ML14175A842

Amend 87 to License DPR-23,revising Tech Spec to Authorize Cycle 10 Operation at Full Power 2,300 Mwt W/New Steam Generators & Incorporate Changes Resulting from Cycle 10 Core Reload Analysis

ML14175A842
Person / Time
Site:	Robinson
Issue date:	11/07/1984
From:	Varga S Office of Nuclear Reactor Regulation
To:
Shared Package
ML14175A843	List: ML14175A842 ML14175A844 ML14175A845 ML20099J505
References
DPR-23-A-087 NUDOCS 8411290082
Download: ML14175A842 (53)

Text

o UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 CAROLINA POWER AND LIGHT COMPANY DOCKET NO. 50-261 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.8 7 License No. DPR-23

1. The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Carolina Power and Light Company (the licensee) dated July 23, 1984 and August 1, 1984, as supplemented by letters dated August 8, 17, and 20(2), and 23, 1984; September 7(2) and 17, 1984; and October 4, 12, and 22, 1984, 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.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No. DPR-23 is hereby amended to read as follows:

8411290082 841107 PDR ADOCK 05000261 P

PDR

-2 (B) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 87, 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 COMMISSION 1 71Steven A. Varga, Chief Operating Reactors Branch #1 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: November 7, 1984

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 87 FACILITY OPERATING LICENSE NO. DPR-23 DOCKET NO. 50-261 Revise Appendix A as follows:

Remove Pages Insert Pages 2.1-1 thru 2.1-8 2.1-1 thru 2.1-4 2.3-1 thru 2.3-6 2.3-1 thru 2.3-6 3.1-1 thru 3.1-3a 3.1-1 thru 3.1-3b 3.1-11 thru 3.1-12 3.1-11 thru 3.1-12 3.5-7 and 3.5-7a 3.5-7 and 3.5-7a 3.5-10a 3.5-10a 3.6-1 thru 3.6-2 3.6-1 thru 3.6-2 3.6-2a 3.6-3 3.8-6 3.8-6 3.10-2 thru 3.10-7 3.10-2 thru 3.10-7b 3.10-12 3.10-12 3.10-14 thru 3.10-20 3.10-14 thru 3.10-20 3.10-22 3.10-22 thru 3.10-24 3.11-1 thru 3.11-2 3.11-1 thru 3.11-2 4.11-1 thru 4.11-3 4.11-1 thru 4.11-3 5.3-1 thru 5.3-2 5.3-1 thru 5.3-2

9 9(HBR-11) 2.0 SAFETY LIHITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT, REACTOR CORE Applicability Applies to the limiting combinations of thermal power, Reactor Coolant System pressure, coolant temperature, and flow when the reactor is critical.

Objective To maintain the integrity of the fuel cladding.

Specification

a.

The combination of thermal power level, coolant pressure, and coolant temperatuie shall not exceed the limits shown in Figure 2.1-1 when.

full flow from three reactor coolant pumps exists.

b.

When full flow from one reactor coolant pump exists, the thermal power level shall not exceed 20%, the coolant pressure shall remain between 1820 psig and 2400 psig, and the-Reactor Coolant System average temperature shall not exceed 590 oF.

c.

When natural circulation exists, the thermal power level shall not exceed 12%, the coolant pressure shall remain between 2135 psig and 2400 psig, and the Reactor Coolant System average temperature shall not exceed 620aF.

d.

The safety limit is exceeded if the combination of Reactor Vessel inlet temperature and thermal power level is at any time above the appropriate pressure line in Figure 2.1-1 or if the thermal power level, coolant pressure, or Reactor Vessel inlet temperature violaces the limits specified above.

Amendment No. 87 2.1-1

Basis To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all operating conditions.

This is accomplished by maintaining the hot regions of the core within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the coolant saturation temperature. The upper boundary of the nucleate boiling regime is termed departure from nucleate boiling (DNB) and at 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 XNB DNB correlation.

The XNB DNB correlation has been developed to predict the DNB flux and the location of DN8 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 indicative of the margin to DNB.

The minimum value of the DNB ratio, DNBR, during normal operational transients and anticipated transients is limited to 1.17.

A DNB ratio of 1.17 corresponds to a 95%

probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating consitionsed The DNB ratio limit of 1.17 is a conservative design limit which is used as a basis for setting core safety limits.

Based on rod bundle DNB tests, no fuel rod damage is expected at this DNB ratio or greater.

The curves of Figure 2.1-1 which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (three loop operation) represent the loci of points of thermal power, reactor vessel inlet temperature, and coolant system pressure for which the DNB ratio is not less than 1.17.

The area where clad integrity is assured is below these lines.

The temperature limits at Low power are considerably more conservative than would be required if they were based upon a minimum DNd ratio of 1.17 but Amendment No. 87 2.1-2

(HBR-11) are set to preclude bulk boiling at the vessel exit.

An arbitrary upper safety limit of 118% thermal power is shown.

This limit is based on the high flux trip including all uncertainties.

Radial power peaking factors consistent with the limit on F given in Specification 3.10.2.1 have been employed in the generation of the curves in Figure 2.1-1.

An additional heat flux factor of 1.03 has been included to account for fuel manufacturing tolerances and in-reactor densification of the fuel.

The safety limit curves given in Figure 2.1-1 are for constant flow conditions.

These curves would not be applicable in the case of a loss of flow transient.

The evaluation of such an event would be based upon the analysis presented in Section 15.3 of the FSAR.

The Reactor Control and Protection 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 a DNB ratio of less than 1.17(2) based on steady state nominal operating power levels less than or equal to 100%, steady state nominal operating Reactor Coolant System average temperatures less than or equal to 575.4'F, and a steady state nominal operating pressure of 2235 psig.

Allowances are made in initial conditions assumed for transient analyses for steady state errors of

+2% in power, +40F in Reactor Coolant System average temperature, and +/-30 psi in pressure. The combined steady state errors result in the DNB ratio at the start of a transient being 10 percent less than the value at nominal full power operating conditiods.

Referentces (1) XN-NF-711(P) Rev. 0, "XNB Addendum for 26 Inch Spacer."

(2)

FSAR Section 15.

Amendment No. 87

2. 1-3

680 I

660 240j HIGH FLUX TRIPI 60 300 Da AT 118% OF I

sia RATED POWER 640.

200 ala 2000 5*

00 si 6 0 20

40) 608010 Dasl 600P ml 580 560 540 520 500 20 40 60 s10 100 1 20i PERCENT OF 2300 MWT CORE PROTECTION BOUNDARIES FOR 3-LOOP OPERATION Figure 2.1-1 2.1-4 Amendment No. 87

(HBR-11) 2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTIVE INSTRUMENTATION Applicability Applies to trip settings for instruments monitoring reactor power and reactor coolant pressure, temperature, and flow and pressurizer level.

Objective To provide for automatic protection action in the event that the principal process variables approach a safety limit.

Specification 2.3.1 Protective instrumentation settings for reactor trip shall be as follows:

2.3.1.1 Start-up protection

a. High flux, power range (low setpoint)

< 25% of rated power.

2.3.1.2 Core protection

a. High flux, power range (high setpoint)

< 109% of rated power

b. -High pressurizer pressure < 2385 psig.

c. Low pressurizer pressure > 1835 psig.

d.

Overtemperature A T (1 + TS)

AT0 (K1 -

K 2 (1 + T2S)

(T -

T') + K3 (P -

P')

f(AI) 1 2.3-1 Amendment No. 87

(HBR-11) where:

ATo

= Indicated AT at rated thermal power; T

= Average temperature, F; P

= Pressurizer pressure, psig; Kl

< 1.1565; K 2

= 0.01228; K

= 0.00089; 1 + TIS

= The function generated by the lead-lag controller for Tavg

1. +

2 dynamic compensation; T, & T2 = Time constants utilized in the lead-lag controller for Tavgs r1 = 20 seconds, T2 3 seconds; T'

- 575.4*F Reference Tavg at rated thermal power; P'

- 2235 psig (Nominal RCS Operating Pressure);

S

- Laplace transform operator, sec and f(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 start-up tests such that:

(1) For (qt -

qb) within +12% and -17%., where q, and qb are percent power in the top and bottom halves of the core, respectively, and q

+ qb is total core power in percent of rated power (2300 Mwt),

f(AI) = 0.

For every 2.4% below rated power (2300 Mwt) level, the permissible positive flux difference range is extended by +1 percent. For every 2.4% below rated power (2300 Mwt) level, the permissible.negative flux difference range is extended by -1 percent.

(2) For each percent that the magnitude of (qt -

qb) exceeds +12% in a positive direction, the AT trip setpoint shall be automatically reduced by 2.4% of the value of AT at rated power (2300 Mwt).

2.3-2 Amendment No. 87

-(HBR-11)

(3) For each percent that the magnitude of (qt b) exceeds -17% in the negative direction, the AT trip setpoint shall be automatically reduced by 2.4% of the value of AT at rated power (2300 Mwt).

e.

Overpower A T t3S SAT0 { Kq -

K5 (

) T K

K6 (T -

T') -

f(AI) }

oT -

51

+

3 where:

AT

= Indicated AT at rated thermal power, F; T

- Average temperature, OF; T'

= 575.4*F Reference Tavg rated thermal power; K4

< 1.07; K5

= 0.0 for decreasing average temperature, 0.02 sec/*F for increasing average temperature; K6

= 0.00277 for T > T' and 0 for T < T; S

= Laplace transform operator, sec-1 T3S I + T;S = The function generated by the rate-lag controller for Tavg dynamic compensation; T3

= Time constant utilized in the rate-lag controller for Tavg, T3 = 10 seconds; f(AI)

= As defined in d. above

f.

Low reactor coolant loop flow > 90% of normal indicated flow.

g.

Low reactor coolant pump frequency > 57.5 Hz.

h.

Undervoltage > 70% of normal voltage.

2.3.1.3 Other Reactor Trips

a.

High pressurizer water level K 92% of span.

b.

Low-low steam generator water level > 14% of narrow range instrument span.

2.3-3 Amendment No. 87

(HBR-11) 2.3.2 Protective instrumentation settings for reactor trip interlocks shall be as follows:

2.3.2.1 The low pressurizer pressure trip, high pressurizer level trip, and the low reactor coolant flow trip (for two or more loops) may be bypassed below 10% of rated power.

2.3.2.2 The single-loop-loss-of-flow trip may be bypassed below 45% of rated power.

Basis The power range reactor trip low setpoint provides protection in the power range for a power excursion beginning from lower power.

This trip value was used in the safety analysis.(1 )

In the power range of operation, the overpower nuclear flux reactor trip protects the reactor core against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. The prescribed set point, with allowance for errors, is consistent with the trip point assumed in the accident analysis.(2)

The source and intermediate range reactor trips do not appear in the specification, as these settings are not used in the transient and accident analysis (FSAR Section 15).

Both trips provide protection during reactor startup. The former is set at about 10+5 counts/sec and the latter at a current proportional to approximately 25% of full power.

The high and low pressure reactor trips limit the pressure range in which reactor operation is permitted.

The high pressurizer pressure reactor trip is also a backup to the pressurizer code safety valves for overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig).

The low pressurizer pressure reactor trip also trips the reactor in the unlikely event of a loss-of-coolant accident.(3 )

2.3-4 Amendment No. 37

(HBR-11)

The overtemperature AT reactor trip provides core protection against DNB for all combinations of pressure, power, coolant temperature, and axial power distribution provided only that (1) the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds)

, and (2) pressure is within the range between the high and low pressure reactor trips. With normal axial power distribution, the reactor trip limit, with allowance for errors,(2) is always below the core safety limit as shown in Figure 2.1-1.

If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to Specification 2.3.1.2.d.

The overpower AT reactor trip prevents power density anywhere in the core from exceeding 118% of design power density as discussed in Section 7.2.2 of the FSAR and includes corrections for axial power distribution, change in density, and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. The specified setpoints meet this requirement and include allowance for instrument (2) errors.

The setpoints in the Technical.Specifications ensure the combination of power, temperature, and pressure will not exceed the core safety limits as shown in Figure 2.1-1.

The low flow reactor trip protects the core against DNB in the event of a sudden loss of power to one or more reactor coolant pumps.

The setpoint specified is consistent with the value used in the accident analysis.(5 ) The undervoltage and underfrequency reactor trips protect against a decrease in flow caused by low electrical voltage or frequency. The specified setpoints assure a reactor trip signal before the low flow trip point is reached.

The high pressurizer water level reactor trip protects the pressurizer safety valves against water relief. Approximately 1150 ft3 of water corresponds to (2) 92% of span.

The specified setpoint allows margin for instrument error and transient level overshoot beyond this trip setting so that the trip function prevents the water level from reaching the safety valves.

2.3-5 Amendment No. 37

(*BR-11 The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents. The specified set point assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the auxiliary feedwater system.(6 )

The specified reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal plant operations.

The prescribed set point above which these trips are unblocked assures their availability in the power range where needed.

Above 10% power, an automatic reactor trip will occur if two reactor coolant pumps are lost during operation. Above 45% power, an automatic reactor trip will-occur if any pump is lost. This latter trip will prevent the minimum value of the DNB ratio, DNBR, from going below 1.17 during normal operational transients and anticipated transients when only two loops are in operation and the overtemperature AT trip setpoint is adjusted to the value specified for three loop operation.

The turbine and steam-feedwater flow mismatch trips do not appear in the specification, as these settings are not used in the transient and accident analysis.

(FSAR Section.15)

References (1) FSAR Section 15.4 (2) FSAR Section 15.0 (3) FSAR Section 15.6 (4) FSAR Section 15.4.2 (5) FSAR Section 15.3 (6) FSAR Section 15.2 2.3-6 Amendment No. 87

(HBR-11) 3.0 LIMITING CONDITIONS FOR OPERATION Except as otherwise provided for in each specification, if a Limiting Condition for Operation cannot be satisfied because of circumstances in excess of those addressed in the specification, the unit shall be placed in hot shutdown within eight hours and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless corrective measures are taken that permit operation under the permissible Limiting Condition for Operation statements for the specified time interval as measured from initial discovery or until the reactor is placed in a condition in which the specification is not applicable.

3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the Reactor Coolant System.

Objective To specify those Reactor Coolant System conditions which must be met to assure safe reactor operation..

Specification 3.1.1 Operational Components 3.1.1.1 Coolant Pumps

a. With reactor power less than 2% of rated thermal power and less than two reactor coolant pumps in operation, one of the following actions shall be taken:

1. maintain.a shutdown margin of at least 4% ak/k, or

2. open the lift disconnect switches for all control rods not fully withdrawn, or

3. open reactor trip breakers.

3.1-1 Amendment No. 87

(HBR-11)

b. Power operation with less than three loops in service is prohibited.

c. At least one reactor coolant pump or residual heat removal pump shall be in operation when Tavg > 200*F and reactor power is less than 2% of rated thermal power. In the event this condition cannot be satisfied, the following actions shall be taken:

1. Proceed to establish a boron concentration in the reactor coolant equal to or greater than that concentration needed to maintain a shutdown margin of 1% Ak/k at 200*F, and

2. Restore at least one reactor coolant pump or residual heat removal pump to operation within one hour, or prepare and submit a Special Report to the NRC within 30 days.

d. A reactor coolant pump may be started (or jogged) only if there is a steam bubble in the pressurizer or the steam generator temperature is no higher than 50dF higher than the temperature of the reactor coolant system.

Basis Specification 3.1.1.1.a contains requirements designed to limit the consequences of the uncontrolled bank withdrawal at low or subcritical power conditions as analyzed in the safety analysis. -The requirement of two reactor coolant pumps in operation below 2% power is consistent with the assumptions utilized in the bounding transient that was analyzed. The specification makes allowance for less than two pumps in operation by specifying either of three actions that must be taken. Either maintaining the specified shutdown margin, opening the lift disconnect switches on the control rods or opening the reactor trip breakers will prevent the occurrence of the postulated uncontrolled bank withdrawal transient, therefore allowing the two pump requirement to be lifted.

Maintaining a shutdown margin of 4% Ak/k is sufficient to prevent a return to criticality if the worth of the two most reactive control rod banks are simultaneously withdrawn as is the assumption of the postulated transient.

3.1-2 Amendment No. 87

(HBR-11)

Specification 3.1.1.1.b requires that all three reactor coolant pumps be operating during power operation to provide core cooling in the event that a loss of flow occurs.

The flow provided will keep DNB well above 1.17.

Therefore, cladding damage and release of fission products to the reactor coolant will not occur.

Specification 3.1.1.1.c is designed to allow for adequate mixing of the reactor coolant to maintain a uniform boron concentration during dilution, and to provide a means of boron injection. Should no residual heat removal pump or reactor coolant pump be available, boration via natural circulation shall be initiated.

A boron concentration corresponding to 1% Ak/k at 200*F (which assumes most reactive rod stuck out) would prevent a return to criticality during the cooldown phase of the postulated steam line break event.

The pressurizer is of no concern because of the low pressurizer volume and because the pressurizer boron concentration will be higher than that of the rest of the reactor coolant.

The.purpose of Specification 3.1.1.1.d is to limit pressure surges exhibited in the RCS during a RCP startup. These pressure surges can be controlled in one of two ways.

One method would be to require a steam bubble in the pressurizer and thus control pressure using pressurizer controls. The other method would be to limit the temperature difference (< 50*F) between the RCS average temperature and the idle pump's cold leg water temperature.

3.1.1.2 Steam Generator At least two steam generators shall be operable whenever the average primary coolant temperature is above 350*F.

Basis One steam generator capable of performing its heat transfer function will provide sufficient heat removal capability to remove core decay heat after a normal reactor shutdown. The reactor cannot be made critical without water in all three steam generators, since the low-low steam generator water level trip prevents this mode of operation. Two operable steam generators are therefore adequate.

3.1-3 Amendment No. 87

(HBR-11) 3.1.1.3 Pressurizer (Pzr)

a. At least one Pzr code safety valve shall be operable whenever the Reactor Head is on the vessel and the RCS is not open for maintenance.

b. The ?zr, including necessary spray and heater control systems, shall be operable before the reactor is made critical.

c.

Whenever the RCS temperature is above 350*F or the reactor is critical:

1. All three pressurizer code safety valves shall be operable.

Their lift settings shall be maintained between 2485 psig and 2560 psig.

2.

At least 125 kw of pressurizer heaters capable of being powered from an emergency power source shall be operable.

d. If the requirements of 3.1.1.3.c.2 are.not met and at least 125 kw or Pzr heaters capable of being powered from an emergency source cannot be provided within 72 hrs., commence a normal plant shutdown and cooldown to an RCS average temperature of less than or equal to 350*F.

Basis The pressurizer is necessary to maintain acceptable system pressure during normal plant operation, including surges that may result following anticipated transients.

Each of the pressurizer code safety valves is designed to relieve 288,000 lbs.

per hr. of saturated steam at the valve setpoint.(1)

Below 350*F and 450 psig in the Reactor Coolant System (RCS), the Residual Heat Removal System can remove decay heat and thereby control system temperature. The pressurizer 3.1-3a Amendment No. 87

(HBR-11) safety valves are sized to protect the RCS against overpressure without taking credit for the steam bypass system.(2)

If no residual heat were removed by any of the means available, the amount of steam which could be generated at safety valve relief pressure would be less than the capacity of a single valve. One valve therefore provides adequate defense against overpressurization of the RCS for primary coolant temperatures less than 350*F and two valves provide protection for any temperature.

ASME Section III of the Code allows a maximum variation in the setpoint of 3 percent above the design set pressure.

The requirement that 125 kw of pressurizer heaters and their associated controls be capable of being supplied electrical power from an emergency power source provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation at hot shutdown.

References (1) FSAR Table 5.4.6-1

-)(2) FSAR Section 15.2 3.1-3b Amendment No. 87

(HBR-11) 3.1.3 Minimum Conditions for Criticality 3.1.3.1 Except during low power physics tests, the reactor shall not be made critical at any temperature, above which the moderator temperature coefficient is greater than:

a)

+5.0 pcm/*F less than 56% of rated power, or b)

+5.0 pcm/*F at 50% of rated power and linearly decreasing to 0 pcm/*F at rated power.

3.1.3.2 In no case shall the reactor be made critical above and to the left of the criticality limit shown on Figure 3.1-1.

3.1.3.3 When the reactor coolant temperature is in a range where the moderator temperature coefficient is greater than as specified in 3.1.3.1 above, the reactor shall be subcritical by an amount equal to or greater than the potential reactivity insertion due to depressurization.

3.1.3.4 The reactor shall be maintained subcritical by at least 1% until normal water level is established in the pressurizer.

Basis During the early part of fuel cycle, the moderator temperature coefficient may be slightly positive at low power levels.

The moderator coefficient at low temperatures or powers will be most positive at the beginning of the fuel cycle, when the boron concentration in the coolant is the greatest.

At all times, the moderator coefficient is calculated to be negative in the high power operating range, and after a very brief period of power operation, the coefficient will be negative in all circumstances due to the reduced boron concentration as Xenon and fission products build into the core.

The requirement that the reactor is not to be made critical when the moderator coefficient-is more positive than as specified in 3.1.3.1 above has been imposed to prevent any unexpected power excursion during normal operations as a result of either an increase of moderator temperature or decrease of coolant 3.1-11 Amendment No. 37

(HBR-11) pressure. This requirement is waived during low power physics tests to permit measurement of reactor moderator coefficient and other physics design parameters of interest.

During physics tests, special operating precautions will be taken.

In addition, the strong negative Doppler coefficient(1) and the small integrated Wk/k would limit the magnitude of a power excursion resulting from a reduction of moderator density.

The heatup curve of Figure 3.1-1 includes criticality Limits which are required by 10 CFR Part 50, Appendix G, Paragraph TV.A.2.c. Whenever the core is critical., additional safety margins above those specified by the AS4E Code Appendix G methods are imposed. The core may be critical at temperatures equal to or above the minimum temperature for the inservice hydrostatic pressure tests as calculated by ASIHE Code Appendix G methods, and an additional safety margin of 40*? must be maintained above the applicable heatup curve at all times.

If the specified shutdown margin is maintained (Section 3.10), there is no possibility of an accidental criticality as a result of an increase of moderator temperature or a decrease of coolant pressure.(I)

The requirement for bubble formation in the pressurizer when the reactor has passed the threshold of one percent subcriticality will assure that the Reactor Coolant System will not be solid when criticality is achieved.

References (1) FSAR Section 4.3 3.1-12 Amendment No. 87

(HIBR-ll).

TABLE 3.5-1 ENGINEERED SAFETY FEATURE SYSTEM INITIATION INSTRUMENT SETTING LIMITS NO.

FUNCTIONAL UNIT CHANNEL ACTION SETTING LIMIT

1.

High Containment Pressure (HI Level)

Safety Injection*

< 5 paig

2.

High Containment Pressure (HI-HI Level)

a. Containment Spray**

< 25 paig

b. Steam Line Ipolation

3.

Pressurizer Low Pressure Safety Injection*

> 1700 paig

4.

High Differential Pressure Between any Safety Injection*

< 150 psi Steam Line and the Steam Line Header

5.

High Steam Flow in 2/3 Steam Lines***

a. Safety Injection*

< 40% (at zero load) of

b. Steam Line Isolation full steam flow

< 40% (at 20% load) of full steam flow

< 110% (at full load) of full steam flow Coincident with Low T or Low Steam

> 541OF T avg avg Line Pressure

> 600 paig steam line pressure C-I-

6.

Loss of Power

a. 480V Emerg. Bus Undervoltage Trip Normal Supply Breaker 328 Volts + I Volt (Loss of Voltage) Time Delay

.75 +.25 sec.

(HBR-11)

TABLE 3.5-1 (Continued)

ENGINEERED SAFETY FEATURE SYSTEM INITIATION INSTRUMENT SETTING LIMITS NO.

FUNCTIONAL UNIT CHANNEL ACTION SETTING LIMIT

6.

b. 480V Emerg. Bus Undervoltage Trip Normal Supply Breaker 412 Volts + I Volt (Cont'd)

(Degraded Voltage) Time Delay 10.0 Second Delay +

0.5 sec.

7.

Containment Radioactivity High Ventilation Isolation

< 2 X Reading at the Time the Alarm is Set with Known Plant Conditions CO Initiates also containment isolation (Phase A), feedwater line isolation and starting of all containment fans.

Initiates also containment isolation (Phase B).

Derived from equivalent AP measurements.

(HBR-11)

TABLE 3.5-3 (Continued)

INSTRUMENTATION OPERATING CONDITIONS FOR ENGINEERED SAFETY FEATURES 3

1 2

OPERATOR ACTION M INIM4 MINIMUM IF CONDITIONS OF CHANNELS DEGREE OF COLUMN I OR 2 NO.

FUNCTIONAL UNIT OPERABLE REDUNDANCY CANNOT BE MET

2.

CONTAINIENT SPRAY

a.

Manual*

2 0**

Cold Shutdown

b.

High Containment Pressure*

2/set 1/set Cold Shutdown (Hi-Hi Level)

3.

LOSS OF POWER

a. 480V Emerg. Bus Undervoltage 2/bus I/bus Main Hot Shutdown (Loss of Voltage)

b. 480V Emerg. Bus Undervoltage 2/bus 1/bus Maintain Hot Shutdown(c)

(Degraded Voltage) me 0C

• Also initiates a Phase B containment isolation.

c* Must actuate two switches simultaneously.

When primary pressure is less than 2000 psig, channels may be blocked.

'When primary temperature is less than 547 0 F, channels may be blocked.

In this case the 2/3 high steam flow Is already in the trip mode.

(a) During testing and maintenance of one channel, may be reduced to 1/bus.

(b) During testing and maintenance of one channel, may be reduced to 0/bus.

(c) The reactor may remain critical below the power operating conditions with this feature inhibited for the purpose of starting reactor coolant pumps.

(RBR-09) 3.6 CONTAINMENT SYSTEM Applicability Applies to the integrity of reactor containment.

Obiective To define the operating status of the reactor containment for plant operation.

Specification 3.6.1 Containment Integrity

a. The containment integrity (as defined in 1.7) shall not be violated unless the reactor is in the cold shutdown condition.

b. The containment integrity shall not be violated when the reactor vessel head is removed unless a shutdown margin greater than 10% Sk/k is constantly maintained.

c. Positive-reactivity changes shall not be made by rod drive motion when the containment integrity is not intact except during any one of the following evolutions:.

1.

rod drop timing test

2.

rod drive mechanism timing test

3.

control rod exercise test

4.

shutdown banks fully withdrawn and control banks withdrawn to < 5 steps.

3.6-1 Amendment No. 87

(E3R-09)

During any of the aforementioned evolutions the shutdown margin shall be maintained > 1% dk/k.

d.

Positive reactivity changes shall not be made by boron dilution when the containment integrity is not intact unless the shutdown margin is maintained > 1% Ak/k.

3.6.2 Internal Pressure If the internal pressure exceeds I psig or the internal vacuum exceeds 1.0 psig, the condition shall be corrected within eight (8) hours or the operator shall start to place the reactor in the hot shutdown condition utilizing normal operating procedures.

3.6.3 Containment Automatic Isolation Trip Valves The following exceptions apply only to automatic containment isolation valves required to be closed during accident conditions-and which are either redundant or installed in a line which is part of a closed system within containment.

With one or more of the automatic containment isolation trip valves inoperable, either:

a.

Restore the inoperable valve(s) to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or

b. Isolate the affected penetration(s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of a deactivated automatic valve(s) secured in the isolation position(s), or

c. Isolate the affected penetration(s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of a closed manual valve(s) or blind flange(s),

or

d. Be in cold shutdown within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Amendment No. 87

3.

6-2

(HBR-09)

Basis The Reactor Coolant System conditions of cold shutdown assure that no steam will be formed and hence there would be no pressure buildup in the containment if the Reactor Coolant System ruptures.

The shutdown margins are selected based on the type of activities that are being carried out. The 10% Ak/k.shutdown margin during refueling precludes criticality, even though fuel is being moved.

When the reactor head is not to be removed, the specified cold shutdown margin of 1% Ak/k precludes criticality.

Regarding internal pressure limitations, the containment design pressure or 42 psig would not be exceeded if the internal pressure before a major loss-of coolant accident were as much as 2 psig.(1)

The containment is designed to withstand an internal vacuum of 2.0 psig.(2)

References (1) FSAR Section 6.2.1 (2) FSAR Section 3.8.1.3 3.6-3 Amendment No. 87

0 (HBR-11)

The relative humidity (R.H.)

of the air processed by the refueling filter systems should be less than the R.H. used during the testing of the charcoal adsorbers in order to assure that the adsorbers will perform under accident conditions as predicted by the test results. Heaters have been installed upstream of the Spent Fuel Building filters to assure of an R.H. of less than 70 percent for the air processed by the Spent Fuel Building filter system. If an R.H. in the containment atmosphere exceeds 70 percent, operation of the containment purge system will be terminated until this specification can be met.

If the Spent Fuel Building filter system is found to be inoperable, all fuel handling and fuel movement operations in the Spent Fuel Building will be terminated until the system is made operable.

The temperature limit specified for the fuel cask handling crane is based on the recorded ambient temperature at the time of the 125% load test.

The limit is imposed to assure adequate toughness properties of the crane structural materials.

References (1) FSAR Section 9.4.1 (2) FSAR Section 4.3 (3) FSAR Section 9.4.1 (4.)

H. B. Robinson Unit 2 Radiological Assessment of Postulated Accidents, KN-NF-84-68(P), July 1984.

3.8-6 Amendment No. 87

(HBR-11) 3.10.1.5 Except for physics tests, if a full length control rod is withdrawn as follows:

at positions > 200 steps and is > 15 inches out of alignment with its bank position, or at positions < 200 steps and is > 7.5 inches out of alignment with the average of its bank position then within two hours, perform the following:

a. Correct the situation, or

b. Determine by measurement the hot channel factors and apply Specification 3.10.2.1, or

c. Limit power to 70 percent of rated power 3.10.1.6 Insertion limits do not apply during physics tests or during period exercise of individual rods.

However, the shutdown margin indicated in Figure 3.10-2 must be maintained, except during the low power physics test to measure control rod worth.and shutdown margin. For this test, the reactor may be critical with all but one full'length control rod inserted.

3.10.2 Power Distribution Limits 3.10.2.1 At all times except during low power physics tests, the hot channel factors, F Q(Z) and FAH, defined in the basis, must meet the following limits:

F Q(Z) < (2.32/P) x K(Z) for P > 0.5 F Q(Z) < 4.64 x K(Z) for P < 0.5 F

< 1.65 (1 + 0.2(1-P) )

3.10-2 Amendment No. 37

(KBR-11) where P is the fraction of rated power (2300 Mwt) at which the core is operating. F (Z) is the measured F (Z) including the measurement uncertainty factor F = 1.05 and the engineering factor FE 1.03.

u Q

P is the measured F including a 1.04 measurement uncertainty factor. K(Z) is based on the function given in Figure 3.10-3, and Z is the axial location of F.

3.10.2.1.1 Following initial loading, or upon achieving equilibrium conditions after exceeding by 10% or more of rated power, the power FQCZ) was last determined, and at least once per effective full power month, power distribution maps using the movable detector system, shall be made to confirm that the hot channel factor limits of Specification 3.10.2.1 are satisfied and to establish the target axial flux difference as a function of power level (called the target flux difference).*

If either measured hot channel factor exceeds the specified limit, the reactor power shall be reduced so as not to exceed a fraction equal to the ratio of the F (Z) or F limit to the measured value, whichever is less, and the high neutron flux trip setpoint shall be reduced by the same ratio.

If subsequent incore mapping cannot, within a 24-hour period, demonstrate that the hot channel factors are met, the overpower AT and overtemperature AT trip setpoints shall be similarly reduced.

3.10.2.2 F Q(Z) shall be determined to be within the limit given in 3.10.2.1 by satisfying the following relationship for the middle axial 80%

of the core at the time of the target flux determination:

F () <2.32 K(Z FQ(Z) < (232) [-]

for P > 0.5 Q

P VZ F Q(Z) < 4.64 [

] for P < 0.5

• During power escalation at the beginning of each cycle, the design target may be used until a power level for extended operation has been achieved.

3.10-3 Amendment Mo. 87

(HBR-1 1) where V(Z) is defined in Figure 3.10-4 which corresponds to the target band and P > 0.5.

3.10.2.2.1 If the relationship specified in 3.10.2.2 cannot be satisfied, one of the following actions shall be taken:

a)

Place the core in an equilibrium condition where the limit in 3.10.2.2 is satisfied and re-establish the target axial flux difference b) Reduce the reactor power by the maximum percent calculated with the following expression for the middle axial 80% of the core:

F (Z) x V(Z)

[ [max. over Z of Q

-1 ] x 100%

2.32 x K(Z) c) Comply with the requirements of Specification 3.10.2.2.2.

3.10.2.2.2 The Allowable Power Level above which initiation of the Axial Power Distribution Monitoring System (APDMS) is required is given by the relation:

2.32 x K(Z)

APL = minimum over Z of F (Z) x V(Z) x 100%

where F (Z) is the measured F (Z), including the engineering factor F Q 1.03 and the measurement uncertainty factor FN 1.05 Q

U at the time of target flux determination from a power distribution map using the movable incore detectors. V(Z) is the variation function defined in Figure 3.10-4 which corresponds to the target band. K(Z) is the function defined in Figure 3.10-3.

The above limit is not applicable in the following core plane regions.

1) Lower core region 0% to 10% inclusive.

2) Upper core region 90% to 100% inclusive.

3.10-4 Amendment No.

'7

(HBR-11)

At power levels in excess of APL of rated power, the APIMS will be employed to monitor F Q(Z).

The limiting value is expressed as:

2.103/P 3

max R.

(1 + a.)

where:

a. P is the fraction of rated power (2300 Mwt) at which the core is operating (P < 1.0).

b. R., for thimble j, is determined from core power maps and is by definition:

6 F

1 Ij R

t E

F(Z)

.S(Z) i=1 ij max F Qj is the value obtained from a full core map including S(Z),

QJ N

but without the measurement uncertainty factor F or the engineering uncertainty factor, F.

The quantity F(Z)ij S(Z) is the measured value without inclusion of the instrument a

N uncertainty factors F.

Those uncertainty factors, F N 1.05, a

Q*E F = 1.02, as well as the engineering factor F = 1.03, have been included in the limiting value.of 2.103/P.

c. a. is the standard deviation associated with the determination of R..

d. S(Z) is the inverse of the K(Z) function given in Figure 3.10-3.

This limit is not applicable during physics tests and excore detector calibrations.

3.10.2.2.3 With successive measurements indicating the enthalpy rise hot N

channel factor, F N, to be increasing with exposure, the total peaking factor, F (Z), shall be further increased by two percent over that specified in Specifications 3.10.2.2, 3.10.2.2.1, and 3.10-5 Amendment No. 87

(HBR-11) 3.10.2.2.2 or FQ(Z) shall be measured and a target axial flux difference re-established at least once every seven (7) effective full power days until two successive measurements indicate enthalpy rise hot channel factor, F6, is not increasing.

3.10.2.3 The reference equilibrium-indicated axial flux difference as a function of power level (called the target flux difference) shall be determined in conjunction with the measurement of F (Z) as defined in Specification 3.10.2.1.1.*

3.10.2.4 The indicated axial flux difference shall be considered outside of the limits of Sections 3.10.2.5 through 3.10.2.9 when more than one of the operable excore channels are indicating the axial flux difference to be outside a limit.

3.10.2.5 Except during physics tests, and except as modified by 3.10.2.6 through 3.10.2.9 below, the indicated axial flux difference shall be maintained within the applicable target band about the target flux difference (defines the target band on axial flux difference).

3.10.2.6 At a power level greater than 90 percent or 0.9 x APL** (whichever is less) of rated power, if the indicated axial flux difference deviates from its target band, the flux difference shall be returned to the target band immediately or reactor power shall be reduced to a level no greater than 90 percent or 0.9 x APL (whichever is less) of rated power.

3.10.2.7 At a power level between 50 percent and 90 percent or 0.9 x APL (whichever is less) of rated power, During power escalation at the beginning of each cycle, the design target may be used until a power level for extended operation has been achieved.

• • APL is the Allowable Power Level defined in Specification 3.10.2.2.2.

3.10-6 Amendment No. 37

(HBR-1 1)

a. The indicated axial flux difference may deviate from its target band for a maximum of one hour (cumulative) in any 24-hour period provided the flux difference does not exceed the limits shown in Figure 3.10-5.

If the cumulative time exceeds one hour, then the reactor power shall be reduced immediately to no greater than 50 percent of rated power and the high neutron flux setpoint reduced to no greater than 55 percent of rated power.

b. A power increase to a level greater than 90 percent or 0.9 x APL (whichever is less) of rated power is contingent upon the indicated axial flux difference being within its target band.

3.10.2.8 At a power level no greater than 50 percent of rated power

a. The indicated axial flux difference may deviate from its target band.

b. A power increase to a level greater than 50 percent of rated power is contingent upon the indicated axial flux difference not being outside its target band for more than two hours (cumulative) out of the preceding 24-hour period.

One-half of the time the indicated axial flux difference is out of its target band up to 50 percent of rated power is to be counted as contributing to the one-hour cumulative maximum the flux difference may deviate from its target band at a power level less than or equal to 90 percent or 0.9 x APL (whichever is less) of rated power.

3.10.2.9 Calibration of the excore detectors will be performed at a power level no greater than 90% or 0.9 x APL (whichever is less) of rated power.

The indicated axial flux difference may deviate from its target band during the calibration provided the flux difference does not exceed the limits shown in Figure 3.10-5.

3.10-7 Amendment No. 87

(HBR-ll) 3.10.2.10 Alarms shall normally be used to indicate non-conformance with the flux difference requirement of 3.10.2.6 or the flux difference time requirement of 3.10.2.7.a.

If the alarms are temporarily out of service, the axial flux difference shall be logged, and conformance with the limits assessed, every hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and half-hourly thereafter.

3.10.2.11 The axial flux difference target band about the target axial flux difference shall be determined in conjunction with the measurement of FQ(Z) as specified in 3.10.2.1.1. The allowable values of the target band are shown in Figure 3.10-4. Redefinition of the target band from more restrictive to less restrictive ranges between determinations of the target axial flux difference is allowed when appropriate redefinitions of APL are made.

Redefinition of the target band from less restrictive to more restrictive ranges is allowed only in conjunction with the determination of a new target axial flux difference.

3.10.3 Quadrant Power Tilt Limits 3.10.3.1 Except for physics tests and during power increases below 50 percent of rated power, whenever the indicated quadrant power tilt ratio exceeds 1.02, the tilt condition shall be eliminated within two hours or the following actions shall be taken:

a. Restrict core power level and reset the power range high flux setpoint to be less two percent of rated values for every percent of indicated power tilt ratio exceeding 1.0, and

b. If the tilt condition is not eliminated after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the power range high flux setpoint shall be reset to 55 percent of rated power. Subsequent reactor operation would be permitted up to 50 percent of rated power for the purpose of measurement and testing to identify the cause of the tilt condition.

3.10-7a Amendment No. 37

(HBR-11) 3.10.3.2 Except for low power physics tests, if the indicated quadrant tilt exceeds 1.09 and there is simultaneous indication of a misaligned rod:

a. The core power level shall be reduced by 2 percent of rated values for every 1 percent of indicated power tilt exceeding 1.0, and

b. If the tilt condition is not eliminated withid two hours, the reactor shall be brought to a hot shutdown condition.

c. After correction of the misaligned rod, reactor operation will be permitted to 50 percent of rated power until the indicated quadrant tilt falls below 1.09.

3.10.3.3 If the indicated quadrant tilt exceeds 1.09 and there is not a simultaneous indication of rod misalignment, except as stated in Specification 3.10.3.2.c, the reactor shall immediately be brought to a hot shutdown condition.

3.10-7b Amendment No. 87

(HBR-11) equilibrium conditions in terms of fuel loading patterns and anticipated control bank worths. These measurements will augment the normal fuel cycle design calculations and place the knowledge of shutdown capability on a firm experimental as well as analytical basis.

Operation with abnormal rod configuration during low power and zero power testing is permitted because of the brief period of the test and because special precautions are taken during the test.

Two criteria have been chosen as a design basis for fuel performance related to fission gas release, pellet temperature, and cladding mechanical properties. First, the peak value of linear power density must not exceed 21.1 kW/ft.

Second, the minimum DNBR in the core must not be less than 1.17 in normal operation or in short term transients.

In addition to the above, the initial steady state conditions for the peak linear power for a Loss-of-Coolant Accident must not exceed the values assumed in the accident evaluation. This limit is required in order for the maximum clad temperature to remain below that established by the ECCS Acceptance Criteria. To aid in specifying the Limits on power distribution the following hot channel factors are defined.

a. FQ, Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

N

b. FN, Nuclear Heat Flux Hot Channel Factor, is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and rod dimensions.

c.

F E Engineering Heat Flux Hot Channel Factor, is defined as the allowance on heat flux required for manufacturing tolerances.

The engineering factor allows for local variations in enrichment, pellet density and diameter, surface 3.10-12 Amendment No. 37

(HBR-11)

e. Axial power distribution control procedures, which are given in terms of axial flux difference within the target band about the target flux difference, are observed. Flux difference refers to the difference in signals between the top and bottom halves of two-section excore neutron detectors.

The flux difference is a measure of axial offset which is defined as the difference in power between the top and bottom halves of the core.

For operation at a fraction P.of full power, the design limits are met, provided the limits of Specification 3.10.2.1 are not exceeded.

The permitted relaxation in F with reduced power allows radial power shape changes with rod insertion to the insertion limits. It has been determined that provided the above conditions a through e are observed, these hot channel factors limits are met.

The procedures for axial power distribution control referred to above include operator control of flux difference to minimize the effects of xenon redistribution on the axial power distribution during load-follow maneuvers.

Basically, control of flux difference is required to limit the difference between the current value of Flux Difference (Al) and a value which corresponds to the full power target flux difference established in conjunction with incore power distribution measurements. The target flux difference varies with power level.

The target value of the flux difference is determined at equilibrium xenon conditions.

The control rods must be positioned in accordance with their insertion limits.

This value, divided by the fraction of full power at which the core was operating is the full power value of the target flux difference. Values for all other core power levels are obtained by multiplying the full power value by the fractional power. Since the indicated equilibrium value was noted, no allowances for excore detector error are necessary and the specified deviation of Al is permitted from the indicated reference value. The periodic updating of the target flux difference is necessary to reflect the impacts of core burnup on power distribution.

3.10-14 Amendment No. 37

Strict control of the flux difference is not possible during certain physics tests, control rod exercises, or during the required periodic excore calibration which require larger flux differences than permitted. Therefore, the specifications on power distribution are not applicable during physics tests, control rod exercises, or excore calibrations; this is acceptable due to the extremely low probability of a significant accident occurring during these operations.

Excore calibration includes that period of time necessary to return to equilibrium operating conditions.. In some instances of rapid plant power reduction, automatic rod motion will cause the flux difference to deviate from the target band when the reduced power level is reached. This does not necessarily affect the xenon distribution sufficiently to change the envelope of peaking factors which can be reached on a subsequent return to full power within the target band; however, to simplify the specification, a limitation of one hour in any period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is placed on operation outside the band.

This ensures that the resulting xenon distributions are not significantly different from those resulting from operation within the target band.

The instantaneous consequence of being outside the band, provided rod insertion limits are observed, is not worse than a 10 percent increment in peaking factor for flux difference in the allowable range shown in Figure 3.10-5 for 90 percent or 0.9 x APL (whichever is less).

Therefore, while the deviation exists, the power level is limited to 90 percent or 0.9 x APL (whichever is less) of rated power or lower depending on the indicated flux difference.

If, for any reason, flux difference is not controlled with the target band for as long a period as one hour, then xenon distributions may be significantly changed and operation at 50 percent of rated power is required to protect against potentially more severe consequences of some accidents.

As discussed above, the essence of the limits is to maintain the xenon distribution in the core as close to the equilibrium full power condition as possible. This is accomplished by using the chemical volume control system to position the full length control rods to produce the required indication flux difference.

3.10-15 Amendment No. 37

(HBR-11)

An upper bound envelope of peaking factors has been determined from extensive analysis considering all operating maneuvers consistent with the technical specifications on power distribution control as given in Section 3.10.2 The specifications on power distribution control ensure that xenon distributions are not developed, which at a later time, could cause greater local power peaking even though the flux difference is then within limits. The results of a Loss-of-Coolant Accident analysis based on this upper bound envelope indicate that a peak clad temperature would not exceed the 2200*F limit. The nuclear analyses of credible power shapes consistent with the power distribution control procedures have shown that the F limit is not exceeded.

For transient events, the core is protected from exceeding 21.1 kw/ft locally, and from going below a minimum of DNBR of 1.17 by automatic protection on power, flux difference, pressure and temperature.

Measurements of the hot channel factors are required as part of startup physics tests and whenever abnormal power distribution conditions require a reduction.of core power to a level based on measured hot channel factors.

In the specified limit of F there is a 5 percent allowance for (5)

Q uncertainties which means that normal operation of the core within the N

defined conditions and.procedures is expected to result in a measured F 5 percent less than the limit, for example, at rated power even on a worst case basis.

When a measurement is taken, experimental error must be allowed for, and 5 percent is the appropriate allowance for a full core representative map taken with the movable incore detector flux mapping system.

N In the specified limit of F

, there is an 8 percent allowance for design prediction uncertainties,.which means that normal operation of the core is N

expected to result in F at least 8 percent less than the limit at rated AH N

power.

The uncertainty to be associated with a measurement of FAH by the movable incore system, on the other hand, is 4 percent, which means that the N

normal operation of the core shall result in a measured F at least 4 percent less than the value at rated power. The logic behind the larger design uncertainty in the case is that (a) abnormal perturbation in the radial power N

shape (e.g., rod misalignment) affects F in most cases without necessarily 3.10-16 Amendment No. 37

(N

()

wieteoeao affecting FQ, and can limit it to the desired value; (b) while the operator has some control over FN through FN by motion of control rods, he has no NQ z

direct control over F

, and (c) an error in the predictions for radial power shape, which may be detected during startup physics tests, can be compensated for in FN by tighter axial control, but compensation for F is less readily available.

Quadrant power tilts are based upon the following considerations.

The radial power distribution within the core must satisfy the design values assumed for calculation of power capability. Radial power distributions, measured as part of the startup physics testing, are periodically measured at a monthly or greater frequency. These measurements are taken to assure that the radial power distribution with any quarter core radial power asymmetry conditions is consistent with the assumptions used in power capability analyses.

It is not intended that extended reactor operation would continue with a power tilt condition which exceeds the radial power asymmetry considered in the power capability analysis.

During normal plant startup, quadrant power tilt ratio may exceed 1.02 due to instrumentation instabilities as a result of rodded configurations and low excore detector signal levels below 50 percent of full power. Sustained power operation below 50 percent of full power would require-a renormalization of the calculational methods for determining power tilt to compensate for change in signal levels once equilibrium conditions are met.

The two-hour time interval in this specification is considered ample to identify a dropped or misaligned rod and complete realignment procedures to eliminate the tilt.

In the event that the tilt conditions cannot be eliminated within the two-hour time allowance, additional time would be needed to investigate the cause of the tilt condition. The measurements would include a full core physics-map utilizing the movable detector system. For a tilt condition < 1.09 an additional 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />' time interval is authorized to accomplish these measurements. However, to assure that the peak core power is maintained below limiting values, a reduction of reactor power of two percent 3.10-17 Amendment No. 37

(HBR-11) for each one percent of indicated tilt is required. Physics measurements have indicated that the core radial power peaking would not exceed a two-to-one relationship with the indicated tilt from the excore nuclear detector system for the worst rod misalignment.

In the event the tilt condition of 1.09 cannot be eliminated after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor power level will be reduced to the range required for low power physics testing. To avoid reset. of a large number of protection setpoints, the power range nuclear instrumentation would be reset to cause an automatic reactor trip at 55 percent of allowed power. A reactor trip at this power has been selected to prevent, with margin, exceeding core safety limits even with a nine percent tilt condition. If a tilt ratio greater than 1.09 occurs which is not due to a misaligned rod, the reactor power shall be brought to a hot shutdown condition for investigation.

However, if the tilt condition can be identified as due to rod misalignment, operation can continue at a reduced power (2 percent for each one percent the tilt ratio exceeds 1.0) for the two-hour period necessary to correct the rod misalignment.

The specified rod drop time is consistent with safety analyses that have been performed.(')

An inoperable rod imposes additional demands on the operator. The permissible number of inoperable control rods is limited to one in order to limit the magnitude of the operating burden, but such a failure would not prevent dropping of the operable rods upon reactor trip.

Normal reactor operation causes significant pellet cracking and fragmentation.

Consequently, handling of irradiated fuel assemblies can result in relocation of these fragments against the cladding. Calculations show that high cladding stresses can occur if the reactor power increase is rapid during the subsequent startup.

3.10-18 Amendment No. 87

(HBR-11)

The 72-hour period allows for stress relaxation of the clad before the ramp rate requirement is removed, thereby reducing the potential harmful effects of possible pellet or fragment relocation.

The 3 percent limit is imposed to minimize the effects of adverse cladding stresses resulting from part pcwer operation for extended periods of time.

The time period of 30 days is based upon the successful power ramp demonstrations performed on Zircaloy clad fuel in operating reactors, resultihg in no cladding failures.

References (1) FSAR Section 15.0 (2) FSAR Section 7.7 (3) FSAR Section 15.4 (4) FSAR Section 15.4 (5) FSAR Section 15 3.10-19 Amendment No. 87

100 G

/

GO'4UF C

'00 LU_

LU

~G QOU DI 0

0 20 40 60 80 100 POWER LEVEL (PERCENT)

BOL EOL Amendment No.

FIGURE H.B. ROBINSON Unit #2 Control Group Insertion Limits For Carolina Three Loop Operation 3.10-1 Power & Light Company Technical Specifications 3.10-20 Amendment No.

37

1.2 LARGE BREAK CURVE LARGE BREAK CURVE (0.5.1.0)

FUEL ROD BURNUPF 1.0 1.0

>9 MWD/kgU LARGE BREAK CURVE

(.,.2 0.8-FUEL ROD BURNUP (0.8,0.80)

<9 MWD/kgU 0

U~U.

0.6 cc CD 0 O SMALL BREAK LIMITING CD N

0.4 (1.0,0.36) 0.2 -

I 0

1.20 2.40 3.60 4.80 6.00 7.20 8.40 9.60 10.80 12.00 ELEVATION IN CORE (ft)

NORMALIZED AXIAL DEPENDENCE FACTOR FOR Fo 2.32 VERSUS ELEVATION FOR FH:1. 6 5 Figure 3.10-3

00 1.18 1.16 (11.25,1.15) 5% TARGET BAND 11.25,1.12)

(9.25.1.11)

,o 1.0 1

1 3% TARGET BAND (9.25,1.08) 08 1.04T-02 i

0 2

4 6

8 10 12 AXIAL HEIGHT (FEET)

V(Z)AS A FUNCTION OF CORE HEIGHT Figure 3.10-4 3.10-23 Amendment No.

87

II-----------------

I1 I

UNACCEPTBLE UNACCEPTABLE 2

OPERATION

(-10,90)

(+10,90)

OPERATION I Z-

-a 90)

a.

Lu r

(-8,90)

(+8,90) a 80-2 c

+G%

TARGET 5TAG 2

x BAND ACCEPTABLE OPERATION BAND

-(-25,50)

(+25,50)

(-23,50

(+23,50) 0 4 3-2I 101 2

30 i-g 0

3.100-2 Amendment N. 8

-3.-0 10-2 0m nmn

10.

2073

0

• (HBR-11) 3.11 MOVABLE IN-CORE INSTRG4ENTATION Applicability Applies to the operability of the movable detector instrumentation system.

Objective To specify functional requirements on the use of the in-core instrumentation systems, for the calibration of the excore symmetrical offset detection system.

Specification 3.11.1 A minimum of 16 total accessible thimbles and at least 2 per quadrant and sufficient movable in-core detectors shall be operable during recalibration of the excore symmetrical offset detection system.

3.11.2 Power shall be limited to 90% of rated power if recalibration requirements for the excore symmetrical offset detection system identified in Table 4.1-1 are not met.

Basis The Movable In-Core Instrumentation System l) has five drives, five detectors, and 48 thimbles in the core. Each detector can be routed to twenty or more thimbles.

Consequently, the full system has a great deal more capability than would be needed for the calibration of the excore detectors.

To calibrate the excore detector system, it is only necessary that the Movable In-Core System be used to determine the gross power distribution in the core as indicated by the power balance between the top and bottom halves of the core.

3.11-1 Amendment No. 37

(HBR-11)

After the excore system is calibrated initially, recalibration is needed'only infrequently to compensate for changes in the core, due, for example, to fuel depletion, and for changes in the detectors.

If the recalibration is not performed, the mandated power reduction assures safe operation of the reactor since it sill compensate for an error of 10% in the excore protection system. Experience at the Beznau No. 1 and R. E. Ginna plants has shown that drift due to the core on instrument channels is very slight.

Thus, limiting the operating levels to 90% of the rated power i very conservative.

Reference (1) FSAR Section 7.7.1.5 3.11-2 Amendment No..87

(BR-11) 4.11 REACTOR CORE Applicability Applies to surveillance of the reactor core.

Objective To ensure the integrity of the fuel cladding.

Specification 4.11.1 APEMS Operation 4.11.1.1 Prior to establishing normal operation with APDMS, at least six maps will be taken to determine applicable values of R and a for surveillance thimbles.

4.11.1.2 Plant operation up to rated power shall be permitted for the purposes of obtaining the initial maps of Specification 4.11.1.1, provided the APDMS is operational and hot channel factors are shown to be below the limiting values set forth in Specification 3.10.2.

Suitably conservative values of R and a shall be derived from maps previously run during the current fuel cycle for use in the APDMS system during this initial period.

4.11.1.3 Subsequent updates of T and a shall employ the last six maps in accordance with Specification 4.11.1.1.

4.11.1.4 Each power distribution map will be based on flux traverses obtained from 36 or more of the 48 monitoring channels.

4.11.2 Except during physics tests and EXCORE calibrations, axial surveillance of F(Z)S(Z) shall consist of traverses with the movable incore detectors in appropriate pairs of detector paths, taken every eight hours, or a frequency of approximately 0, 10, 4.11-1 Amendment No. 37

(HBR-11) 30, 60,

120, 180, 240, 360, and 480 minutes following accumulated control rod motion in any one direction of five steps or more, exclusive of control rod movement within 15 steps from the top of the core. From the traverses, determination of F(Z)S(Z) shall be made and shown to result in a value less than the limiting value specified in 3.10.2. If the APDMS is out of service, reactor operation above APL of rated power can be continued for fourteen equivalent full power days provided that traverses are taken manually at equivalent frequencies, and a log of accumulated rod motion and time of manual traverses is kept.

4.11.3 The following criteria will be used for selecting the channels for measuring F(Z)S(Z):

a. The channel is not acceptable if it contains a control rod allowed by the insertion limits at power levels requiring APDMS.

b. For the latest full core power map, i, channels, j, are acceptable if:

R.

R i

< 20 R.

Basis The R technique provides a means for using many of the monitoring thimbles to determine F Q(Z) without fully mapping the core. Frequent core maps assure that appropriate values of T are being used for each thimble.

Upon return to power following a refueling outage or other situation where establishment of normal APDMS operation is required, power operation above APL of rated power is desirable to establish hot channel factors at full power.

4.11-2 Amendment No. 87

(HBR-I1)

By using maps that have been previously obtained during the power ascension and deriving conservative values of i and a from these maps for use in the APDMS, operation of the plant within the peaking factor limitations can be ensured.

If the APDMS is out of service, adequate monitoring of the core power distribution can be maintained for a limited period of time by manual actuation of the flux mapping system and calculation of the values of F(Z)S(Z).

4.11-3 Amendment No. 87

(HBR-1>1) 5.3 REACTOR 5.3.1 Reactor Core 5.3.1.1 The reactor core contains approximately 68 metric tons of uranium in the form of natural or slightly enriched uranium dioxide pellets.

The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods which are all pre-pressurized. The reactor core is made up of 157 fuel assemblies. Each fuel assembly contains 204 fuel rod locations occupied by rods consisting of natural or slightly enriched uranium pellets, solid inert materials, or a combination of the aforementioned.(1 )

5.3.1.2 Deleted.

5.3.1.3 Reload fuel will be similar in design to the initial core. The enrichment of reload fuel will be no more than 3.9 weight percent of U-235.

5.3.1.4 Deleted.

5.3.1.5 There are 45 full-length RCC assemblies-in the reactor core. The full-length RCC assemblies contain 144 inch segments of silver indium-cadmium alloy clad with the stainless steel.(2) 5.3.1.6 Up to 10 grams of enriched fissionable material may be used either in the core, or available on the plant site, in the form of fabricated neutron flux detectors for the purposes of monitoring core neutron flux.

5.3.2 Reactor Coolant System 5.3.2.1 The design of the Reactor Coolant System complies with the Code requirements.(3) 5.3-1 Amendment No. 87

(BBR-11) 5.3.2.2 All piping, components and supporting structures of the Reactor Coolant System are designed to Class I requirements.

5.3.2.3 The nominal liquid volume of the Reactor Coolant System, at rated operating conditions, is 9343 cubic feet. (4 References (1) FSAR Section 4.2.1 (2) FSAR Section 4.2.2 (3) FSAR Table 3.2.2-1 (4) FSAR Table 5.1.0-1 5.3-2 Amendment No. 87