteria and containment design pressure rating.
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- 2. NPSHA for RS and LHSI Pumps Table 2 shows that there is no decrease in NPSHA for the inside and outside RS pumps. There is a slight decrease in the LHSI pump l NPSHA but it remains above NPSH requirements at design flow.
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- 2. Containment Main Steam Line Break (MSLB) Analysis The basis of the steam line break analysis is the full guillotine main steam line break at the no-load (hot shutdown) condition. The no-load Reactor Coolant System and Steam Generator temperature and pressure remain unchanged subsequent to the uprate. Although there are additional energy releases at power operating conditions due to increased (uprated) Steam Generator pressure and tempera-ture, the no-load condition remains the limiting case. There-l fore the Main Steam Line Break post-accident conditions remain j as previously analyzed.
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- 3. Subcompartment Analysis Subcompartment analyses were performed and are documented in the UFSAR for the reactor cavity, steam generator cubicle and pres-surizer cubicle. Our calculations have confirmed that for the subcooled -
reactor coolant system, mass and energy releases decrease with increased reactor coolant temperature. The analyses documented in the UFSAR are therefore bounding for the uprate.
s.
- 4. Equipment Qualification
- 1. Inside Containment Equipment qualification inside the containment is based on the Main Steam Line Break and LOCA post-accident environ-mental ccaditions. Containment design pressure is used for qualification. The current Main Steam Line Break and LOCA analyses are either unchanged or negligibly impacted for the uprate conditions
,, as discussed in Sections 1 and 2. Post-accident containment pres-
- sure remains below the design pressure.
- 2. Outside containment Post-accident environments outside containment which are used to generate equipment qualification envelopes are based on the following high energy line breaks:
- a. Primary System Branch Line Break The Letdown Line Break is part of the basis for the environmental qualification in the charging pump cubicle in the auxiliary building. The letdown line temperature increases one degree above the normal operating temper-ature used in the original analysis. The resulting environmental temperature due to a postulated line break is not significantly affected by this change.
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- b. Secondary System Break The Main Steam Line Break affects the Main Steam Valve House, the Service Water Valve Pit and the Turbine Building.
The Main Steam Valve House environmental envelope is based on the no-load power condition which is unaffected by the up-rate. Equipment qualification temperature and pressure in the Service Water Valve Pit and Turbine Building is limited by the Turbine Building siding pressure retaining capability.
Any change to break effluent due to the uprate has no effect on this pressure or temperature, and therefore, on equipment qualification.
- c. Auxiliary Steam Line Break This break affects the Auxiliary Building and the Service Water Valve Pit. The releases are based on the Auxiliary Steam Line relief valve pressure setting which is unchanged by the uprate. Additionally, the Auxiliary Steam System pressure is controlled by a pressure reduction valve tied into the Main Steam header. The increased Main Steam Oper-ating pressure will not affect the operation of this valve and therefore the Auxiliary Steam System pressure will remain unchanged.
- d. Steam Generator Blowdown Line Break t This affects the Pipe Tunnel and the Auxiliary Building. The releases are based on the bounding condition of no-load steam generator pressure which is unchanged by the uprate.
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1 Table 1 LOCA Containment Integrity Analysis Results 7.5'F Current Uprate Required Containment Peak Pressure (psig.) 40.27 40.62 <45.
Containment Depressurization Time .
(Sec.) 3300 3320 <3600 Containment Subatmospheric Peak (psig.) .12 .11 <0.
I Table 2 LOCA NPSH Analysis Results 7.5*F Required Current Uprate NPSH*
Inside Recire Spray Pumps (ft.) 11.7 11.7 8.8 Outside Recire Spray Pumps (ft.) 17.0 17.0 9.7 Low Head Safety Injection Pumps (ft.) 15.6 15.2 13.4
- At design flow 7
r D. TRANSIENT ANALYSIS As a plant operational consideration, the 50% load rejection and turbine trip capability is being investigated. Any changes in load rejection cap-abilities will be evaluated prior to uprating implementation and be in-corporated in a subsequent UFSAR amendment as necessary.
E. PIPE STRESS AND SUPPORTS For all systems within the scope of the uprating program, the appropriate maximum stress levels and fatigue analysis results have been reviewe,d for the plant uprate conditions. This review indicates that in all cases the associated piping stress and fatigue limits will not be exceeded as a result of the uprate. The following systems were evaluated for the uprated conditions:
System Safety Class Main Steam 2 Feedwater 2 Reactor Coolant 1 RC Loop Letdown 1 RC Loop Excess Letdown 1 Pressurizer Spray 1 Residual Heat Removal 1,2 Low Head Safety Injection 1 High Head Safety Injection 1 CVCS Seal Water Inlet 1 CVCS Seal Water Outlet 1 RC Loop Fill 1 Resistance Temperature Detection 1 8
System Safety Class RC Loop Drain 1 RC Loop Charging 1 Steam Generator Blowdown 2 Steam Generator Wet Layup 2 Component Cooling to RC Pumps 2
! The uprate also has minimal impact on the existing stress analysis of the unmodified Resistance Temperature Detector (RTD) lines in the Reactor Coolant System. Because the Reactor Vessel Level Indication System (RVLIS) now ties into the RTD lines, a review of portions of these lines will be necessary. This review will be performed prior to implementation of the uprate and it is expected that the usage factors and stresses re-sulting from the RVLIS tie-in will not exceed code allowables.
l The above systems were reviewed with respect to pipe stress and the ade-quacy of pipe support designs. The maximum expected increase in the ther-mal expansion due to the uprate is approximately 2%. Considering that pipe support design loads include several loacing components other than thermal,
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it is concluded that, with the exception of the Unit 2 Main Steam Mono-Ball supports identified during the 2.5'F uprate program review, all existing pipe support and equipment nozzle design loads will remain valid.
The Mono-Ball supports will be modified to accommodate both the 2.5*F and 7.5'F uprating, prior to implementing the 2.5*F uprating. ,
A review of safety related piping for fatigue effects of thermal transients during the 2.5'F program review revealed that only the reactor coolant letdown line required reanalysis. The 2.5 F increase in Tavg subjected the letdown line to a more severe reinitiation of letdown flow transient chan was originally l
used as a design basis. The results of the analysis showed that the fatigue effects on the piping would remain acceptable subsequent to both the proposed 2.5*F and 7.5 F uprates. The stress report will be revised to include this information l
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It has been determined that previously calculated break points in piping systems, used for the design of rupture restraints, jet shields and large pipe supports will remain unchanged as a result of the uprating. The points of maximum stress within any piping system remain unchanged as the stress increases due to the uprating are uniform.
In conclusion, all existing pipe stress and support analyses within the scope of this evaluation have been determined to remain valid under the conditions of the 7.5"F uprate in Reactor Coolant System Tavg.
F. MAJOR EQUIPMENT SUPPORTS AND PIPE RUPTURE RESTRAINTS Design calculations for major equipment supports, seismic tanks, vessels, pipe rupture restraints and shields and miscellaneous mechanical equipment were reviewed with the uprated system parameters for 100% power and no-load conditions.
Approximately 450 calculations were reviewed. Of these calculations, a number did not use system press ires and temperatures which bounded the uprated conditions. Each one of those calculations was reviewed with respect to the uprated conditions. Only one of those calculations j was found to have revised results. The remaining calculations were deter-mined to be acceptable at the upra*ed conditions. The revised results con-sisted of an increased loading to a radial wall in the containment due to a postulated Main Feedwater pipe break. This loading is used as input to a structural calculation which was reviewed for the new loads due to the uprate. The results of the review are in Section G. below.
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l The data used to assess the effect of LOCA loading on major equipment supports remain applicable for both the 2.5*F and 7.5 F uprated conditions. Our evaluation concluded that an increase in Reactor Coolant System temperature results in a decrease in LOCA loadings with the frequency content of those j loadings being unchanged. The amplitudes and time history of the LOCA loads data previously supplied by Westinghouse was verified as being applicable by Westinghouse.
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G. STRUCTURES The only structural loads subject to change as a result of the uprate were those resulting from system parameter changes such as pressure, temperature and flow and subsequent postulated pipe breaks as developed in the review in Section F above, those resulting from peak containment pressure subsequent to a LOCA and subcompartments pressurizations reviewed in Section C above.
The new post LOCA peak containment pressure is less than containment design pressure and subcompartment pressures decrease.
Only one structural calculation (referenced in Section F above) was further reviewed because of increased loads from the uprating. This review was involved with the impact of a Main Feedwater pipe in the containment on a radial wall during a postulated break. The revised calculation showed that the structural integrity of the Reactor Containment would not be impacted by the postulated pipe break. A review of the expected spalling (concrete debris) zones was performed in order to determine if the spalling would have an adverse impact on any safety related components that would be required to operate to mitigate the consequences of that break. It was determined that the spalling would not adversely impact any safety related components in such a way that would preclude a safe shutdown of the plant subsequent to that break.
In summary, the uprating caused some structural leads to increase (due to a small increase in post accident containment peak pressure and feedwater pipe rupture discussed above), but those increased loads were found not to change the basis of the original plant design, or the struc-tural integrity of any safety-related structures.
H. BOP SYSTEMS AND NSSS INTERFACES
- 1. Main Steam System The Main Steam System piping is designed for 560*F and 1100 psia.
These conditions bound the uprated Main Steam conditions of 547 F/1020 psia at no-load and 536*F/930 psia at 100% power.
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l The Main Steam Safety valves have a total relieving capacity of 12,826,260 lb/hr. Based on Heat Balance 13075.01-HM-7A the total Main Steam flow rate will be 12,251,367 lb/hr. The Main Steam Safety Valves must be capable of passing the entire flow without taking credit for; operation of the Power Operated Relief Valves or the Main Steam Dump System. Using this criterion, it has been shown that the Main Steam Safety Valves are adequately designed for use at the uprated conditions.
The Main Steam Trip and Non-Return Valves were evaluated for rapid i closure impact loads applied subsequent to Main Steam System Pipe rupture at uprated conditions. The results of computer runs that modeled the transient's effect on the valves showed that they would I close as required during a Main Steam Pipe break without jeopardizing
! the integrity of' the pressure boundary.
! 2. Auxiliary Feedwater System l
The Auxiliary Feedwater pumps are designed to deliver rated flow to the Steam Generator at a static discharge head equivalent to the set pressure of the lowes. set Main Steam Safety Valve, 1085 psig. Because this set point will not change and the Main Steam Safety Valves have been shown to be acceptable at the uprated conditions, it is concluded that the resistance parameters associated with the Auxiliary Feedwater System are unchanged.
The Westinghouse NSSS safety evaluation states that the existing Auxiliary Feedwater flow requirements (based on 2910 MW core power plus 2*.) for North Anna are unchanged.
I Our conclusion is that the existing Auxiliary Feedwater System is adequate
- at the uprated condition as the flow requirements and system resistance l parameters are unchanged as a result of the uprating.
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- 3. Extraction Steam System, Feedwater Heaters and Flash Evaporator A heat balance was developed for the proposed 7.5'F increase in Tavg l
at 2787 MW g /930 psia steam pressure. From this heat balance (13075.01-HM-7A) the values were taken for the temperatures and pressures of the Extraction Steam lines.
We have determined that the uprated operating pressures fall within the design conditions for all extraction lines. It was observed that l the uprated operating temperatures of the third point extraction lines on both units were above the design temperature. An analysis was performed to determine the thermal stresses produced in the third point extraction lines at the uprated condition. The maximum thermal stress in the third point extraction lines were found to be below the code allowable stress, confirming that all Extraction Steam lines
! (including the third point elitractions) are adequate to operate at the t
l uprated conditions.
We have also verified that the uprated extraction pressures are within l
the s'lell side design pressures for all of the Feedwater Heaters and the Flash Evaporator. It was conservatively assumed that the pressure at the turbine extraction nozzle was equivalent to the operating pressure at the Feedwater Heater shell without considering pressure drop in the extraction lines.
- 4. Condensate and Feedwater Systems l
l To determine if the Condensate and Feedwater System piping would be adequate at the uprated conditions, a comparison was made between the current North Anna heat balance (13075.01-HM-1) and the uprated heat balance (13075.01-HM-7A). It was shown that the entire Condensate /
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Feedwater piping train uprated temperatures were within one degree of the current condition at all locations. The existing feedwater temper-ature entering the Steam Generator is approximately 1*F higher than j at the uprated condition. This same relationship exists throughout the Condensate and Feedwater systems at each Feedwater Heater inlet and outlet location. Discharge pressure at the Condensate pump dis-r charge increased by approximately 1 psia and discharge pressure at the i
Feedwater pump discharge decreased by approximately 2 psia.
Therefore the predicted change in Condensate and Feedwater System temperature and pressure parameters due to the uprating is insignificant.
l These small changes are within the capability of the current system.
l The total Condensate and Feedwater System resistance was evaluated for l
the new flow rates and Steam Generator pressure perta:.ning to the i 7.5*F increase in Tavg. The overall system resistance increases i
slightly due to the higher Steam Generator pressure and greater fric-
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tion losses due to increased flow rate. However, it has been deter-mined frem analyzing the existing Condensate Feedwater System calcu-1 lation that the existing pumps have sufficient head to overcome the slightly increased total system resistance with two Condensate and two Steam Generator Feed Pumps in operation at the uprated condition. T" he NPSHA at the suction of the Condensate and Feedwater pumps was eval-uated at the uprated conditions. It was determined that sufficient NPSHA exists to allow acceptable operation at the uprated flows.
- 5. Feedwater Regulating Valves A review of the operation of the Main Feedwater Regulating and Bypass Valves has shown that increasing Tavg by 7.5"F will result in a significant decrease in flexibility of the Feedwater Regulating Valve. To ensure the Feedwater Regulating Valves can pass the required flow, the Bypass Valves may need to be maintained partially open or a change in Feedwater Regulating Valve trim may be required.
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- 6. Low Pressure and High Pressure Heater Drain System As with the Condensate and Feedwater systems, the uprated temperature and.
pressure conditions associated with the Low and High Pressure Heater Drain Systems are within one degree of temperature and one psia of pressure from i 1
current conditions. The Low and High Pressure Hester Drain Pumps have been l shown to be adequate at the uprated flow conditions. Uprated NPSHA has been evaluated at the pump suctions and has been determined to be acceptable for pump operation.
- 7. Steam Generator Blowdown System l
A review of the Steam Generator Blowdown System has indicated that uprating l
l Reactor Coolant System Tavg by 7.5*F will not affect the present safety aspects or operability of the system.
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l The design of the excess flow high energy line break isolation valves was for an inlet p'ressure of 1100 psig which is higher than the lowest Main i Steam Safety Valve setpoint and is therefore acceptable with regard to the 1
uprate.
All remaining portions of the Steam Generator Blowdown System including l
flow control valves, safety valves, tanks and pressure control valves were 1
i reviewed for any expected temperature and pressure changes and are un-affected by the uprate.
- 8. Condensate Polishing System l The design conditions for the Condensate Polishing System were evaluated to determine the ability of the polishers to operate at the uprated con-ditions. The proposed temperature and pressure of the condensate at the polisher inlet (470 psig and 103*D) remain below the polishing system design conditions of 700 psig and 180*F. l l
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From heat balance 13075.01-HM-7A, the condensate flow rate is given as 8,001,103 lb/hr or 2.68 gpm per sq. ft. of Condensate Polishing System filtering surface area. This flow rate is less than the vendors guaranteed filtering capacity of 4.0 gpm per sq. ft. of surface area at design pressure drop. Therefore it has been determined that the condensate polishing system is adequate to operate'at the 7.5'F uprated conditions, although at in-creased flow rates, filter differential pressure will increase at a faster rate and the backwash frequency will be slightly higher.
- 9. Auxiliary Steam PCV The Auxiliary Steam Pressure Control Valve PCV-AS-105 has been reviewed for the uprated pressure conditions. The valve is designed for a maximum inlet pressure of 1200 psig. This design pressure is higher than the maximum possible upstream pressure of 1020 psia at no-load conditions. It is there-fore concluded that the Auxiliary Steam PCV is adequate for the 7.5* uprated conditions and will reduce uprated Main Steam pressure as originally designed to 150 psig.
- 10. Component Cooling System The. increased RCS cold leg temperature increases the heat loadings on the Component Cooling Water System due to the Chemical and Volume Control System heat exchangers which are designed to remove heat from the letdown flow stream. The letdownfline is tied into one RCS cold leg.
The affected heat exchangers are the Non-Regenerative, Excess Letdown and Seal Water Return heat exchangers. The cumulative heat loading to the component Cooling Water System at the uprated normal operating condition remains below the design value used for the original plant design. Con-sequently, the Service Water System will not be impacted as a result of the uprating.
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- 11. Spent Fuel Pool Cooling System There is no impact on the Spent Fuel Pit Heat Loads as a result of the uprating since core power and associated decay heat levels remain unchanged.
- 12. Containment Air Recirculation System The heat input into the containment has been estimated Eo increase approximately 2% because of the increased average Reactor Coolant System temperature. This increased heat load is well below the installed design capacity of cooling units in the containment air l
recirculation system. Since Technical Specifications limit the ambient temperature inside the containment to 105'F, no change in l
the operating procedures are necessary and containment temperatures will not exceed that limit.
I. HEAT BALANCE CALCULATIONS A heat balance diagram was developed for analytical and operational use at the uprated conditions of 2787 MW (2775 MWgcore power plus 12 MWg pump heat) and 930 psia steam pressure. This document, 13075.01-BM-7A, was transmitted to Vepco in September 1982 and was used to evaluate secondary system adequacy at the uprated conditions.
J. CONTROLS SYSTEMS AND INSTRUMENTATION The 7.5'F uprate Heat Balance Diagram (13075.01-EM-7A) has been reviewed l
to determine the effect of this uprate on BOP instrumentation and control valves in the Feedwater, Condensate, Main Steam and Heater Drains Systems. The changes in pressures, temperatures and flows were reviewed against the design parameters of the existing equipment. It was determined that the presently installed equipment is adequate for the 7.5'F uprate. No setpoint changes are required as a result of this uprating in addition to those specified by Westinghouse.
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K. ELECTRICAL SYSTEMS An uprated generator output of 954,698 KW (from Heat Balance 13075.01-HM-7A) at .9 PF yields 27,847 Amps to the Iso-Phase Bus Duct. The existing bus duct is rated at 30,500 Amps continuous and is adequate for the uprated conditions.
A review has been performed to evaluate the increased loading on the Feed-water, Condensate and Heater Drain Pump motors resulting from increases in fluid flow rate. As a result of this review, it has been determined that the motors and their associated power feeds are adequate for the uprated conditions. ,
Because the increased loadings on the above mentioned pumps does not exceed the rated horsepower for the respective motors, there is no impact on the Station Service Transformers, Normal Buses or connecting cables due to the uprate.
The Emergency Buses are not affected because the uprating causes none of the emergency loads to increase.
The GDC-17 Confirmatory Analysis Studies have been evaluated with respect to the uprating and it has been determined that no chanSes in the results of that study have been evidenced.
l L. REVIEW OF THE TECHNICAL SPECIFICATIONS l
The Technical Specifications have been reviewed to determine if any sections could be affected by the proposed 7.5 Tavg increased from 580.3'F to 587.8' F.
With the exception of the Technical Specification revisions in Enclosure 3, no additional sections are affected.
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ENCLOSURE 3 North Anna Technical Specifications Changes for Tavg of 587.8 F Note: Changes denoted by a double bar, ll,areidenticalto those in the most recent LOCA analysis NRC submittal, Vepco letter Serial No. 490, dated August 16, 1982.
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f S $ s s T n u n E - Dn Dn u r g U e - E o E o D o t i L l R tR Tc T c [ c s s A b tE nE A A l g g n e V a nW iW oO Rc a Re n At b
0 3 3 i i i d c iO pP fl fi i
1 s s
- E i l
oP p t ot ol f e e p p f o
f p ooo t
p tL eL o x l l l
l A p eA SA Ta %a o o 0 5
% %l A SH H % 3 H H 6 9 W
0 R hR S. h 5. h 0 3 e e 8
1 3
2 3
9 9
8r
> .e t t wE gE St St 1 e e o oi l i l i i i S S g i i p 5w $
i l
A H Ll i
l T 5w S
_ T N
_ I O
- P I
_ E S h h D t i
t i
P f I D l w w R R E A E 1 T R Rs Rs
_ A Ed Ed W H R f Wn Wn O n 0 o Oo Oo P d a f Pc Pc p 1
T o % e e L n s A 9 Ls Ls A o w 1
- T % 0 A A M c t o 2 N 5 1 H2 H2 R e n l E 2 R R fl s e f 2 M 1 E1 E1 i r i
m U $ l H T n
E R -
i Tt T t e p
i r g L T e - n n D t i l S I l R tR Da Da E s s
_ l N H b tf nE ft ft T s g n e A iW l s l s A t g
_ a n d T
nM I l i oO An An R 1 2 i i
_ 0 c i Ro u s s
- M P i oP pP Ro e e p p f f p E T p t c c f o o oo l
o T E p t L eL f f c t t a o S S p eA SA oem oe o 0 51
% %l SM H % b H H 7 !
Y A hi %i %i tn S 0 1 3 2 0 S P R t 2 1 e e 1
1 2 9 9 r t wf gf 5t 5t e I
R o l i li e e >_p P
I T N lo. T i
i l T 5a 5a 1 i S S 1 5 1 p.
R o T h o R
u l O n x i l
l r .
I o w l
g l
e C x x x r u -
- p
- A u u l u t u
l f l n.
i i
l l
E l l f e - - e m t
i f f H n - - v eg n i s n o e e r
e p o o o . r r L r re re e t T u u r 0 i
r t tt t t g u A s s s e 0
T u ua ua eR n
a H e
e s
e e t 0, e eR r r r a 5 r H H N R u P W t
o ,
ei
.ev ei
,ev e e g
t a
I A
P r r r w
o 9
s T c e g gt gt t
a n r e e e l i I a a e r z z z f N e n ni na ag i
p e i w as d R i i l
l R a Ro R e e m w r r r f o R
P H m e e o u u u o l
_ L A
l a r r eh r
eh r
ex c
r t
r p
r s
s s.
s s
s s f
- N u e wg wg t u u e e e e e s n O n w oi nl o v v r r r o u I a o oi Pl i
Pl i f S O O P P P l t
I M P I
s C . . . . . . . . . . e tl . .
3 4 S 6 7 1 1 9 0 1 2 D l
i I 2 1 1 1 ^
f M :il5 58~ -
ei $
1; i J
g TABLE 2.2-1(Continuedl 9
- REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPolllTS E
g H0TATION HOTE 1: Overiemperature AT 1 AT,[K j-K2 *1 (T-T')iK(P-P')-f(AI)!
3 j
-' ,l2S where: AT,
= Indicated AT at RATED TilERNAL POWER T = Average temperature. *F
=
T' Indicated T,yg at RATED DEM POWR.1587.8S' E P = Pressprlzer pressure, psig o
P' = 2235 psig (Indicated RCS pominal operating pressure) 8 o>
IIYg5 '
j, g= The function senerated by the lead-lag controller for T,,g dynamic compensation l -
t j & v2 = Tine constants utlitzed in the lead-lag controller for T,,, ty = 25 secs.
f 1 t
2 " 4 5"CS*
5 = 1.aplace transform operator (sec-I)
5
$ TABLE 2.2-1 (Continued) x P,
f REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS
$ NOTATION (Continued) i
& Operation with 3 Loops operation with 2 Loops Operation with 2 Loops
~
" (no loops' isolated)* (1 loop isolated)*
K = 1.078 Kg =
( ) K g
=
( )
g K = 0.0143 K "
( ) K 2
~ ( )
2 2
~ ~
K = 0.000674 K 3 3 3
\
and fg (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuc1 car ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:
(i) for q - qh between - 36 percent and + 11 percent, f (AI) '= 0 l 7 (where q and q are Percent RATED TilERNAL POWER in the top and bottom b
- halves oI the core respectively, and q + q isb total THERMAL POWER in percent of RATED TilERMAL POWER).
l (ii) for each percent 'that the magnitude of (q - q b) exceeds - 36 percent, l
the AT trip setpoint shall be automatica1ky reduced by 1.2 percent of I its value at RATED TilERMAL POWER.
l l (iii) for each percent that the magnitude of (q - qb) exceeds + 11 percent, the AT trip setpoint shall be automaticalIy reduced by 1.6 percent of its value at RATED TilERMAL POWER.
l
- Values dependent on NRC approval of ECCS evaluation for these operating conditions.
. _ . . - - - - - - - - - . .-. . .. . . .- _. --.- . - - .-- -- - . ~ .
I 2
5
' TABLE 2.2-1 (Continued)
$ REACTOR TRIP SYSTDI INSTRUMENTATION TRIP SETPOINTS s'
i NOTATION (Continued)
! g NOTE 2: Overpower AT 1AT,[K4-K5 I
/TS) '
~
6 (T-f,)-f2IAI)l l
! (I + T3j S
i AT g Indicated A T at RATED TIIERMAL POWER Where: =
a T = Average temperature. *F T" = Indicated T at RATED Ti!ERMAL POWER < 587.8*F r avg -
K = 1.091 l 4 K = 0.02/* F for increasing ' average temperature 5
!. o = 0 for decreasing average temperatures K
- 5 J K 6
~ * # "I 6
" '# 1" T3S = The function generated by the rate lag controller for T "#E dynamic compensation 1 +T 3
T = Time constant utilized in the rate lag controller for T avg T = 10 secs, 3
i S = Laplace transform operator (sec_1) i
< f (A I) = 0 for all AI 2
Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than 2 percent span.
f d
J
7CLE DIS *?l30T!CN I,DC-S EZAT 7 LUX HOT CHANNIL FACTOR-F (Z) i I
LD t NC CC'*DITION 70R C?!RA*ICN 3.2.2 7 q(2) shall be limited by the folleving rela:1:: ships:
Fq ( ) i df [,K( )}for ? > 0.5 P
Fq(Z) 1 (4.403 (,I(2)]for ? 1 C.3 i.
(
l
' where ? = "EIRMAL PCLE RATID I".nF&u. FCkB f
i and K(Z) is :he functics obtaised f := Figure 3.:-2 fer a given core heigh: locacic=.
AF?LICA3ILIM MCDE 1.
[
AC"!CN With 7 (Z) exceeding its 11=1 :
Q
- a. Comply with either of the felicving ACT!CNS:
l
- 1. Reduce TEI??.AL PCku a: least 1 for each 1: ?q( ) exceeds the limit within 15 minu:ss and similarly reduce the 7:ver 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />;Range7Cku Neutron Flux-Eigh Trip Setpois:s with1= the nex:
CPERA~ ION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequen:
PCkB CPERATICN may proceed provided the Cverpever iT Trip Setpoints have been reduced a: leas: 12 for each 1: Tq (:)
The Overpever AT Trip Se: point redue:1:n~
exceeds the 1121:.
shall be performed vi:h the reae:or in at least ECT STAiT3Y.
l
- 2. Reduca THI32'.AL PCWER as necessary t: =eet the Id-*:s of Specification,3.2.6 us1=g the AP T.S vi:h the latest i= core sap and upda:ed R.
i
- b. Identify and co hec: the cause of the cut of limit c:ndition prior to increasing TEI72d.AL PCkB above the reduced 11=1: required by.2, above; TEI32'.AL PCku may then be increased provided qT (2) is
't. l demonstrated th:: ugh incere mapping to be within 1:s l' i
l 3Cr3 ANNA - UNIT 1 3/4 2-5
/
-- , . . _ . ~ - . - .
. ;_ .g ;, . ,.; ., - - . - .
g,---.
_ , . : 3 _ ,.__
--.,y
.n=: I n=: . . .t: .=.:r- - - - i . r..n _ ~L.-. .=_ :-. _ r.__ - .. _ r = . . c.. .-- - : .
jj=.::.------ - _ . - -
.:.:;. ..i.j:. a:y. i : nn j
- . :: -.: r p=~=.==.
-. = T".::
-- . ==:-----'- -
_ - - ,:=.::=- ;- _=_=::. _ z-_-..t
._..:..- o - _:- --, ._.==
. . w.:w: . . -. -__ _., _r --. _.-
.-,2,2 22 _
1.O
_g
.. a n e ts _ u , &
.. a .
_,_;,,__,.,,_.;.....*.T...._..g...._p.._.....__
....t....
-... .1.
= . .u;..:... . ..._-...... .
. . . . ..i; . .r.
u ;.;,1...
- ., ;- 2
- g - ... 1:; .. : . ., . u; ..
- p.p-M'ia:M? .';r.:i; ^
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g '*'
- ! .k i I : i .-
SI.5ET: _;i l l: - lT?:::ili: -l*N ":!l~i! Ti.:n di:}:!hi :I cr -
.. ntj n ~.{nk....l
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. ut;. . . .. :
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- E * * ** i: :b k:.; "*Mi: . ! * - *;h.* * . : : ~k. :i-
- *-- * ** i 2:sej==m -= q- ;h g= gi:-it= gja q ==sp.s-t"i=5lgii=ig==;\ =p. r . g--
3
- b +El?i
- + : l-i Fini:i:-st ;s .iil.*-...:ii - - -i:
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..line\+. -
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a- -~ -
z.. . t : = j.~.:: .. t .. t.. ..
==q_; t.: ...ir. .1 2.
. -' ..v -t g t: .::::= r:: . * ::.:: ' : . . l .=. _-l::.n . :l. .n: rtu mu.ra.~ ::t.:
- O: : .:':n.: ..
- l:; T.*'..*.;..~-.
7%
n
-:.t=: - _ .-....... .== 4.:..-. = :.{...- : .f =* C:::M \ :!. .. ..!:== = = --= .= ~ ~ == - m' J.";"*C..tr r u - ...
.W4 34)._
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g
.- . . .;u-:.: '{
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=.: r :.,:.o h -
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- - -- e.---u : r : -
g g
-- =:r -- -l _ l:.:-
_ ;; ;- - r .. . ..g s.- . l. .
v l --.. . :. . ,; ;; ,.ul-
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t 02 c-- .: a.:::.
i u::
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1:
- p: . :.l.
.l-t
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. ..... . u.,; ..
. . . . . #. ;; -. . .g_..
- ll 92nh;ri' & *3**
- 2 SElisi p; u.
,5:i'Jii' a=.+.
JE-i.9 ?!!! 1..
!"i!!S i'-tr "rl'
. ;. + . r 4 6 8 10 12 0 2 CORE HEIGHT (FI)
Figure 3.2-2 K(Z) - Nor=ali::ed Fq(Z) as a Function of Core Height NORTH ANNA - 1: NIT 1 3/4 2-8 l
I a
TABLE 3.2-1 h
- DNB PARAMETERS h
I LIMITS g
2 Loops In Operation ** 2 Loops In Operation ** -
~ 3 Loops In & Loop Stop & Isolated Locp Operation Valves Open Stop Valves Closed PARAMETER Reactor Coolant System T 1592*F ,,
yg Pressurizer Pressure > 2205 psig*
Reactor Coolant System > 285,000 gpm Total Flow Rate M
n Y
d
- Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% RATED TIIERMAL POWER per minute or a TilERMA1, POWER step increase in excess of 10% RATED TIIERMAL POWER.
- Values dependent on NRC approval of ECCS evaluation for these conditions.
PCh'IR DIS II30"l* ION LIMI S AIIAL PCh'ER DISTRI3CICN LWC CONDI !ON FOR OPERMION 3.2.6 The axial power distribution shall be limited by the following relationship:
=
r 2. 201 rx(z)1 3(Z) 3 (I))(?g)(1.03)(1 + c'))(1.07)
Whers:
- a. T)(Z) is the cor 2H rad axial pcver distribution f :m thi=bla j at core elevation Z.
- b. ? is the fraction of R&D M.AL ?Cb"ER.
- c. K(Z) is the functicu obta1=ed f:em 71gure 3.2-2 for a given core height location.
- d. I , for t '-kla h j , is determined f :n at least n=6 i= core 3
fluz zaps covering the full configuratien of permissible rod patterns above ?, of RfD N.AL PCVII in accordance with:
a = =111 r R J 13 Where: v **
T0*1 R
ij = [Fq(2)]
and(Tg3 (2)]g is the mard=-= value of the to:malized sxial distribution at elevatics : f := thi=ble j i= map i which had a measured peaking facter without uneartainties v
or densificatien allevance of 7 *.Q ass .
I i NORI3 ANNA - UNI 1 3/4 2-16
1-i~
3/4.2 POWER DISTRIBUTION LIMITS BASES
\
The specifications of this section provide assurance of fuel integrity i during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature & cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides i assurance that the initial conditions assumed for the LOCA analyses are met-and the ECCS acceptance criteria limit of 2200 F is not exceeded.
The definitions of certain hot channel and peaking factors as used in these specifications are as follows:
l j Fq (Z) Heat Flux Hot Channel Factor, is defined as the maximum local
- heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for man-ufacturing tolerances on fuel pellets and rods.
l F H
Nuclear Enthalpy Rise Hot Channel Factor, is defined as the
! ratio of the integral of linear power along the rod with the l
highest integrated powe, to the average rod power, F Radial Peaking Factor, is defined'as the ratio of peak power
- Y(Z) density to average power density in the horizontal plane at i core elevation Z.
3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) i The limits on AXIAL FLUX DIFFERENCE assure that the Fq(Z) upper bound envelope, as given in Specification 3.2.2, is not exceeded during either ll normal operation or in the event of xenon redistribution following power changes.
Target flux difference is determined at equilibrium xenon conditions.
The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. The value of the l- target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by mutiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux dif ference value is necessary to reflect core burnup j considerations.
l NORTH ANNA - UNIT 1 B 3/4 2-1 l
[
\
l
O N
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. - l' ,
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. - l_ : --l . - - .. j . .9. _ t. 4. _..
g j ..l. ..l l . :- "
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.i =l c
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- r-. . . . u.
[.i?-s-/-fN 555!5M-[Iii 'II' T@/ 'c" I 5IW . !? O C C C C O C e e e e w ~ m m o e e e J. * (N 1 + # ) 12/t 3E01W3dW3130W3M NORTH ANNA - UNIT 2 2-2
IAlliE 2.2-I
!s
- REACIOR IRIP SYS11H INSIRilHENIATION TRIP SEIP0lNIS
]
si filNCIl0NAl. UNIT 1 RIP SE1PulH_I_ All0WAlllE VAlllES
. l. Manual Reactor Irlp Hot Applicalile Hot Applicable G low Setpoint - 1 26% of RATED G 2. Power llange, Heutron Elux low Setpoint - 1 25% of RATED
lilERHAL POWER IllERHAl. POWER ers Iligh Setpoint - 1 109% of RAIED liigh Setpoint - 1 1I0% of RAIED lilERHAt. POWER lilERHAL l>0WER
- 3. Power Range, Heutron flux, 5 5% of RAIED IllERHAL POWER with 5 5.5% of RAlED IllERHAL POWER 111 01: Positive Rate a time constant 1 2 seconds with a time constant 1 2 seconds
- 4. Power Range, Heutron Elux, 5 5% of RAIED lilERHAl. POWER with $ 5.5% of RAIED lilERHAL POWER 111 0 1: Negative Rate a time constant 1 2 seconds with a time constant 12 secoruls
- 5. Intermediate Range, Heutron < 25% of RA1ED lilERHAL POWER < 30% of RATED lilERHAL POWER
', Elux o.
- 6. Source Range, Neutron flux 1 10 counts per second 5 1.3 x 10 counts per second
- 7. Overtemperature AI See Hote 1 See Hote 3
- 11. Overpower Al See Hole 2 See Hate 3
- 9. Pressurizer Pressure--low 1 1870 psig 1 1060 psig
- 10. Pressurlier Pressure--liigli 123115 psig i 2395 psig II. Pressurizer Water level--liigli- 5 92% of instrimient span 5 93% of instrument span
- 12. l os:. of flow 1 90% of design flow 1 (19% of design flow per loop" per loop ^
^ Design flow is 95,000gpm per loop. l e
w.-
g- I Alli t 2.2-1 (Continised}
Rf ACIOR TRIP SYSIIH INSTHilHINTAll0N IRIP SfiPolNIS II NOIAIION ji C
3 N0ll 1: Overtemperattare Al 1 AI,[K g-K2 18:i5 (i'i )'K 3 (P'P )'t l (Ot))
-8 los 2, S
n where: Al = Indicaleal Al at itAlfD flifRMAL POWEN T = Av rage temperatisre. *f I" = Indicated I at 1(AllD lillRHAl. POWER $ 587.8 F l P = Pressairlier pressiare, psig P' = 2235 psig (tsuficatest HC5 nomissal operating pressiere)
[ -
ici,5 = lhe inanction llesieratual by the least-lag controller for 1 dynamic compensation y,
& = lime const.ints utilized in the lead-lag controller for T s = 25 secs, y
u, 2 v
2"4**"'
S = laplace tratistoim operator (sec" )
I O
e
8 so
$ TABLE 2.2-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS h
NOTATION (Continued)
[
- .2 O Operation with 3 Loops Operation with 2 Loops Operation with 2 Loops e2 (no loops isolated)* (1 loop isolated)*
Kg = -1.078 Kg =
( ) Kg =
( )
K = 0.0143 K "
( ) K 2
~ ( )
2 2 K = 0.000674 K =
( ) K ~
( ) -
3 3 3
and fg (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant .startup tests such that:
(i) for q - g between - 36 percent and + 11 percent, f (AI) =0 l b
{ (where q and qh are percent RATED TilERMAL POWER in the top and bottom halves of the core respectively, and q + qb '8 " I" percent of RATED TilERMAL POWER).
(ii) for each percent that the magnitude of (q - qb) exceeds - 36 percent, the AT trip setpoint shall be automatica1Iy reduced by 1.2 percent of its value at RATED Ti!ERMAL p0WER.
(iii) for each percent that the magnitude of (q - q ) exceeds + 11 percent, b
the AT trip setpoint shall be automatica1Iy reduced by 1.6 percent of its value at RATED TilERMAL POWER.
- Values dependent on NRC approval of ECCS evaluation for these operating conditions.
g TABLE 2.2-1 (Continued) d
- REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS t,*
NOTATION (Continued)
E Overpower AT <aT [K4-K5 l r $s )
H 6 (T-T")-f2(AI
~
NOTE 2: I N (1 + T 38i Where: A,T = Indicated AT at RATED TifERMAL POWER T = Average temperature, *F T" = Indicated T at RATED TilERMAL POWER < 587.8 F I avg - i K 1.091 4 K = . r n r asing average temperature 5 g K 5
~ "# "" "E """#"E * "E #
K = 0.00126 for T >T"; K 6
= 0 for T <T" 6
T S 3
= % e function generated by the rate lag controller for Ta ,g dynamic compensation 1+T S 3
T = Time constant utilized in the rate lag controller for T M 1 = secs. 3 S = Laplace transform operator (sec~I) f2 (AI) = 0 for all AI Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than 2 percent span.
1 i POWER DISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR-F (Z) LIMITING CONDITION FOR OPERATION 4 l 3.2.2 F (Z) shall be limited by the following relationships: q Fq (Z) 1 '2.200 , K(Z) for P > 0.5
,P .I Fq (Z) 1 [4.40] K(Z) for P 10.5 where P = THERMAL POWER RATED THERMAL POWER i
and K(2) is the function obteiMed from Figure 3.2-2 for a given core height location. i APPLICABILITY: MODE 1. i ACTION: With F (Z) exceeding its limit: q
- a. Comply with either of the; following ACTIONS: ,
- 1. Reduce THERMAL POWER at least 1% for each 1% qF (Z) exceeds the limit within.15. minuter and similarly reduce the Power Range s
Neutron Flux-High Trip Setpoints within the next 4 hours; POWER
- OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower aT Trip Setpoints have been reduced at least 1% for each 1% Fq (Z) exceeds the limit. The Overpower AT Trip Setpoint reduction shall be performed with the reactor in at least HOT STANDBY.
- 2. Reduce THERMAL POWER as necessary to =ect the limits of Specification;3.2.6 using the APDMS with the latest incore map l and updated R.
i i
- b. Identify and correct the cause of the out of limit condition prior I
to increasing THERMAL POWER above the reduced limit required by a, above; THERMAL POWER may then be increased providedqF (Z) is demonstrated through incore mapping to be within its limit. .t NORTH ANNA - NL'IT 2 3/4 2-5 4 9
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3 . 2 4 6 8 10 12 I O CORE HEIGHT (FI) Figure 3.2-2 K(2) - Nor::talized F q(Z) as a Fune:1cn of Core Height NORTH ANNA - UNIT 2 3/4 2-8
z
@ TABLE 2.2-1 E!
UNB PARAMETERS g N
' LIMITS y 2 Loops In Operation ** 2 Loops In Operation **
y 3 Loops In & Loop Stop & Isolated Loop PARAMETER Operation Valves Open Stop Valves Closed Reactor Coolant System T < 592* F Pressurizer Pressure > 2205 psig* Reactor Coolant System >285,000 gpm Total Flow Rate M t O Y
- Limit not applicable during either a THERMAL POWER ramp increase in excess of 5% RATED TilERMAL POWER per minute or a TIIERMAL POWER step increase in excess of 10% RATED TIIERMAL POWER.
** Values dependent on NRC approval of ECCS evaluation for these conditions.
PCLTR DISTRI3t*TICN LDCTS ATTAL PCkTR DISTRI3CCCN LDdtlNG CCNDIMON FOR 0?! RATION 3.2.6 The axial power distribu:1on shall be 11=ited by the fo11 v1=g relationship: I2.24 IR(z9~ [F3 (Z)]s * (R))(?g)(1.03)(1 + o'))(1.07) klere:
- a. F (2) is the normalized axial power dis::1butice frc= th1=ble 3
j at core eleva:1cn Z.
- b. P is the frac:fon of RAID MAL PChTR.
- c. K(Z) is the fune:1on obtained f :s Figure 3.2-2 for a given core height loca:1cn.
- d. less: n=6 incere R), for thimble j , is deter =ined f =m a:
flux maps covering the full configuration of pe =issible cd pa : erns above PJ. of RAID MAL ?CkTR i= accordancs with: T3 R, , 1= $.1 1
~
kiers: yMeas 01 R ij . {Fq(2)]g and (Tg )(2)]g is the maximum value of the nor=alized axial dist:1buti:n at eleva:1cn : f :s chimble j in map i FORT 3 ANNA - UNIT 2 3/t. 2-17
\
3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature & cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded. The definitions of certain hot channel and peaking factors as used in these specifications are as follows: Fq(Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for man-ufacturing tolerances on fuel pellets and rods, i F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the
~
ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. F Radial Peaking Factor, is defined as the ratio of peak power
*I(Z) density to average power density in the horizontal plane at core elevation Z.
i 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) The limits on AXIAL FLUX DIFFERENCE assure that the Fq (Z) upper bound envelope, as given in Specification 3.2.2, is not exceeded during either ll i normal operation or in the event of xenon redistribution following power
- changes.
Target flux difference is determined at equilibrium xenon conditions. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER l for the associated core burnup conditions. Target flux differences for other j NORTH ANNA - UNIT 2 B 3/4 2-1
I ENCLOSURE 4 Core Surveillance Reports for North Anna 1 Cycle 4 and North Anna 2 Cycle 2 f l l
TABLE 1 MORTH AMMA UMIT 1, CYCLE 4 CORE SURVIILLAMCI LIMI'.S, TS = 2.20 I. The T-XY limits fo: RATED THERMAL POWER uithin spec fic core planes shall be:
- 1. Txy-RTP $ 1.71 for all core planes containing bank "D" control rods, and
- 2. Txy-RTP < 1.65 for all un:cdded co:e planes hetueen 15 % and 25 % of co:e height, or
- 3. Txy-RTP :E 1.70 for all unrodded core planes hetueen 25 % and 55 % of co:e height, or
- 4. Txy-RTP $ 1.65 for all un:odded core planes hetueen 55 % and 85 % of core height.
II. The axial poue: distribution surveillance th:eshold power level shall bei
- 1. Pm = 100% of RATE 3 THERMAL PCMER.
l l J
\
NORTH ANNA UNIT 1 CYCLE 4 MAXIMUM FO-TOTALaP VS. RXIRL CORE HEIGHT DURING NORMAL CORE OPERATION 2.2 x x xx l
, y .x) v xx x %
xx x x
' X x'x X 2.0 x X.<
1 4' X x ! b 1.8 . , ; M - 1 A j X 1 M U 1.6 ; I M F G :
\ .
T 1.4 0 . 3 T l R y L
; X o
P 1. 2
. I' 4
I l < 1.0
- x 0.8 . . . .
i . . 0 2 4 6 8 10 12 CORE HEIGHT (FEET) TECHNICAL SPECIFICATIONS LIMIT X CALCULATED DATR .
o . TABLI 1 NORTH ANNA UNIT 2, CYCLI 2 CORI SURVE*LLANCE LIMITS, T2 = 2.20
- 2. The T-XY limits fo: RATI: THERMAL POWER within specific coze' planes shall be:
- 1. Txy-RTP g 1.71 for all cc:e planes containing bank "3" control :ods, and i
; 2. Txy-RTP s 1.65 for all unrodded co:e planes between 15 % and 25 % of core height, or
- 3. Txy-RTP s 1.70 for all un:odded co:e planes between 25 % and 55 % of co:e height, o:
4 Txy-RTP s 1.65 fo: all un:odded co:e planes between 55 % and 85 % of core height. f l TI. The axial power distribution surveillance th:eshold poue: level shall be
- 1. Pm = 100% of RATED THERMAL POWER.
l l I e l l l
r NORTH ANNA UNIT 2 CYCLE 2 MAXIMUM FG-TOTAlmP VS. AX1AL CCRE HEIGHT f I OURING NORMAL CORE OPERAT10N l l l l * - ' - * * -' ' ' ' ' 2.2 ' - \ . y Y xx x x x x 'x
'. x x '
2.0 ' I*xv'vxxvvvx xx vxx X xX^
- x ;
il x ! l x v 1.8
- M A :
x X ~ ! I xo
- M V 1.6 '
II ( ( M x F ' x ' l 0 l \ T 0 1.4 ' } 1 T : A L . m P 1.2 l i - y 1.0 [
^ l 0.8. . . . . . .
0 2 4 6 8 10 12 l CORE HEIGHT (FEET) l l TECHNICAL SPECIFICATIONS LIMIT t X CALCULATEC GATA .}}