ML20091F479

From kanterella
Jump to navigation Jump to search
Amends 149 & 133 to Licenses NPF-4 & NPF-7,respectively, Revising Tech Specs Re Control Rod Insertion Limits
ML20091F479
Person / Time
Site: North Anna  Dominion icon.png
Issue date: 11/22/1991
From: Berkow H
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20091F483 List:
References
NUDOCS 9112060376
Download: ML20091F479 (31)
v • d • e


Text

-___-

[ga meo [o, UNITED STATES NUCLEAR REGULATORY COMMISSION-e o

g j

WASHINGTON. D. C. 20555

%...+ /

VIRGINIA ELECTRIC AND POWER COMPANY O!O DOMINION ELECTPIC COOPERATIVE DOCKET NO. 50-338 NORTH ANNA POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICEN,S_EE Amendment No.149 License No, t!PF-4 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated August 11, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), ano the Comission'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 Comission; 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 Comission's regulations; D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

9112060376 911122 PDR ADOCK 05000338 p

PDR

'c

]

1

-?-

2.

Accordingly, the license is amended by changes to the Technical Speci-ficatinns as indicated in the attachment to this license amendment.

- and te graph 2.0.(2) of Facility Operating License No, NPF-4 is hereby anend;. to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.149, 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 issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMf11SS10N H rbert N. Berkow, Director Project Directorate 11-?

Division of Reactor Projects - 1/11-Office of Nuclear Reactor Regulation

Attachment:

Chances to the Technical Specifications Date of Issuance:

November-22, 1991 l

l I

i

s' ATTACHMENT TO LICENSE At1ENDMENT NO.

149 TO FACILITY OPERATING LICillSE NO.-NPF-4 l

DOCKET NO. 50-338 Replace the following pages of the Appendix "A" Teclaical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness, Pace W

1-5 B 2-2 3/41'3 3/4 10-1 B 3/4 1-4 i

.Pr

=

s.

M 4' 88 me

- er 0

1.0 DEFINITIONS (Continued)

ENGNEEREQ SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monttored param9ter exceeds itt ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e. the valves travel to their required positions, pump discharge pressures reach their required values, etc.;. Times shall 1

include diesel generator starting and sequence loading delays where applicable.

FREQUENCYNOTATION 1,13 The FREQUENCY NOTATION cpecified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

FULLY WITHORAWN 1.13a The control bank FULLY-WITHDRAWN position snall be within the interval of 225 to 229 steps withdrawn, inclusive. Definition of the FULLY WITHORAWN position for each specific cycle shall be documented in the rod insertion limit operator curve.

GASEOUS RADWASTE TREAThENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is the system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. The system is composed of the waste gas decay tanks, regenerative heat exchanger, waste gas charcoal filters, process vent blowers, waste gas surge tanks and waste gas diaphram compressor.

IDEfEIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing 'eaks that are captured and conducted to a sump or collecting tank, or b.

Leakage-into the containment atmosphere from sources that are both specifically-located ar.d known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or c.

Reactor coolant system leakage through a steam generator to the secondary system.

MEMBER (S) OF THE PUBUC 1.16 MEMBER (S) OF THE PUBLIC shall include all individuals who by virtue of their occupational status have no formal association with the plant. This category shall include non employees of the licensee who are permitted to use potilons of the site for recreational, occupational, or other purposes not associated with plant functions. This category shall an1 include non employees such as vending machine servicemen or postman who, as part of their formal job function, occasionally enter an area that is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.

NORTH ANNA - UNIT 1 1-3 Amendment No. 16.59,1/6, 149 i

l

's 1.0 DEFINillONS (Continued)

~

OFFSITE DQ$E CALCULAT!ON MANUAL t.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and l parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid efflimt monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Moe ring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and ?,adiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semi. annual Radioactive Efiluent Release Reports required by Specifications 6.9.1.8 and 6.9.1.9.

OPER ABLE. OPER ABILITY 1.18 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s), and when attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsysicm, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i e., MODE) shall corresoond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified f able 1.1.

PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the ftndamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.21 PRESSURE BOUNDARY LEAKnGE shall be Isakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall o wall.

PROCESS CCMROL PROGRAM 1.22 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling,

{

analyses, tests and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20,61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of the radioactive waste.

PURGE - PURGING 1.23 PURGE or PURGING is the controlled process of discharging air or gas from a l confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purity the confinement.

NORTH ANNA - UNIT 1 14 Amendment No. 75,JS,

INo146,

2-1.0 DEFfNITIONS -(Continued)-'

QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of-the maximum upper ex-core detector calibrated output to the average of the upper excore. detector calibrated outputs._ or the ratio of the maximum lower excore detector-calibrated-output to the average of the lower excore detector calibrated outputs, whichever-is greater.

With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2893 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME

!.26 The REACTOR. TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored. parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT 1.27 A REPORTABLE ~ EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

SHUTDOWN MARGIN-1.28 ~ SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from'its present condition assuming alltfull length rod. cluster assemblies (shutdown and control)'are fully-inserted except-for the single rod cluster assembly of highest reactivity worth which is assumed to_be FULLY WITHDRAWN.

SITE BOUNDARY 1.29~ The SITE BOUNDARY shall be that line beyond which the land-is not owned.,

. leased or otherwise controlled by the licensee.

SLAVE RELAY-TEST

'1.30 A SLAVE RELAY TEST shall be the energization of each slave relhy and verification of OPERABILITY of each relay. -The SLAVE RELAY TEST shall include a continuity check', as a minimum, of assoc 14ted testable actuation devices.

SOURCE CHECK 1.31 A_ SOURCE: CHECK shall be the qualitative assessment-of channel l response when the channel sensor is exposed-to radiation. This-applies to instal _ led radiation monitoring systems.

P

.. NORTH ANNA - UNIT;i

,_. i 1-5

_Amendagt No. JSJM3&%

t

1.0 DEFINITIOWS (Continued _)

i STACCERED TEST B_ ASIS 1.32 A STAGGERID TEST BASIS sha*1 ccmist of:

a.

A test schedule for a systems subsystems, tt'ains or other designated components obtained by dividing the specified test interval into a equal sUintervals, b.

The testing of one system, subsystem, train or other designated coepencut at the beginning of es.h subinterval.

THERMAL pOVEP.

1.33 THERMAL POWER shall be the total ree.: tot core heat transfer rate to the l

reactor coolant i

UNIDENTIFIED LEAKAGE 1.34 UNIDENTIT ED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNKESTRICTED M EA 1.35 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY l

where access is not controlled by the licensee for purposes of protection of individuals f rom exposure to radiation and radioactive materials or any area within the SITE BOVNDARY used for residential quarters or for industrial, connercial, institutiot.al, and/or recreational purposes.

VENTILATION EIRAUST TREATMENT SYSTEM l

1.36 A VENTILATION IIHAUST TREATMENT SYSTDi is the system designed and installed to reduce gaseous radiciodine or radiczetive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such s system is not considered to have any effect on noble gas effluents).

Engineered Safety Teature (EST) 'tmospheric cleanup systems are not considered to be VINTil.ATION EXHAUST TREA'lNIFT SYSTEM components.

VENTING 1.37 VENTING is the controlled crocess of discharging air or gas from a con finement to maintain tenperature, pressure, humidity, concentration or other operating condition, in such a asuner that replacement rir or gas is not provided or required during VENTING.

Vent, used in s/ stem names, does not imply a VENTING process.

NORTH ANNA - UNIT 1 1-6 Amendment No.15, 4 g ' 125

e a

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel. cladding is prevented by restricting fuel operation to within the nucleat; boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime-could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pree pre have been related to DNB through a correlation.

The DNB correlation has in developed f-to predict the DNB flux and the location of DNB for axially u...ot.a and non-uniform heat flux distributions.

The local DNB heat flux ratio DNBR, defined as the ratio of the heat flux : hat would cause DNB at a particular core location to the local heat flux, is indicative of the marQin to DNB, The DNB design basis is as follows:

there must be at least a 95 percent probability that:the minimum DNBR of the limiting rod during Condition I and 11 events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB vill not occur when the minimum DNBR is at the DNBR limit.

In meeting this design basis, uncer71nties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95% probability that the minimum DNBR for the limitir.g rod is (rtater than or equal to the ONBR limit. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty.

This DNBR ur. certainty, combined with the correlation DNBR l'imit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties. As an

(

additional criterion, meeting the DNBR limit also ensures that a least 99.9%

of the core avoids the onset.of DNB when the plant is at the DNBR limit.

The curves of' Figures 2.1 1, 2.1-2, and 'i.1 3 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and averag;. temperature for which the minimum DNBR is no less than the design limit DNBR, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid, NORTH ANNA - UNIT 1 B 2-1 Amendment No.M. 112 s

1

e SAFETY LIMITS BASES The curves are based on an enthal y hot channel factor, F)d, of 1 and a reference cosine with a peak of

.55 for axial power sha An 3

allowance is included for an increase in F at reduced power based on "H

a the expression:

IF f = 1.49 [l+ 0.3 (1.P))

t.H where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods FUl1Y WITHDRAWN to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f(a!) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power im-tilance effect on the Overtemperature at trips will reduce the setpoints to provide protection consistent with core safety limits, i

2.1.2 REACTOR COOLANT SYSTEM PRESSURE I

The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the l

release of radionuclides contained in the reactor coolent from reaching the containment atmosphere.

I The reactor pressure vessel and pressurizer are design'd to Section

I of the ASME Code fnr Nuclear Power Plant which permits a maximum

.transient pressure of 110% (2735 psi'g) of design pressura. The Reactor Coolent System piping, valves and fittings, were initially designed to ANSI B 31.1 1967 Edition and ANSI B 31.7 1969 Edition (Table 5.2.1-1 of FSAR) which permits a maximum transient pressure of 120% (29B5 psig) of component design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig.125%

of design pressure, to demonstrate integrity prior to initial operation.

l l

I NORTH ANNA-UNIT 1 B 2-2 Amendment No. 4 (, S.149 1

' REACTIVITY CONTROL SYSTEMS ROD OROP TIME LIMITING _ CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) rod drop time from the 229 STEP withdrawn position shall be < 2.7 seconds from beginning l

~

of decay of 5 tationary gripper coil voltage to dashpot entry with a.

T

> 500'F, and avg b.

All reactor coolant pumps operating.

APPLICABILITY:

0 DES 1 and 2.

ACTIOJ:

a.

With the drop time of any full length rod detennined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b.

With the rod drop times within limits but determined with 2 reactor coolant pumps operating, operation may proceed pr vided THERMAL POWER is restricted to:

1.

< 66% of RATED THERMAL POWER when the reactor coolant

~ top valves in the nonoperating loop are open, or 2.

< 71% of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are closed.

SURVEILLANCE REQUIREMENT 5 e

i 4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

a.

For all rods following each removal of the reactor vessel..ead, b.

For specifically affected individual rods following any main-tenance on or modification to the control rod drive system w;1ich could affect the drop time of those specific rods, and c.

At least once per 18 months.

NORTH ANNA-UNIT 1 3/4 1-23 Amendwnt No. NP, %9

REACTMTY OONTROL SYSTEMS SHUTDOWN ROD INSERTION UMIT UMITING CONDITON FOR OPERATION 3.1.3.5 All shutdown rods shall be limited in physical insertion as specified in the CORE OPEFMTING UMITS REPORT.

AE211CABILITY: MODES 1' and 2'#

M:

With a maximum of one shutdown rod inserted beyond the insertion limit specified in the CORE OPF. RATING UMITS REPORT, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

a. Restore the rod to within the insertion limit specified in the CORE OPERATING UMITS REPORT, or g
b. Declare the rod to be inoperable and apply Specification 3.1.3.1 4

SURVElt1ANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion !!mit specified in the CORE OPERATING UMITS REPORT

a. Within 15 minutes prior to in1tial control rod bank withdrawal during an approach to reactor criticality, and
b. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter, t

t,

  • See Special Test Exceptbns 3.10.2 and 3.10.3.
  1. With K,gg 21.0 NORTH ANNA UNIT 1 3/4 1 24 Amendment No. 138, 146

t l

3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGig LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided the reactivity equivalent to at least the highest estimated control rori worth is available for trip insertion from OPERABLE control rod (s).

APPLICABillTY: MODE 2.

ACTIONt a.

With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, boric acid solution or its equivale.,10 gpm of at least 12.950 ppm initiate and continue boration at y nt until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored, b.

With all full length control rods inserted and the reactor sub.

critical by less than the above reactivity equivalent, imedictely initiate and continue boration at 2 10 gpm of at least 12,950 ppm boric acid solution or its equivale,nt until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length rod either partially or rutLY WITHORAWN shall be detennined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4.10.1.2 Each full length rod that is not fully inserted shall be demonstrated capable of full insertion when tripped from at least 50%

withdrawn position within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the SHUTDOWN MARGIN to less than the limits of Sp?cification 3.1.1.1.

NORTH ANNA - UNIT 1 3/4 10-1 Amendment No./14 58, 149

4 1

SPECIAL TEST EXCEPTIONS GROUP HE!CHT INSERTION AND POWER O!STRIBUTION LIMITS LIMITING CONDITION FOR OPERATION

(

=

3.10.2 The group height, insertion and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, and 3s2.4 may be susbended t

during the performance of PHYSICS TESTS providedt Tne THERMAL POWER is maintained < 85% of kATED THERMAL POWER, a.

ar.d b.

The limits of Specifications 3.2.2 and 3.2.3 are maintained and detennined at the frequencie, specified i'i Specification 4.10.2.2 below.

APPLICABILITY: MODE 1.

ACTION:

With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, and 3.2.4 are suspended, either:

Reduce THERMAL POWER sufficient to satisfy the ACTION require-a.

ments of Specifications 3.2.2 and 3.2.3, or b.

Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be < 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

I 4.10.2.2 The Surveillance Requirements of Specifications 4.2.2 and 4.2.3 shall be performed at the following frequencies during PHYSICS TESTS:

a.

Specification 4.2.2 - At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.

Specification 4.2.3 - At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

RORTH ANNA - UNIT 1 3/4 10-2 Amendment No. YS.105 I

a

.o REACTIVITY CONTROL SYSTEMS BASES 3/4,l.2 B09AT10N SYSTEMS (Continued)

With the RCS average temperature above 200'F a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable.

Allowable out-of service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The boration capabioity of either system is sufficient to provide a SHUTOOWN MARGIN from expected operating conditions of 1.77% ak/k af ter xenon decay and cooljown to 200'F.

This expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 6,000 gallons of 12,950 ppm borated water from the boric acid storage tanks or 54,200 gallons of 2300 ppm borated water from the refueling water storage tank.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 324"i provides assurance l

that a mass addition pressure transient can be relieved by the operation of a single PORV.

With the RCS temperature below 200"F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

The boron capability required below 200'F is sufficient to provide a SHUTOOWN MARGIN of 1.77% ak/k after xenon decay and cooldown from 200*F to 140'F. This condition requires either 1378 gallons of 12.950 ppm borated water from the boric acid storage tanks or 3400 gallons of 2300 ppm borated water from the refueling water storage tank.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

  • I The OPERAB!t.ITY of one boron injection system during REFUELING ineures that this system is available for reactivity control while in MODE 6.

NORTH ANNA - UNIT 1 B 3/4 1-3 Amendment No. 5, 16, 6S,7J,11' l

REACTMTYCONTROLSYSTEMS EMSES T412 BORATION SYSTEMS (ContinuM)

The limits on contained water volume and boron concentration of the RWST ensure a pH value of between 7.7 and 9.0 for the solution recirculated within the containment after a LOCA.

This pH minimizes the evolution of lodine and minimizes the effect or chloride and caustic stress corrosion on mechanical systems and components.

At least one charging pump must remain operable at all times when the opposite unit is in MODE 1,2,3. or 4. This is required to maintain the charging pump cross connect system operational.

T4.11 MOVABLE CONTROL ASSEVBUES The specifications of this section (1) ensure that acceptable power distribution limits are maintaired, (2) ensure that the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalignment on associated accident analyses. OPERABILITY of the movable Lontrol assemblies is established by obrerving rod motion and determining that rods are positioned wimin i 12 steps (Indicated position) of the respecth e demand step counter position. The OPERABILITY of the individual rod position Indication system 1.s established by appropriate periodic CHANNEL CHECKS, CHANNEL FUNCTIONAL TESTS, and CHANNEL CALIBRATIONS. OPERABILITY of the individual rod position indicators is required to determine control rod position and thereby ensure compilance with the control rod alignment and insertion limits. The OPERABLE condition for the individual rod position indicators is defined as being capable of indicating rod position within : 12 steps of the associated demand position indicator.

For power levels below 50 percent of RATED THERMAL POWER, the specifications of this section permit a maximum one hour in every 24 stabilization period (thermal " soak time") to allow stabilization of known thermal drift in the individual rod position indicator channels during which time the indicated rod position may vary from demand position indicr. tion by no more than t 24 steps. This "1 in 24' feature is an upper limit on the frequency of thermal soak allowances and is available both for a continuous one hour period or one consisting of several discrete intervals. During this stabilization period, greater rellance is placed upon the demand position indicators to determine rod position. In addition, the t 24 stepfhour limit is not applicable when the control rod position is known to be greater than 12 steps from the rod group step counter demand position indication. Above 50 percent of RATED THERMAL POWER, rod motion is not expected to induce thermal transients of sufficient magnitude to exceed the individual rod position indicator instrument accuracy of i 12 steps. Comparison of the demand position Indicators to the bank Insertion limits with venfication of rod position by the individual rod position indicators (after thermal soak following rod motion below 50 percent of RATED THERMAL POWER)is sufficient venfication that the control rods are above the insertion limits.

The control bank FULLY WITHDRAWN position can be varied within the interval of 225 to 229 steps withdrawn, inclusive.

This interval permits periodic repositioning of the parked RCCAs to minimize wear, while having minimal impact on the normal reload core physics and safety evaluations.

Changes of the RCCA FULLY WITHDRAWN position within this band are administrative 1y controlled, using the rod insertion limit operator curve.

NORTH ANNA + UNIT 1 B 3/4 14 Amendment No. 16.2$,138. Fli 149

q

[(pwn.jo,,

UNITED STATES p,

NUCLE AR HEGULATORY COMMISSION I

WASHING TON, D, C. 20566

  • t,

/

.....f VIRGIN 1A ELECTRIC AND POWER COMPANY OLD 00 MIN 10N ELECTRIC COOPERAT!VE DOCKET NO. 50-339 NORTil ANNA p0WER STATION UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE l

Amendment No. 133 License No. NPF-7 i

.1.

The Nuclear Regulatory Comission (the Comission) has found thet:

A.

The application for amendment by Virginia Electric and Power Company, et a1., (the licensee) dated August 11 1989, complies with the standards and requirements of the AtomIt Energy Act of 1954, as amended (the-Act), and the Comission'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 Comission;-

C.

There is reasonable assurt.nce (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the tublic, and (ii) thet such activities will be conducted in compliance with the Comission's regulations;-

D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2 2.

Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. Npf.7 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 133, 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 issuance and shall be implemented within 30 days.

FOR THE NilCl. EAR REGULATORY COMMISSION 9

Herbert N. Berkow, Director project Directorate 11-2 Division of Ceactor Projects - 1/11 Office of Nuclete Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: November 22, 1991 l

i 1

ATTACHMENT TO l.1 CENSE At1ENDt'ENT P0. 133 TO FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50 339 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated.

The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

paae W

1-5 B 2-?

3/4 1-21 3/4 10-1 0 3/4 1-4 i

1.0 DEFINITIONS (Continued)

QQJFFRED SAFETY FEATURE RESDOtacF TIME 1.12 The ENGINEERED SAFETY FRATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting arid equence loading delays where applicable.

FREOUENCYPCTATON 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

PhlLY WITHDRAWN 1.13a The control bank FULLY WITHORAWN position shall be within the interval of 225 to 229 steps withdrawn, inclusive.

Definition of the FULLY WITHDRAWN position for each specific cycle shall be documented in the rod insertion limit operator curve, GASEOUS RADWASTE TPEATIKNT SYSTUd 1,14 A GASEOUS RADWASTE TREATMENT SYSTEM is the system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactMty prior te release to the environment. The system is composed of the waste gas decay tanks, regeneratrve heat exchanger, waste gas charcoal filters, process vent blowers, waste gas surge tanks and waste gas diaphram compressor.

IDEPR1FIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packhg leaks that are captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfero with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

[

c.

Reactor coolant system leakage through a steam gensrator to the secondary systeni.

MEMBERIS) OF THE PMUQ 1.16 MEMBER (S) OF THE PUBLIC shall include all individuals who by virtue of their occupational status have no formal association with the plant. This category shall include non.

employees of the licensee who are permitted to use portions of the site for recreational, occupational, or other purposes not associated with plant funct6ons. This category shall nat include non employees such as vending machine servicemen or postman who, as part of their formal job function, occasionalty enter an area that is controlled by the licenses for purposes of protection of Individuals from exposure to radiation and radioactive malertals.

NORTH ANPA - UNIT 2 13 Amendment No. 31,130, 133

e 1.0 DEFINITIONS (Continued)

QEES!IE DOSE CALCULATION h%NUAL 1.17 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and l parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Satpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semi annual Radioactive Effluent Release Heports required by Specifications 6.9.1.8 and 6.9.1.9.

OPERABLE OPERABILITY OPERABILITY when it is capable of performing its specified function (s), and whe 1.18 A system, subsystem, train, component or device shall be OPERABLE or have attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other cuxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related suppott function (s).

OPERATIONALMODE MODE 1,19 - An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination l of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.1, PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear l characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOLNDARY LEAKAGE 1.21 PRESSUAE BJUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) l

- through a non isolable fault in a Reactor Coolant System component body, pipe wati or vessel

wall, EPOCESSCONTROLPROGRAM i

4 1.22 The PROCESS CONTROL PMGRAM (PCP) shall contain the current formulas, sampling. l analyses, tests and determinations to oe made to ensure that the processing and packaging of solid radioactive wastes based on dersonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20,61, and 71, State regulations, burial ground requirements, and other requirements Governing the disposal of the radioactive waste.

PURGE. PURGING 1.23 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purity the confinement.

NORTH ANNA UNIT 2 14 Amendment No. H, !! A,130,

1.0 DEFIN!TIONS (Continued)

QUADRANT POWER TILT RATIO 1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper ex.

core detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greate*.

With one excore detector inoperable. the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2893 MWt.

gACTORTRIPSYSTEMRESPONSETIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channdl sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

SHUTDOWN MARGIN 1.2B SHUTOOWN MARr.!N shell be the instantaneous amount of reactivity by which the reactor u subtritical or would be subcritical from its present condition assuming all full length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be FULLY WITHDRAWN.

SITE BOUNDARY 1.29 The SITE BOUNDARY shall be t, hat line beyond which the land is not owned, leased or otherwise controlled by the licensee.

SLAVE RELAY TEST _

l. 30 A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay.

The SLAVE RELAY TEST g

shall include a continuity check, as a minimum, of associated testable actuation devices.

SOURCE CHECK 1.31 A. SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to radiation.

This applies to installed radiation monitoring systems.

l NORTH ANNA - UNIT 2 1-5 Amendment No. 37.$7,77. NP. M '

l J ih 133

1.0 DEFINITIONS (Continued)

STACCERED TEST BASIS I

1.32 A STAGGERED TEST BASIS shall consist oft a.

A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal sabintervals, b.

The testing of one system, subsystem. train or other designated component at the beginning of each subinterval.

THERMAL POVER

(

1.33 THERMAL F0VER shall be the total reactor core heat transfer rate to the reactor coolant.

UNIDENTIFIED LEAKAGE l

1.34 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAltAGE.

UNRESTRICTEL AREA 1.35 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY where access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or any area within the $1TE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/cr recreational purposes.

VENTILATION EXRAUST TREATMENT SYSTEM l

1.36 A VENTILATION EIHAUST TREATMENT SYSTEM is the system designed and installed to reduce gaseous radiciodine or radioactive material in particulate e,

form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents).

Engineered Safety Feature (EST) atmospheric cleanup systess are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING f

1.37 VENTING is the controlled process of discharging air or gas f rom a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a r.anner that replacement air or gas is not provided or required during VENTING.

Vent, used in systen names, does not imply a VINTING process.

NORTH ANNA - UNIT 2 1-6 Amendment No. g y,109

t 2.1 SAFETY LlHITS BASES 2 1.1 REACTOR CORE 2

The restrictions of this safety limit prevent overheating of the fuel and post.ible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through a correlation.

The.DND correlation l

has been developed to predict the 04D flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB, heat flux ratio. DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is as follows:

there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and Il events is greater than or equal to the DNBR limit of the DND correlation being used (the WRB-1 correlation in this applicetion). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DND will not occur when the minimum DNBR is at the DNBR limit.

In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel faL'ication parameters are considered statistically such that there is at least a 95% probability that the minimum DNBR for the limiting rod is greater than or equal to the DNBP. limit. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DMDR uncertr'9ty, combined with the correlation DNBR limit, establishes a design DNBR valm which must be met in plant safety analyses using values of input parameters without uncertainties.

As an additional criterion, meeting the DNBR limit also ensures that at least 99.9%

of the' core avoids the onset of DNB when the plant is at the DNBR limit.

The curves of Figures 2.1-1, 2.1-2, and 2.1-3 show the loci of points of THERMAL PDWER, Peactor Coolant System pressure and, average temoerature for which the minimum DNBR is no less than the design limit DNBR, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

NORTH ANNA - UNIT 2 B21 Amendment No. /J, 100

i 4

SATET! LIMITS BASES The curves are based on an enthalpy hot channel factor. FY, of 1.49 and a g

reference cosine with a peak of 1.55 for axial power she'ps.

An allowance is included for an increase in F;g at reduced power based On the expression:

g Tfg = 1.49 [l+0.3 (1-P)]

where P in the f raction of RATED THEPRAL Pok'ER These limiting heat flux conditions are higher than those calculated for the range of all control rods FULLY WITHDRAWN to the maximum allowable control rod l

insertion assuming the axial power imbalance is within the limits of the f(delta I) function of the Overteeperature trip, k'h en the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature at trips vill reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM FRESSURE The restriction of this Saft f Limit trotects the integrity of the Reactor Coolant System from overprzeauritation and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atnesphere.

The reactor pressure vessel and pressurizer are designed to cection 111 of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 peig) of design pressure.

The Reactor Coolant System piping, valves and fittings, were initially desig0ed tc ANSI B 31.1 1967 Edition and ANSI B 31.7 1969 Edition (Table 5.2.1-1 c' FSAR) unich permits a maximum transient pressura of 120% (2985 psig) of component design pressure.

The Snfety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

entire Reactor Coolant System is hydtotested at 3107 psig. 125% of design

.gsure, to demonstrate integrity prior to initial operation.

i NORT11 ANNA - UNIT 2 L 2-2 Amendment No. 8 f 77.133

_____m REACTIVITY CONTROL SYSTEMS RQOOROPTIME LIMITING CONDITION FOR OPERATION 3.1. 3. 4 The individual full length (shutdown and control) rod drop time ft'om the '229 STEP withdrawn position shall be less than or equal to ?.7 teconds from l

beginning of decay of stationary gripper coil voltage to dashpot on.ry with:

T,yg greater than or equal to 500*F and a.

b.

All reactor conlant rumps operating.

APPLICABILITY:

MODES 1 and 2.

ACTION:

With the drg? time of any full length rod determined to exceed a.

the aoove limit, restore the rod drop time to within the above limit prior to proceedit.g to MODE 1 or 2.

b.

With the rod drop times within limits but determined with 2 reactor ecolant pumps operating, operation may proceed provided THERMAL POWER is restricted to:

I.

Less than or equal to 66% of RATED THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are open, or 2.

Lass than or equal to 71% of RATFD THERMAL POWER when the reactor coolant stop valves in the nonoperating loop are closed.

.S_UR_VEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full length rods shall be demonstrated through nessurement prior to reactor criticality:

a.

For all rods following each removal of the reactor vessel head, b.

For specifically affected individual rods followir,g any main-tenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and c.

At least once per 18 months.

NORTH ANNA - UNIT 2 3/4 1-21 Amendment No. AES: 133

. =..

REACTMTYCONTROL SYSTEMS SHl1TDOWN ROD INe#RTION LIMIT UMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be lintiled in physical insertion as ;ipeelfled in the CORE OPERATING UMfTS REPORT.

1 APPLICABILITY: MODES 1* and 2'#

ACTION:

With a maximum of one shutdown rod inserted beyond the insertion limil specified in the CORE OPERATING LIMITS REPORT. except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:

a. Restore the rod to within the insertion limit specified in the CORE OPERATING UMITS REPORT, or
b. Declare the rod to be inoperable and apply Specification 3.1.3.1 1

4 SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be within the insertion limit specified in the CORE OPERATING UMrTS REPORT a.

Within 15 min;:tes prior to initial control rod bank withdrawal during an approach to reactor criticality, and

b. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

',i

'See Special Test Exceptions 3.10.2 and 3.10.3.

  1. With K,tg greater than or equal to 1.0 NORTH ANNA UNIT 2 3/4 1 22 Amendment No.121,130,

.. _. ~

e i

3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN HARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided the reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s),

i APPLICABILITY:

MODE 2.

ACTION:

a.

With any full length control rod not fully inserted and with less than the above-reactivity equivalent available for trip insertion, initiate

- ana continue boration-at greater than or equal to 10 gpm of a solution containing at least 12,950 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b.

With all full length control rods inserted and the reactor subcritical by less than_ the above reactivity equivalent, inmediately initiate and-continue boration at greater than or equal to 10 gpm of a solution containing at least 12.950 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is' restored.

SURVE!LLANCE REQUIREMENTS 4.10.1.11.The position of each control rod either partially or-FULLY WITHDRAWN shall be determined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4.10.1.2 Each control rod that is not fully inserted shall be demonstrated-capable of full insertion when tripped from at least 50% withdrawn position

= '

-within-24 hours prior to reducing the. SHUTDOWN MARGIN to less than the limits

- l of Specification 3.1.1.1.

-NORTH ANNA UNIT 2 3/4 10-1 Amendment No. 64, 133 1

-,n-

-..,,.r.,,,,,-,w.-,,

-,-,,vr

,nr,v,,

n,-,,,n,.-,-,

~ -

,n.w.-,,,-,,.-,.w,

,..., ~. - -,, - - - - <,-,, -,,. -, - - - -,., -, -, -,,, -,,, -

-~-

e i

l SPECIAL TEST EXCEPTIONS GROUP HEIGHT. INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERAr!ON i

3.10.2' The group height, insertion and power distribution limits of i

Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, and 3.2.4 may be suspended l

during the performance of PHYSICS TESTS provided:

a.

The THERMAL POWER is maintained less than or equal to 85% of RATED

-THERMAL POWER, and b.

The limits of Specifications 3.2.2 and 3.2.3 ars maintained and determined at the frequencies specified in Specification 4.10.2.2 below.

APPLICABILITY: MODE 1.

ACTION:

With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, and 3.2.4 l

are suspended, either:

a.-

Reduce THERMAL POWER sufficient-to satisfy the ACTION require-ments of Specifications 3.2.2 and 3.2.3, or b.

Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS i

4.10.2.1 - The THERMAL POWER shall be determined to be less than or equal to i

85% of RATED THERMAL POWER at least once.per hour during PHYSICS TESTS.

4.10c2.2 The Surveillance Requirements of the below listed Specifications shall be performed at'least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during PHYSICS TESTS, a.-

Specification 4.2.2.2 and 4.2.2.3.

b.

Specification 4.2.3.1 and 4.2.3.2.

l NORTH ANNA - UNIT 2 3/4 10-2 Amendment No. 64

REACT!VITY CONTROL SYSTEMS _

BASES 3/a.1.2 BORAT!0N SYSTEMS The boron injection system ensures that negative reactivity control is available during each rzde of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps,

3) separate flow paths, 4) boric acid transfer pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 200'F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN KARGlN from expected operation conditions of 1,77% delta x/k after xenon decay l

and cooldown to 200'F.

The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 6000 gallons of 12.950 ppm borated water from the boric acid storage tanks or 54,200 gallons of 2300 ppm borated water from the refuelin'g water l

storage tank.

With the RCS temperature below 200'F one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 340'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The boron capability required below 200'F is sufficient to provide a e

SHUTDOWN MARGIN of 1.77 delta k/k after xenon decay and cooldown from 200'F to 140*F. This condition requires either 1378 gallons of 12.950 ppm borated i

water from the boric acid storage tanks or 3400 gallons of 2300 ppm borated water from the refueling water storage tank.

NORTH ANNA - UNIY 2 B 3/4 1 3 Amendn ent No. SV,78

PEACTIVITY CONTROL SYSTEMS BASES 3/4.1.2 BORAT10N SYSTEMS (Continued)

The contained water volume imits include allowance for water not available because of discharge itne location and other physical characteristics.

The OPERABIL11Y of or.e boron injection system during REFUELING insures that this system is available for reactivity control while in MODE f.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.7 and 9.0 for the solution recirculated within the containment after a LOCA.

This pH minimizes the evolution of iodine and minimizes the vffect of chloride and caustic stress corrosion on mechanical systems and components.

At least one charging pump nust remain operable at all times when the opposite Unit is in MODE 1, 2, 3, or 4 This is required to maintbin the charging pump cross-connect system operational.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section (1) ensure that acceptable power distribution limits are maintained, (2) ensure that the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalignment on associated accident analyses.

OPERABILITY of the movable control assemblies is established by observing rod motion and determining that rods are positioned within + 12 steps (indicated position) of the respective demand step counter position. The OPERABILITY of the individual rod position indication system is established by appropriate periodic CHANNEL CHECKS, CHANNEL FUNCTIONAL TESTS, and CHANNEL CAL 1 BRAT 10NS. OPERABILITY of the individual rod position indicators is required to determine control rod position and thereby ensure compliance with the control red alignment and insertion limits.

The OPERABLE condition for the indivioual rod position indicators is defined as being capable of indicating rod position within + 12 steps of the associated demand position indicator.

For power levels below 50 perceiit of RATED THERHAL POWER, the specifications of this section permit a maximum one hour stabilization in every 24 period (thermal " soak tir.e") to allow stabilization of known thermal drift in the individual rod position indicator channels during which time the' indicated rod position may vary from demand position indication by no more than + 24 steps. This "1 in 24" feature is an upper limit on the frequency of thermal soak allowances and is available for both a continuous one hour period or one consisting of several discrete intervals. During this stabilization period, greater reliance is placed upon the demand position indicators to determine rod position.

In addition, the + 24 step / hour limit is i'

not applicabla when the control rod position is known to~be greater than 12 steps from i

the rod group step counter demand position indication. Above 50 percent of RATED THERMAL POWER, rod motion is not expected to induce thermal transients of sufficient magnitude to exceed the individual rod position indicator instrument accuracy of

+ 12 steps. Comparison of the demand position indicators to the bank insertion limits

- ~ith verification of rod position by the individual rod position indicators (after thermal soak following rod motion below 50 percent of RATED THERMAL POWER) is sufficient verification that the control rods are above the insertion limits.

The control bank FULLY WITHDRAWN position can be varied within the interval of 225 to 229 steps withdrawn, inclusive.

This interval permits periodic repositioning of the parked RCCAS to minimize wear, while having minimal impact on the normal reload core physics and safety evaluations.

Changes of the RCCA FULLY WITHDRAWN position within this band are administrative 1y controlled, using the rod insertion limit operator curve.

NODTH ANNA - UNIT 2 8 3/4 1-4 Amendment No. !?!,179,133

_

Retrieved from "https://kanterella.com/w/index.php?title=ML20091F479&oldid=4336913"