Text

Amend 97 to License DPR-61,supporting Operation of Plant for Cycle 15 & Reflecting Major Efforts in Upgrading design-basis Accident Analyses & in Reformatting Tech Specs as Part of Conversion to Westinghouse STS
	ML20236S373
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Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	11/12/1987
From:	Boyle M; Office of Nuclear Reactor Regulation
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+ jo,, UNITED STATES 3 NUCLEAR REGULATORY COMMISSION !

[ g i WASHINGTOW, D. C. 20666

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CONNECTICUT YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-213 HADDAM NECK PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 97 i License No. DPR-61 l J

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

A. The application for amendment by Connecticut Yankee Atomic Power ,

Company (the licensee), dated June 1, 1987, as modified by letter j dated July 22, 1987, complies with the standards and requirements !

of the Atomic 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; j C. There is reasonable assurance (i) that the activities authorized ;

by tnis amendment can be conducted without endangerir:3 the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the common i defense and security or to the health and safety of the public; and l i

E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regule.tions and all applicable requirements have ;

been satisfied. 1 1

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8711250112 g71112 3 DR ADOCK 050

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. DPR-61 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 97, are hereby incorporated into 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 E

Michael L. Boyle, Acting Director Integrated Safety Assessment Project Directorate Division of Reactor Projects III/IV/V and Special Projects

Attachment:

Changes to the Technical !

Specifications i Date of Issuance: November 12, 1987 f

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ATTACHMENT TO LICENSE AMENDMENT N0.97 FACILITY OPERATING LICENSE NO. DPR-61 DOCKET N0. 50-213 Revise Appendix A Technical Specifications by removing the pages identified .

below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

REMOVE INSERT l 1-3 1-3 1-6 1-6 2-1 2-1 Figure 2.2-1 Figure 2.2-1 Figure 2.2-2 Figure 2.2-2 2-2 2-2 2-3 2 Blank 2 -S 2-5 2-7 2-7 2-8 2-8 3-3 3-3 3-4 3-4 3-4a 3-4a 3-4b 3-4b 3-4c 3-4c 3-4d 3-4d

- 3-4e <

- 3-4f

- 3-4g

- 3-4h

- 3-4i j

- 3-43 l

- 3-4k I

- 3-41

- Figure 3.3-1

- 3-4m

- 3-4n

- 3-40

- 3-4p

- 3-4q 1

- 3-4r '

- 3-4s

- 3-4t

_ _ - - - - _ l

REMOVE INSERT 3-8 3-8 3-9 3-9 3-11 3-11 3-15a 3-15a 3-16 3-16 3-17 3-17 3-17a 3-17a 3-17b 3-17b 3-17c 3-17c 3-17d 3-17d 3-17e 3-17e

- 3-17f 3-179

- 3-17h

- 3-171

- 3-17,1

- 3-17k

- 3-171

- 3-17m

- 3-17n

- 3-17o Figure 3.10-1 Figure 3.10-1

- 3-17p

- Figure 3.10-2

- 3-17q

- 3-17r

- 3-17s 3-18 3-18 3-19 3-19 3-23 3-23 3-24 3-24 3-27 3 Blank 3-28 3 Blank 3-30 3-30

- Figure 3.17-1A

- Figure 3.17-1B 3-31 3-31

- Figure 3.17-2A

- Figure 3.17-2B 3-31a 3-31a

- 3-31b

- 3-31c

- 3-31d Figure 3.17-1 Figure 3.17-3

l REMOVE INSERT 3-31e 3-31f 3-31g 3-31h '

.l 3-311 3-31j i 3-31k l 3-311 !

Table 3.17-1 l 3-31m 3-31n 3-31o

\ 3-32 3-33 3-34 3 Blank 3 Blank 3 Blank 3-34a -

3-35 3 Blank 3-36 3 Blank Figure 3.18-la -

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Figure 3.18-1c -

Figure 3.18-2a -

Figure 3.18-2b -

Figure 3.18-2c -

3-39 3 Blank 3-40 3 Blank 3-47 3-48 3-49 3-50 4-18 4-18 T

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s a _ - - _ _ _ _

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DEFINITIONS CHANNEL FUNCTIONAL TEST l

1.11 A CHANNEL FUNCTIONAL TEST shall.be the injection of a simulated signal into the channel as close to the primary sensor as prac-ticable to verify OPERABILITY including alarm and/or trip functions.

CORE ALTERATION 1.12 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

SHUTDOWN MARGIN 1.13 SHUTDOWN MARGIN shall be the instantaneous. amount of reactivity by which the reactor is suberitical or would be suberitical from its present condition assuming all rod cluster assemblies are fully inserted except for the single rod cluster of highest reactivity worth which is assumed to be fully withdrawn.

IDENTIFIED LEAKAGE 1.2,4 IDENTIFIED LEAKAGE shall be:

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

b. Leakage into the containment atmospbere from sources that are both specifically located and known either not to interfere with the operation of UNIDENTIFIED LEAKAGE monitoring systems or not to be PRESSURE BOUNDARY LEAKAGE.

UNIDENTIFIED LEAKAGE' I.15 UNIDENTIFIED LEAKAGE shall be all leakage which has not been identified.

1-3 Amendment No. 97

TABLE 1.1 OPERATIONAL MODES REACTIVITY % RATED AVERAGE COOLANT MODE CONDITION, Keff THERMAL P0k'ER* TEMPERATURE

1. P0k'ER OPERATION >0.99

>5% >350*F l l

2. < >

STARTUP _

.99 _5% _350'F g

3: HOT STANDBY < .99 0 >350 F

4. HOT SHUTD0kW < .99 0 350'F > T,yg > 200'T

5. COLD SHUTDok3 < .99 0 $200*F

6. PITUELING** 5 94 0 <140'T

Excluding decay heat.

- Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

1-6 Amendment No. 29, 97

SECTION 2 .

SAFETY LIMITS AND MAXIMUM SAFETY SETTINGS ;

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2.1 INTRODUCTION

I Safety Limit's are defined in order to protect the fuel cladding and I the reactor coolant system. The integrity of these barriers must be maintained to prevent an uncontrolled release of radioactivity.

Maximum Safety Settings are also established for protective devices related to the process variables on which the safety limits are based. The maximum safety settings are chosen,such that protective action will prevent the safety limits from being exceeded.

2.2 SAFETY I.IMITS REACTOR CCRE 2.2.I. The combination of THERMAL POWER, pressurizer pressure, l and the highest operating loop inlet temperature (T } '

shall not exceed the limits shown in Figures 2.2-I INd 2,2-2 for four and three loop operation, respectively.

APPLICABILITY: MODES I and 2 ACTION:

Whenever the point define'd by the combination of the highest operating loop inlet temperature (T ) and pressurizerpressurehasexceededtheapprohEh!tepercent of rated thermal power (RTP) line, be in HOT STANDBY within I bour, and comply with the requirements of Speci-fication 6.7.1.

2-1 Amendment No. 97

610 -

600 90 IE ~

100:

580 1105

, 570 -

118 E 560 -

1251 g

x

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y 550 -

540 530 520

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$10 1700 1800 1900 2000 2100 2200 2300 1600 PRES $Utt. PSIA FIGURE 2.21 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION Anendnent No. 3, 97 e * ,

53c -

. 6 60s 620 -

W.

610 - .

605 l

65:

600 W.

$90 00:

580 90s 570 -

u0 -

f550 540 530 52c -

l a a 1 i i 510 m i 1600 1700 1800 1930 2000 2100 2200 2300 PRI$$URE. P51A FIGURE 2.2 2 REACTOR CORE SAFETY LIMIT - THREE LOOPS IN OPERA 11oN, l Amndent No. $1, 97

BASES 2.2.1 REACTOR COLE I

The restrictions of this Safety Limit prevent overheating 1

of the fuel and possible cladding perforation which would result in the relcase 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 esultant sharp reduction in beat transfer coefficient. DhB is not a directly measurable parameter during operation and, therefore, THER".AL P0kT.R and Reactor Coolanc Temperature and Pressure have been related to DNB th' rough the W-3 correlation. The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uaiform and nonuniform heat flux distri-butions. The local DNB heat flux ratio (DNBR) is defined '

as the ratio of the heat flux that would cause DNB at a psrticular core location to the local heat flux, and'is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is cbosen as an appropriate margin to DNB for all operating conditions.

The curves of Tigures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, pressurizer pressure and core inlet temperature for which the minimum DNBR is no less than 1.30, and the core outlet void fraction is no greater than 0.32.

These curvgs are based on nuclear enthalpy hot channel o

factors,F"No,n,f1.60and1.64forfourloopandthree loop opera respgetively. An allowance is included for an increase in T"g at reduced power.

These limiting hot channel factors are higher than those calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod l insertion. This insertion limit is described in The required reduction Specification 3.10.2.6 and 3.10.2.7.

in power level as dictated by Tigures 3.10-1 and 3.10-2 insures that the DNB ratio is always greater than 1.30, 2.- 2 Amendment No. 97

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Intentionally Lef t Blank l

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l Amendment No. 97 2-3

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l 2.4 MAXIMUM SATETY SETTINGS - PROTECTIVE INSTRUMENTATION Applicability: Applies to trip settings for instruments monitoring reactor power and reactor coolant pressure, temperature, and flow. -

Objective: To provide for protective action in the event that the principle process variables approach a safety limit.

Specification: Protective instrumentation trip settings shall be as follows:

Tour Reactor Coolant Three Reactor Coolant Pumps enerating Pumps Operating (1) Pressurizer Prer,sure ff.300 psic $2300 psig l

(2) Pressurizer Level * [86% of range 186% of range (3) Variable Low Pressure *** 117.4 >17.4

-8850 (T""8+1.17AT)

-8850(T *8 + 1.17AT)

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(4) Noelear Overpower ** 1109% of rated power 174% of rated power

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(5) Low Coolant Flow *** 190% of nominal four 184% of nominal loop flow three loop flow (6) Reactor Coolant Loop $20'T $20*F Valve-Temperature Interlock (7) High Steam Flow 110% of full load 110% of full load steam flow steam flow (8) High Startup Rate **** 15 decades per minute 15 decades per minute

May be bypassed when the reactor is at least 1.5%Ak suberitical.

- The nuclear overpower trip is based upon a symmetrical core power distribution. When the reactor power is (10% the overpower trip )

setpoint is reduced to 25 percent of rated power i

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- eMay be bypassed below 10 percent of rated power.

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- - - Not required if the reactor trip breakers are cpen or the control rod drive lift coils are de-energized; say be bypassed above 10 percent of rated power.

Basis: The reactor protective system is designed and constructed such that no single failure in any of the instrument systems will prevent the desired safety actiou if an applicable parameter exceeds a r.afety setpoint.

1 Amendaent No. 2), /f, BA, 97 ;

2-5 i

and shutdown. It is safe to block this trip below 10 percent power since the protection afforded by this trip is not required at this low level. Removal of unnecessary trip signals will reduce the number of spurious trips.

(4) Nuclear Overpower As explained above, the nuclear overpower ,

reactor trip, in conjunction with the variable low pressure reactor trip, provides overpower, overtemperature protection. The nuclear overpower trip channels will respond first to rapid reactivity insertion j rates, detected by the increase in flux, before there are any significant changes in the system trocess variables. A maximum error of 9 percent of full power due to setpoint, instrumentation, and calorimetric determination (see Section 4.3.6 of the FDSA) is considered in establishing the setpoint. In order to reduce the time to trip for certain accidents occurring at low s power, the overpower setpoint is lowered to 25 percent when reactor power is below 10 percent. This low overpower trip would .

terminate the postulated large steamline break accident from the hot zero power condition. The lower setting for three loop operation provides protection at the reduced power level equivalent to that provided by the setting for four loop operation at full power.

(5) Low Coolant Flow The low coolant flow reactor trip protects the core against an increase in coolant temperature resulting from a reduction in

, coolantflowwhily3 e reactor is at substantial power This trip will !

prevent DNB in any loss-of-flow incident, which eliminates the possibility of clad damage. Flow detection in each reactor coolant loop is from a measurement of pressure drop from inlet to outlet of each steam generator. The 90 percent and 84 percent low flow signals are high enough to activate a trip in time to prevent DNB, and low enough to reflect that a loss-of-flow j i

condition truly exists. A maximum instrument and setpoint error of 5 percent full flow is considered in determining the setpoint.

Loss-of-flow protection is also provided by i reactor coolant pump breaker and from undervoltage 2-7 k n k nt No. U , H , W

on a reactor coolant pump motor bus.

This trip signal may be defeated below 10 percent power since at this level, natural circulation, if need be, could cool the core. Spurious trips can be eliminated by defeating the signal.

(6) Reactor Coolant Loop Valve -- Temperature Interlock The reactor coolant loop valve-temperature interlock prevents the return.to service of an isolated loop, whose temperature is substantially below the highest co leg temperatureoftheoperatingloops{gy The .

setting of 20 F (plus 10 F instrument and set point error) limits the positive reactivity insertion that could occur due to admission of cooler water to the reactor coolant system to a value below that which would result in DNB.

(7) High Steam Flow This circuit providesprgggetionagainsta large steam line rupture . This signal I closes the main steam line isolation valves and. trips the reactor, thereby limiting the cooldown. An error of 5 percent full steam flow is used in determining the trip settings.

l l (8) High Startup Rate The Intermediate Range High Startup Rate trip provides core protection during l reactor startup. This trip function provides protection for large reactivity insertion events initiated from a suberitical mode of operation. This trip function is credited in the rod withdrawal from suberitical analysis.

J

References:

(1) FDSA - Section 7.2 {

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(2) FDSA - Section 10.3.5 J

(3) FDSA - Section 10.3.2 i (4) FDSA - Section 10.2.2 1

(5) FDSA - Section 10.3.3 2-8 Amendment No. 97

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3.3 REACTOR COOLANT SYSTEM l 3.3.1.1 REACTOR COOLAt1T LOOPS AND COOLANT CIRCULATION

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STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION i

3.3.1.1 The following reactor coolant loops sha!! be in operation, with the associated loop stop valves OPERABLE:

a. All reactor coolant loops in operation with the reactor above 65%

of RATED THERMAL POWER, or -

b. At least three coolant loops in operation

with the reactor less than or equal to 65% of RATED THERMAL POWER.

APPLICABILITY: MODES I and 2.

s ACTION:

I With less than the above recuired reactor coolant loops in operation or the associated loop stop valves not OPERABLE be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . !

i SURVEILLANCE REQUIREMENT

1. At least once pu 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the above required reactor coolant loops shall i

' be verified to be in operation and circulating reactor coolant, and that ;

power is available to the loop stop valves. !

2. At least once per 18 months, cycle the loop stop valves through one l' complete cycle of full travel.

The loop out of service may be idled (cold leg stop valve closed) or isolated (cold and hot leg stop valves closed).

Amendment No. 20, 33, f 2, 97 3-3

REACTOR COOLANT SYSTEM ,

l HOT STANDBY LIMITING CONDITION FOR OPERATION 3,3.1.2 The following number of Reactor Coolant Loops listed below shall be OPERABLE and in operation.* l

a. At least three reactor coolant loops shall be OPERABLE and at least two reactor coolant lonps shall be in operation if the reactor trip breakers are closed and the control rod drive lif t coils energized, or

b. At least two reactor coolant loops shall be OPERABLE and at least one reactor coolant loop shall be in operation if the reactor trip breakers are open or the control rod drive lift coils are de-energized.

The reactor coolant loops are;

a. Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump, ,

b. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,

c. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump, and

d. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump.

APPLICABILITY: MODE 3.

ACTION: a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. With only one reactor coolant loop in operation and the reactor trip system breakers closed and the control rod drive lif t coils energized, within I hour either open the reactor trip system breakers or de-energize the control rod drive lif t coils,

c. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System,immediately open or verify open the reactor trip system breakers, and initiate corrective action to return the required loop to operation. i

All reactor coolant pumps may be de-energized for up to I hour provided:(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 100F below saturation temperature.

3-4 Amendment No f2, SA, 97

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REACTOR COOLANT 5YSTEM HOTSTANDBY SURVEILL ANCE REQUIREMENT 5

a. At least once per 7 days the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE.

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the required steam generators shall be j

determined OPERABLE.

c. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the required reactor coolant loops shall be verified in operation and circulating reactor coolant.

d. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,if required, verify that the reactor trip .

system breakers are open or the control rod drive lift coils are de-l energized, s

3-4a Amend, tent No. 20, 33, A$, /J, 97 l

i RE ACTOR COOLANT SYSTEM l HOTSHUTDOWN MMITING CONDITION FOR OPER ATION 3.3.1.3 The following number of heat removalloops listed below shall be OPERABLE

and in operation:

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a. At least three reactor coolant loops shall be OPERABLE and at least two reactor coolant loops shall be in operation if the reactor trip breakers are closed and the control rod drive lif t coils energized, or

b. At least two heat removal loops (RCS or RHR)' shall be OPERABLE and at least one heat removal loop (RCS or RHR) shall be in operation if the reactor trip breakers are open or the control rod drive lif t cells are de-energized.

'l The heat removal loops are;

a. Reactbr Coolant Loop 1 and its associated steam generator and reactor coolant pump, :

b. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,"'

c. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump,

d. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump,

e. RHR loop A, and

f. RHR loop B.

APPLICABILITY: MODE 4 ACTION: a. With less than the above required loops OPER ABLE, immediately initiate corrective action to return the required loops to OPERABLE status and open or verify open the reactor trip system breakers within I hour, if the remaining OPERABLE loop is an RHR loop, be in COLD SHUTDOWN

'. within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

! With no loop in operation, suspend all operations involving a reduction in

b.

boron concentration of the Reactor Coolant System,immediately open i or verify open the reactor trip system breakers and initiate corrective action to return the required loop to operation.

All reactor coolant pumps and RHR pumps may be deenergized for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant Systern boron concentration, and (2) core outlet temperature is maintained at least 100F below saturation temperature.

An RHR loop consists of a Residual Heat Removal (RHR) pump, a dedicated RHR heat exchanger either from the same train or from the opposite train and all other necessary piping and components required to receive and cool reactor coolant.

* % reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 31)oF unless the secondary water temperature of each steam generator is less than 200F above each of the Reactor Coolant System cold leg temperatures.

Amendment No. 33, 3$, 4 , 97

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REACTOR COOL ANT SYSTE\1 SURVEILL ANCE REQUIREMENTS O At least once per 7 days the required reactor coolant pump (s),if not in i operation, shall be determined OPERABLE.

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b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the required steam generator (s) shall be determined OPERABLE by verifying secondary side narrow range water level to be greater than or equal to 25%.

c. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months one reactor coolant or RHR loop shall be verified in operation and circulating reactor coolant.

d. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,if required, verify that the reactor trip system breakers are open or the control rod drive lif t coils are de-energized.

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l 3-4c Amendment No. 33, 3$ M. M, $$, 97 !

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{

REACTOR COOLANT SYSTEM l i

COLD SHUTDOWN - LOOPS FILLED ,

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{

LIMITING CONDITION FOR OPERATION 1

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3.3.1.4.1 At least one RHR loop shall be OPERABLE and in operation,* the reactor trip system breakers shall be open or the control rod drive lift coils shall be de-energized, and either: i i

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a. One additional RHR loep shall be OPERABLE, or

b. The secondary side narrow range water level of at least two unisolated steam generators shall be greater than 25%

APPLIC ABILIT Y: MODE 5 with reactor coolant loops filled.*

ACTION:

a. With one of the RHR loops inoperable and with less than the required steam generator level, immediately initiate corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator water level as soon as possible. l

b. With no RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

c. With the reactor trip system breakers closed and the control rod drive lift coils energized, within I hour either open the reactor trip system breakers or de-energize the control rod drive lif t coils.

SURVEILLANCE REQUIREMENTS

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the secondary side water level of at least two steam generators when required shall be determined to be within limits,

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , an RHR loop shall be determined to be in operation and circulating reactor coolant.

c. The RHR loop not in operation but required shall be determined OPERABLE at least once per 7 days by verifying breaker alignments and indicated power availability.

d. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that the reactor trip system breakers are open or the control rod drive lif t coils are de-energized.

The RHR pump may be deenergized for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 100F below saturation temperature.

A reactor coolant pump in an unisolated loop shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 3150F unless the secondary water temperature of each steam generator is less than 200F above each of the Reactor Coolant System cold leg temperatures.

Amendment No. M, fJ , f t, 97 3-4d

REACTOR COOLANT SYSTEM

' COLD SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.3.1.4.2 Two RHR loops shall be OPERABLE

with at least one RHR loop in operation ** and either the reactor trip system breakers shall be open or the control rod drive lif t coils shall be de-energized.

APPLICABILITY: MODE 5 with reactor coolant loops not filled.

A_CTION:

a. With less than the above required RHR foops OPERABLE,immediately i initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.

b'. With no RHR loop in operation, suspend all operations involving a i reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop-to operation.

c. With the reactor trip system breakers. closed and the control rod drive

! lift coils energized, within I hour either open the reactor trip system breakers or de-energize the control rod drive lif t coils.

l SURVEILLANCE REQUIREMENTS i

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a RHR loop shall be determined to be in operation and circulating reactor coolant.

b. At least once per 7 days the RHR loop not in operation shall be determined OPERABLE by verifying breaker alignments and indicated power availability.

c. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months verify that either the reactor trip system breakers are open or the control rod drive lif t coils are de-energized.

One RHR loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other RHR loop is OPERABLE and in operation.

- The RHR pump may be de-energ: zed for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 100F below j saturation temperature. {

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I Amendment No. 97 3-4e

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REACTOR COOLANT SYSTEM' ,

i ISOLATED LOOP I

J LIMITING CONDITION FOR OPERATION l l 3.3.1.5 The RCS loop stop valves of an isolated loop

sha!! be shut and either: . !

f . a. The power removed from the valve operators, or

b. The boron concentration of the isolated loop shall be maintained greater than or equal to the boron concentration of the operating

loops.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

With the requirements of the above specification not satisfied, either:

a. Retnove power from the valve operators within one hour, or-

b. Increase the boron concentration of the isolated loop to within the limits hithin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or

c. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS

a. At east l once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,if required, verify that power is removed fron, the valve operators.

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,if required, verify that the boron concentration of an isolated loop is greater than or equal to the boron concentration of the operating loops.

A loop is considered to be isolated when the hot and cold leg stop valves are both closed.

Amendment No. 97 3-4f

REACTOR COOLANT SYSTEM i 50 LATED LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.3.1.6 A reactor coolant loop shall remain isolated untilt

a. The temperature at the cold leg of the isolated loop is within 200F of the highest cold leg temperature of the operating loops,

b. The boron concentration of the isolated loop is greater than or equal to the boron concentration of the operating loops,

c. The reactor is subcritical by at least 1000 pcm.

APPLICABILITY: MODES 3,4,5 and 6.

ACTION:

With the requirements of the above specification not satisfied, do not open the Isolated loop stop valves.

SURVEILL ANCE REQUIREMENTS

a. The isolated loop cold leg temperature shall be determined to be within 200F of the highest cold leg temperature of the operating loops within 30 minutes prior to opening the cold leg stop valve.

b.- The reactor shall be determined to be subcritical by at least 1000 pcm within 30 minutes prior to opening the cold leg stop valve.

c. Within 30 minutes prior to opening the loop stop valves, the isolated loop shall be determined to have a boron concentration greater than or equal to the boron concentration of the operating loops.

i Amendment No. 97 i i 3-4g I i

_ REACTOR COOLANT SYSTEM IDLED LOOP _

LIMITING CONDITION FOR OPERATION The cold leg loop stop valve of an idled loop (s)* shall be shut and either:

3.3.1.7

a. The power removed from the valve operator, or

b. The boron concentration of the idled loop (s) shall be maintained greater than or equal to the boron concentration of the operatin6 loops.

APPLICABILITY: MODES 1, 2, 3, 4, and 5, ACTION:

With the requirements of the above specification not satisfied, either:

s

a. Remove power from the valve operator within one hour,

b. Increase the boren concentration of the idled loop (s) to within the limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or

c. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDO within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,if required, verify that power is removed from the valve operators,

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , if required, verify that the boron concentration of an idled loop (s)is greater than or equal to the boron concentration of the operating loop (s).

A loop is considered to be idled when the hot leg stop valve is open and the cold leg stop valve is closed. l l

Amendment No. 97 3-4h

REACTOR COOLANT SYSTEM IDLED LOOP STARTUP LIMITING CONDITION FOR OPERATION 3.3.1.8 A reactor coolant loop shall remain idled until:

a. - The temperature at the cold leg of the idled loop (s) is within 200F of the highest cold leg temperature of the operating loop (s),

b. The boron concentration of the idled loop (s) is greater than or equal to the boron concentration of the operating loop (s),

c. The reactor is no greater than 60% RATED THERMAL POWER if only one loop is idled or is subcritical by at least 1000 pcm if more than one loop is idled.

APPLICABILITY: All MODES.

ACTION:

With the requirements of the above specification not satisfied, do not open the idled loop stop valve.

SURVEILLANCE REQUIREMENTS

a. The idled loop cold leg temperature shall be determined to be within 200F of the highest cold leg temperature of the operating loops within 30 minutes prior to opening the idled loop stop valve.

b. Within 30 minutes prior to opening the idled loop stop valve, the reactor shall be determined to be either:

a. Less than 60% RATED THERMAL POWER if only one loop is idled, or

b. Subcritical by at least 1000 pcm if more than one loop is idled.

c. Within 30 minutes prior to opening the loop stop valve, the idled loop shall be determined to have a boron concentration greater than or equal to the boren concentration of the operating loops.

Amendment No. 97 3-41

_ ___.________________j

REACTOR COOLANT SYSTEM 3.3.2 SAFETY VALVES .

SHUTDOWN LIMITING CONDITION FOR OPERATION l

3.3.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with the lif t setting associated with the OPERABLE Code safety valve l

l within gl% of its design Setpoint*.

APPLICABILITY: MODE 4 except when Specification 3.3.4.2 is applicable.

ACTION:

With no pressurizer Code safety valve OPERABLE,immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

s SURVEILL ANCE REQUIREMENTS No additional Surveillance Requirements other than those required by Specification 4.2.

The lif t setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

3-4j Amendment No. 97

7 REACTOR COOLANT SYSTEM OPERATING l -

LIMITING CONDITION FOR OPERATlON .

l 3.3.7.2 . All pressurizer Code safety valves shall be OPERABLE with respective l lif t settings of 2485 psig g 1%,2535 psig g 1% and 2585 psig g 1% in accordance with their respective nameplates.*

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With one pressurizer Code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS No additional Sur eillance Requirements other than those required by Specification 4.2.

The lif t setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

3-4k Amendment No. 97

l REACTOR COOLANT SYSTEM 3.3.3 PRESSURIZER LIMITING CONDITION FOR OPERATION l

3.3.3 The pressurizer shall be OPERABLE with: !

a. Water level within 5% of the programmed level of Figure 3.3-1 during periods when THERMAL POWER is maintained constant,*

and

b. At least two groups of pressurizer heaters capable of being powered from an emergency power source and each having a -

capacity of at least 150 kW. ,

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With only one group of pressurizer heaters OPERABLE, restore at least I two groups to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT I STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the i following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . I

b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the Reactor Trip System breakers open within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS I

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the pressurizer water level shall be J determined to be within its limit.

i

b. At least once per 92 days the capacity of each of the above required l groups of pressurizer heaters shall be verified. l l

i During periods when THERMAL POWER is being changed the pressurizer water level may be outside the 5% band for periods not to exceed I hour.

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i Amendment No. 97 3-4 L

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. PRESSURIZER PROGRAMMED WATER LEVEL Amendment No. 97 I

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REACTOR COOLANT SYSTEM 3.3.4 RELIEF VALVES LIMITING CONDITION FOR OPERATION l 3.3.4.1 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE. The Setpoint for the PORY's and their block ;

valves sha!! be greater than or equal to 2325 psig and less than or equal to 23% psig. The emergency control air supply shall have a minimum pressure of 118 psig.

APPLICABILITY: MODES 1, 2, and 3. ;

I ACTION:

a. With one or more PORV(s) inoperable, because of excessive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close/ verify close the associated block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one PORV inoperable due to causes other than excessive seat leakage or low emergency control air supply pressure, within I hour l either restore the PORV to OPERABLE status or close/ verify close the 1

associated block valve and remove power from the block valse; restore the PORY to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With both PORV(s) inoperable due to causes other than excessive seat leakage or low emergency control air supply pressure, within I hour either restore each of the PORV(s) to OPERABLE status or close/ verify close their associated block valve (s) and remove power from the block valve (s) and be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the fo!!owing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

d. With the emergency control air supply pressure for the PORVs less than i 118 psig restore the emergency control air supply to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS

1. In addition to the requirements of Specification 4.2, each PORV shall be demonstrated OPERABLE at least once per 18 months by:

l a. Performance of a CHANNEL FUNCTIONAL TEST, and

b. Operating the valve through one complete cycle of full travel.

1 l

i l

Amendment No. 97 3-4m

.9 Ri1CTOR COOLANT SYSTEM g.'.igEILL ANCE REQUIREMENTS 2, Each PORY block valve shall be demonstrated OPERABLE at least once per 18 months by performance of a CHANNEL FUNCTIONAL TEST.

3. Each PORY block valve shall be demonstrated Or)ERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.3.4.1.

4. The control air supply for the PORVs shall be demonstrated OPERABLE at least once per 18 months by verifying that the control air supply does not drop more than 0.3 psiin one hour when isolated from Containment Air Supply System.

5. The air supply for the PORVs shall be verified by local pressure indication once every 92 days.

I Anm.Snent No. 97 3-4n

REACTOR COOLANT SYSTEM ,. j LOW TEMPERATURE OVERPRESSURE PROTECTION SYSTEMS -

LIMITING CONDITION FOR OPERATION i

3.3.4.2 At least one of the following Low Temperature Overpressure Protection (LTOP) Systems shall be OPERABLE:

a. Two [. TOP spring-loaded relief valves (SLRV) with a lif t setting of 380 psig (13%) with their respective motor-operated isolation valves in the open position, or

b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 7 square inches (3 inches nominal 1

. diameter).

APPLICABILITY:

MODE 4 when the temperature of any RCS cold leg of .n unisolated loop is less than or equal to 3150F. MODE 5 and MODE 6 witt, the teactor vessel head on. The above overpressure protection system shall be placed in service prior to placing the i RHR system into service and shall remain in service unless the requirements of Specification 3.3.4.2.b are satisfied.

ACTION:

i

a. With one LTOP SLRV inoperable, restore the inoperable LTOP SLRV to OPERABLE status within 7 days, or depressuriz,e and i

vent the RCS through at least a 7 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. With both LTOP SLRVs inoperable, depressurize and vent the RCS through at least a 7 square inch vent within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c. In the event either the LTOP SLRVs or the RCS vent (s) of Specification 3.3.4.2.b are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the LTOP SLRVs or the above RCS vent (s) on the I transient, and any corrective action necessary to prevent recurrence.

4 3-4o Amendment No. 97

REACTOR COOLANT SYSTEM ,

SURVEILLANCE REQUIREMENTS

a. Each LTOP SLRV shall be demonstrated OPERABLE by verifying the SLRV isolation valve is open at least once per 31 days when the LTOP SLRV is required to be OPERABLE.

b. The RCS vent (s) of Specification 3.3.4.2.b shall be verified to be open at least once per 31 days when the vent (s) are being used for overpressure protection.

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Amendment No. 97 3-4p

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l 3.3 REACTOR COOLANT SYSTEM j i

BASES 3.3.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and anticipated transients. In MODES I and 2 with one reactor coolant loop not in operation this specification requires that the plant remain below 65% power. With less than the required reactor coolant loops in operation, the plant shall be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The loop isolation valves are required to be OPERABLE in the operating loops in order to terminate the primary to secondary leak path in the event of a steam generator tube rupture.

In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident. Single failure considerations require that three loops be OPERABLE. A single reactor coolant loop provides sufficient heat removal capability for decay heat if a bank withdrawal accident can be prevented (i.e., by opening the reactor trip system breakers or de-energizing the control rod drive lif t coils). Single failure considerations require that two loops be OPERABLE.

In MODE 4, two reactor coolant loops provide sufficient heat removal capability for removing decay heat even in the event of a bank withdrawal accident. Single failure considerations require that three loops be OPERABLE. A single reactor coolant or RHR loop provides suf ficient heat removal capability for decay heat if a bank withdrawal accident can be prevented, i.e., by opening the reactor trip system breakers or de-energizing the control rod drive lif t coils. Single failure considerations require that two loops be OPERABLE.

In MODE 5 with reactor coolant loops filled, a single RHR loop provides sufficient heat :

removal capability for removing decay heat. A bank withdrawr,1 accident is prevented 1 by opening the reactor trip system breakers or de-energizing the control rod drive lif t coils. Single failure considerations require that at least two RHR loops be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat. A bank withdrawal accident is prevented by opening the reactor trip system breakers or de-energizing the control rod drive lif t coils. Single failure considerations and the unavailability of the steam generators as a heat removing component require that at least two RHR loops be OPERABLE.

l i

The operation of one Reactor Coolant Pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes j during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of l operator recognition and control. i j l l

1 Amendment No. 97 3-4q

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1 3.3.1 REACTOR COOLANT SYSTEM BASES l

i 3.3.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION (continued) !

The restrictions on starting an RCP with one or more RCS cold legs less than or equal to 3150F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 30. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the i

secondary water temperature of each steam generator is less than 200F above each of !

the RCS cold leg temperatures. !

}

The requirement to maintain the boron concentration of an isolated / idled loop greater I than or equal to the boron concentration of the operating loops ensdres that no reactivity addition to the core could occur during startup of an isolated / idled loop. j Verification of the boron concentration in an isolated / idled loop prior to opening the stop valves provides a reassurance of the adequacy of the boron concentration in the l isolated / idled loop. 1 i

Startup of an isolated / idled loop could inject cool water from the loop into the core.

The reactivity transient resulting from this cool water injection is minimized by prohibiting isolated / idled loop startup until its temperature is within 200F of the operating loops. Making the reactor subcritical prior to isolated loop startup prevents any power spike which could otherwise result from this cool water-induced reactivity transient.

3.3.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 293,300 lbs. per hour of saturated steam at 2485 psig. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, {'

connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip until the first Reactor Trip System Trip Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss-of-load) and also assuming no operation of the power-operated relief valves or steam dump valves.

Demonstration of the safety valves'lif t settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler ;

and Pressure Code.

Amendment No. 97 3-4r i l

l

l REACTOR COOLANT SYSTEM .

BASES 3.3.3 PRESSURIZER, The limit on the water level in the pressurizer assures that the parameter is maintained within that assumed in the safety analyses. The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation.

- I 1

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f Amendment No. 97 3-4s

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REACTOR COOLANT SYSTEM BASES 3.3.4 RELIEF VALVES Operation of the power-operated relief valves (PORVs) minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves and provide an alternate means of core cooling. Each PORY has a remotely operated block valve to provide a positive shutoff capability should a PORV become inoperable. One of two redundant PORY relief trains must be OPERABLE to adequately cool the core in the event that the steam generators are not available to remove core decay heat. When in automatic mode, all PORVs and block valves open automatically on high pressurizer pressure. The PORVs and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.

The OPER' ABILITY of two spring-loaded rel'ief valves (SLRVs) or an RCS vent opening of greater than 7 square' inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 3150F. Either SLRV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either:(4 the start of an idle RCP with the secondary water temperature of .

the steam generator less than or equal to 200F above the RCS cold leg temperatures, or (2) the start of a char'ging pump (centrifugal) and its injection into a water-solid RCS.

The Maximum Allowed SLRV Setpoint for the Low Temperature Overpressure Protec-tion System (OPS)is derived by analysis which models the performance of the OPS l

l assuming various mass input and heat input tran'sients. Operation with a SLRV Setpoint less than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for a maximum pressure overshoot beyond the SLRV Setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. To ensure that mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of all but one charging pump (centrifugal or metering) while in MODES 4,5, and 6 with the reactor vessel head installed and disa!!ow start of an RCP if secondary temperature is more than 200F above RCS cold leg temperature.

Amendment No. 97

3.5 CHEMICAL AND VOLUME CONTROL SYSTEM 4

Applicability: Applies to the operational status of the chemical and volume control system.

Obj ective: To specify those limiting conditions for operation of the chemical and volume control system which must be met in order to ensure safe reactor operation.

Specification: A. The reactor shall not be critical unless the following chemical and volume control

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system conditions are at-t:

(1) Either two charging pumps.or the metering pump and one charging pump operable.

(2) Two borie acid pumps or one boric acid pump and gravity fill line to metering pump operable.

(3) The boric acid tank sball contain at least 12,000 gallons of solution whose concentration shall be at least 8 percent boric acid, but not greater t' 13 percent boric acid. The to ,serature shall be 140*F or higher.

(4) Maintenance, which requires draining of the boric acid mix tank, shall be allowed only when the plant is shut down and the reactor coolant system l borated to the refueling boron concen- {

tration. J (5) System piping and valves operable to the extent required to establish two i flow paths for boric acid injection to the reactor coolant system.

(6) Valve BA-V-399 shall be locked open and shall not be closed except when the plant is shutdown and the reactor coolant system borated to the refueling boron concentration.

B. A maximum of one centrifugal charging pump shall be operable whenever the temperature

! of one or more nonisolated RCS cold legs is I less than or equal to 315'T and the RCS is not vented by a minimum opening of a 3-inch diameter.

h l Amendment No. 22, 33, 94, 97 3-8

Basis: The chemical and volume control system provides control of the reactor coolant systed boron inventpry.

Either a 360 gpm charging pump or a 30 gpm metering pump is capable of injecting concentrated boric acid l !

j directly into the coolant system. Approximately 10,000 gallons of the 8 percent solution boric acid i 1

are required to meet cold shutdown requirements.

Thus, a minimum of 12,000 gallons in the mix tank is specified. An upper concentration limit of 13 percent boric acid in the mix tank is specified to maintain solution solubility at the specified low temperature limit of 140*F. Limits on draining the tank are specified to afford opportunity for scheduled main-tenance. The refueling boron concentration is specified before boric acid mix tank maintenance is undertaken to preclude a return to criticality under any circumstances, even fuel movement.

When boric acid blender and concentrated boric acid (20,000 ppm boron) are used to change the reactor coolant system boron concentration, a flow of 25 spm can be realized which results in a change of 600 ppm /br in the system. If a blender is bypassed, a flow of 80 gpm can be realized which will change the reactor coolant system concentration at the rate of 2000 ppm per hour.

i Specification B ensures that the assumptions of the Low Temperature RCS Overpressurization Analysis are set by allowing a maximum of one charging pump to be operabic when low temperature overpressurization protection is required.

References:

(1) FDSA Section 4.2 Amendnent No. 33, 97 3-9

3.7 . MINIMUM WATER ' VOLUME AND BORON CONCENTRATION IN THE REFUELING WATER STORAGE TANK Applicability: Applies to the inventory' of borated refueling water for core cooling systems, or containment spray.

Objective:

~

To protect against fuel damage and reduce containment pressure by ensuring immediate availability of core cooling and containment spray water.

Specification: Whenever the core cooling or containment spray systems are specified to be operable, the refueling water tank shall contain not less than 230,000 gallons and shall .have a .

boron concentration of not less than 2200 ppm.- - 1 Basis: .The volume and boron concentration requirements of

, 230,000 gallons and 2200 ppm, respectively, are consistent '

with the transient and LOCA analyses. The volume

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requirements bound the maximum expected boration 1 capability requirement based on end of cycle conditions' i

g. ' - af ter xenon decay and cooldown to 1400F - i I

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h L

-3.10 REACTIVITY CONTROL SYSTEMS 3.10.1 BORATION CONTROL SHUTDOWN MARGIN FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.10.1.1- The SHUTDOWN MARGIN shall be greater than or equal to 1800 pcm.

APPLICABILITY: MODES I and 2' (Four Loop Operation).

i ACTION:

With the SHUTDOWN MARGIN less than 1800 pcm,immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 14,000 ppm boron or its equivalent until the required SHUTDOWN MARGIN is restored.

1 SURVEILL ANCE REQUIREMENTS

1. The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1800 pcm: .

a. Within I hour af ter detection of a6 inoperable control rod (s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaf ter while the rod (s)is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN sha!! be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s);

b. When in MODE 1 or MODE 2 with Keff greater than or equal to I at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawals within the limits of Specification 3.10.2.6; !

c. When in MODE 2 with Ke gg less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by versfying that the predicted critical control rod position is within the limits of Specification 3.10.2.6; :

d. Prior to initial operation above 5% RATED THERMAL POWER af ter each fuelloading, by consideration of the f actors in 2 below, with the control banks at the maximum insertion limit of Specification 3.10.2.6 and

See Special Test Exceptions Specification 3 24.1. l Amendment No. U , y , 97 3-16

REACTIVITY CONTROL SYSTEM 5 SURVEILL ANCE REQUIREMENTS (Continued) -

2. During MODE 1 operation, the overall core reactivity balance shall be compared to predicted values to demonstrate agreement withini 1000 pcm at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least those factors stated below. The predicted reactivity values shall be adjusted (normalized),if required, to correspond to the actual core conditions prior to exceeding a f uel burnup of 60 EFPD af ter each fuel loading.

a) Reactor Coolant System boron concentration, '

b) Control rod position, c) Reactor Coolant System average temperature, d) Fuel burnup based on gro's s thermal energy generation, e) Xenon concentration, and f) Samarium concentration.

l l

3-17 Amendment No. 29, f/, 97

s REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1.2 The SHUTDOWN MARGIN shall be greater than or equal tot

a. 1800 pcm with no reactor coolant loop idled or isolated,' or

b. 2600 pcm with one or more reactor coolant loops idled or isolated, except during the startup of a reactor coolant pump.

APPLICABILITY: MODE 3.

ACTION: ,

With the SHUTDOWN MARGIN less than the required value,immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 14,000 ppm boron or its equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILL ANCE REQUIREMENTS The SHUTDOWN MARGIN r. hall be determined to be greater than or equal to the required value:

a. Within I hour af ter detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaf ter while the rod (s)is inoperable. If the

- inoperable control rod is immovable or untrippable, the SHUTDOWN MARGIN shc!! be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1) Reactor Coolant System boron concentration,

2) Control rod position,

3) Reactor Coolant System average temperature,

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration.

Idled or isolated refers to the position of the loop stop valves, not the reactor coolant pump operating status.

3-17a Amendment No. /p, ff, 97

F REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1.3 The SHUTDOWN MARGIN shall be greater than or equal to 2600 pcm.

APPLICABILITY: MODE 4 and 5.

ACTION:

With the SHUTDOWN MARGIN less than 2600 pcm,immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 14,000 ppm boron or its equivalent until the required SHUTDOWN MARGIN is restored.

l SURVEILLANCE REQUIREMENTS The SHUTDOWN MARGIN shall be determined to be greater than or

% equal to 2600 pcm

a. Within I hour af ter detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaf ter while the rod (s) is inoperable. If the Inoperable control rod is immovable or untrippable, the SHUTDOWN l

MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod (s); and I

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1) Reactor Coolant System boron concentration,

2) Control rod position, J

3) Reactor Coolant System average temperature, l

4) Fuel burnup based on gross thermal energy generation,

5) Xenon concentration, and

6) Samarium concentration. :

)

Amendment No. f/, 97 3-17b

REACTIVITY CONTROL $YSTEMS SHUTDOWN MARGIN THREE LOOPS OPER ATING LIMITING CONDITION FOR OPERATION __

3.10.1.4 The SHUTDOWN MARGIN shall be greater than or equal to 2600 pcm.

APPLICABILITY: MODES 1 and 2' (Three Loop Operation).

_ ACTION:

With the SHUTDOWN MARGIN less than 2600 pcm, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 14,000 ppm boron or its equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS

1. The SHUTDOWN MARGIN sha!! be determined to be greater than or equal to 2600 pcm:

a. Within I hour af ter detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaf ter while the rod (s)is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an incre'ased allowance for the withdrawn worth of the immovable or untrippable controf rod (s);

b. When in MODE 1 or MODE 2 with Kegg greater than or equal to 1 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.10.2.7;

c. When in MODE 2 with Kegg less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.10.2.7;

d. Prior to initial operation above 5% RATED THERMAL POWER af ter each fuelloading, by consideration of the factors in 2 below, with the control banks at the maximum insertion limit of Specification 3.10.2.7; !

and I l

See Special Test Exceptions Specification 3.24.1.

l L

l l

f l

1 " " "* N ' N '

3-17c .

i l

L_ _ _ _ _ _ -- )

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

2. During MODE 1 operation, the overall core reactivity balance shall be compared to predicted values to demonstrate agreement within +1000 pcm at least once per 31 Effective Full Power Days (EFPD). ThTs comparison shall consider at least those f actors stated below. The predicted reactivity values shall be adjusted (normalized) to correspond

to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD af ter each fuel loading, a) Reactor Coolant System boron concentration, b) Control rod position, c) Reactor Coolant System average temperature, d) Fuel burnup based on gross thermal energy generation, c) Xenon concentration, and I) Samarium concentration.

j 1

n h nt No. 29, 47, M , 97 I 3-17d

REACTIVITY CONTROL SYSTEMS l

MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.10.1.5 The moderator temperature coefficient (MTC) shall be:

i

a. Less positive than 5 pcm/0F for the a!! rods withdrawn, beginning of cycle life (BOL), hot zero THERMAL POWER condition; and

b. Less positive than 0 pcm/0F for the a!! rods withdrawn, BOL, R ATED THERMAL POWER ccndition; and

c. Less negative than -29 pcm/0F for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition.

APPLICABILITY: MODES I and 2" * *.

ACTION:

a. With the MTC more positive than the limits of Specifications 3.10.1.3a or 3.10.1.5b above, operation in MODES I and 2 rnay proceed provided:

1. Control rod withdrawal limits are established and maintained sufficient to restore the MTC to l'ess positive than 5 pcm/0F at hot zero THERMAL POWER or 0 pcm/0F at RATED THERMAL POWER within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits shall be in addition to the insertson limits of Specification 3.10.2.6 or 3.10.2.7; ;

)

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and 3 A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2, within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the MTC to within its limit for the all rods withdrawn condition. 1

{

i

b. With the MTC more negative than the limit of Specification 3.10.1.5c above, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

'With Keff greater than or equal to 1. .

l

*See Special Test Exceptions Specification 3.24.2.

] ,

I i

l 3-17e

?

REACTIVITY CONTROL SYSTEMS 9

SURVEILLANCE REQUIREMENTS The MTC sha!! be determined to be witnin its limits during each fuel cycle as follows:

a. The MTC shall be measured and compared to the BOL limit of Specification 3.10.1.5a. above, prior to initial oper& tion above 5% of RATED THERMAL POWER, af ter each fuel loading;

b. The measured MTC shall be adjusted to the BOL, RATED THERMAL POWER conditions and compared to the BOL limit of Specification 3.10.1.5b above, prior to initial operation above 5% of RATED THERMAL POWER, af ter each fuel loading; and

c. The measured MTC shall be adjusted to EOL, RATED THERMAL POWER conditions and compared to the EOL !!mit of Specification 3.10.1.5c above, prior to initial power operation above 5% of RATED THERMAL POWER, af ter each fuel loading.

3-17f

i l

REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.10.1.6 The Reactor Coolant System operating temperature (Tavg) shall be greater than or equal to 5250F.

APPLICABILITY: MODES 1 and 2+ * *. I l

1 ACTION:

less than 5250F restore With to a Reactor within its Coolant System limit within operating 15 minutes temperature or be,in HOT STA (Tavb)DBY within the next 15 Tavg ;

minutes.

SURVEILLANCE REQUIREMENTS The Reactor Coolant System temperature (Tavg) shall be determined to

be greater than or equal to $250F: ,

a. Within 15 minutes prior to achieving reactor criticality, and

b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System Tavg is less than 5300F with the Tavg -Tref Deviation Alarm not reset.

With Keff greater than or equal to 1.

*See Special Test Exceptions Specification 3.24.2.

3-17g Amendment No. 97 i

i.

s REACTIVITY CONTROL SYSTEMS 3.10'.2 MOVABLE CONTROL ASSEMBLIES l BANK HEIGHT j LIMITING CONDITION FOR OPERATION p 3.10.2.1 All shutdown and control rods shall be CPERABLE and positioned within 1 24 steps indicated position (RPI) of their bank position, as indicated by the average of the RPI for the respective bank.

APPLICABILITY: MODES l a and 2*.

I ACTION:

a. With one or more rods inoperable due to being immovable as a result of excessive friction or mechanicalinterference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.10.1.1 or Specification 3.10.1.4 is satisfied within I hour and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. With one rod trippable but inoperable due to causes other than addressed by ACTION a., above, or misaligned from its bank position by more than124 steps, l POWER OPERATION may continue provided that within I hour:

1. The rod is restored to OPERABLE status within the above alignment requirements, or

2. The rod is declared inoperable and the remainder of the rods in the bank with the inoperable rod are aligned to within124 steps of the inoperable j rod while maintaining the rod sequence and insertion limits of Figure 3.10-1 or Figure 3.10-2. The THERMAL POWER Level shall be restricted pursuant to Specification 3.10.2.6 during subsequent four loop operation or Specification 3.10.2.7 during subsequent three loop operation, or

3. The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.10.1.1 for four loop operation or Specification 3.10.1.4 for three loop operation is satisfied. POWER OPERATION may then continue provided that:

a) The THERMAL POWER level is reduced to less than or equal to 65% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the Nuclear Overpower Trip Setpoint is reduced to less than or equal to 74% of RATED THERMAL POWER.

b) The SHUTDOWN MARGIN requirement of Specification 3.10.1.1 for four loop operation or Specification 3.10.1.4 for three loop operation is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

'See Special Test Exceptions Specifications 3.24.2.

3-17h Amendment No. 97 l

REACTIVITY CONTROL SYSTEMS LIMITING CONDITIONS FOR OPERATION ACTION (Continued) c) A power distribution map is obtained from the movable incore 1 detectors and LHCR and FhH are verified to be within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and d) A reevaluation'of each accident analysis of Table 3.10-1 is performed within 10 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions;

c. With more than one rod misaligned from its bank position by more than 3,24 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS

a. The position of each rod shall be determined to be within the position limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

, except during time intervals when the rod position deviation monitor (Datalogger) is inoperable, then verify the individual rod positions at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . -

b. Each rod not fully inserted in the core shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

1 i

Amendment No. 97

- _ _ _ _ - _ _ _ _ _ _ _ k

TABLE 3.10-1 ACjjpENT ANALYSIS REQUIRING REEVALUATION gigEVENT OF AN INOPERABLE CONTROL ROD .

Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment t

Loss of Reactor Coolant from small Ruptured Pipes or from Cracks in Large Pipes Which Actuates the Emergency Core Cooling System Major Reactor Coolant Systerr Pipe Ruptures (Loss of Coolant Accident)

Major Secondary System Pipe Rupture Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly .

Ejection) i l

i i

l l

I 3-17j Amendment No. 97 j

REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS-OPERATING LIMITING CONDITION FOR OPERATION 3.10.2.2 The Digital Rod Position Indication System (step counters) and the Analog Rod Position Indication System (RPI) shall be OPERABLE and capable of determining the control rod positions within116 steps.

APPLICABILITY: MODES I and 2. t ACTION:

a. With a maximum of one analog rod position indicator per bank inoperable, restrict the movement of the rod bank which includes the nonindicating rod toi 8 steps from the position last determined prior to loss of the nonindicating rod, if the position of the nonindicating rod is not determined within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , declare 1 the rod inoperable and refer to Specification 3.10.2.1. l k b. With a maximum of one digital rod position indicator per bank inoperable either:

1. Verify that all analog rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 32 steps of each other at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or

2. With four loops operating, reduce THERMAL POWER to less than 6596 of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

3. With three loops operating, reduce THERMAL POWER to less than 1696 of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c. With more than one analog rod position indicator or digital rod position indicator per bank inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months with l the reactor trip breakers open or the control rod drive lif t coils deenergized. l 9

s SURVEILLANCE REQUIREMENTS Each digital and analog rod position indicator shall be determined to be OPERABLE by verifying that the bank average analog rod position and the digital rod position agree within 16 steps at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod position deviation monitor (Datalogger)is inoperable, then compare the analog rod position and the digital rod position at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

3-17k

REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEMS-OPERATING LIMITING CONDITION FOR OPERATION 3.10.2.2 The Digital Rod Position Indication System (step counters) and the Analog Rod Position Indication System (RPI) shall be OPERABLE and capable of determining the control rod positions within 116 steps.

APPLICABILITY: MODES I and 2.

ACTION:

a. With a maximum of one analog rod position indicator per bank inoperable, restrict the movement of the rod bank which includes the nonindicating rod to +~

8 steps from the position last determined prior to loss of the nonindicating rod.

If the position of the nonindicating rod is not determined within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , declare the rod inoperable and refer to Specification 3.10.2.1.

b. With a maximum of one digital rod position indicator per bank inoperable either:

1. Verify that all analog rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn ,

rod of the bank are within a maximum of 32 steps of each other at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or

2. With four loops operating, reduce THERMAL POWER to less than 6596 of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

3. With three loops operating, reduce THERMAL POWER to less than 1696 of RATED THERMAL POWER within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c. With more than one analog rod position indicator or digital rod position indicator per bank inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months with the reactor trip breakers open or the control rod drive lif t coils deenergized.

SURVEILLANCE REQUIREMENTS Each digital and analog rod position indicator shall be determined to be OPERABLE by verifying that the bank average analog rod position and the digital rod position agree within 16 steps at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the rod position deviation monitor (Datalogger)is inoperable, then compare the analog rod position and the digital rod position at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

1 3, Amendment No. 97

k i

- i REACT 3VITY CONTROL SYSTEMS f POSITION INDICATIO'N SYSTEM-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.10.2.3 The analog rod position indicator shall be OPERABLE and capable of ;

determining the control rod position within f,16 steps for each shutdown or control rod not fully inserted.

APPLICABILITY: MODES 3* ", 4 + ", and 5 * ". .

ACTION:

With less than the above required rod position indicator (s) OPERABLE, immediately open the Reactor Trip System breakers.

SURVEILLANCE REQUIREMENTS Each of the above required analog rod position indicator (s) shall be determined to be OPERABLE at least once per refueling by verifying that the digital rod position indicators agree with the analog rod position indicators within 16 stepa when exercised over the full-range of rod travel.

l

With the Reactor Trip System breakers in the closed position. 1

'See Special Test Exemptions Specification 3.24.3.

3-17L

l REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.10.2'.4 The measured individual rod drop time from fully withdrawn position shall be less than 2.5 seconds from the fully withdrawn position to the bottom of the dashpot with Tavg Breater than $250F and four reactor .

coolant pumps operating. A drop time of 2.45 sec. shall be used if only 3 reactor coolant pumps are operating while the drop tests are rnade.

APPLICABILITY: MODES I and 2.

ACTION:

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

SURVEILLANCE REQUIREMENTS The rod drop time shall be demonstrated through measurement prior to s reactor criticality:

a. For all rods following each removal of the reactor vessel head, For specifically affected individual rods following any maintenance j b.

on or modification to the control rod drive system whi::h could j affect the drop time of those specific rods, and l c.. ' At least once per 18 months.

l l

3-17m Amendment No. 97

l-s .

i REACTIVITY CONTROL SYSTEM _S _

SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION F'OR OPERATION 3.10.2.5 All shutdown rods shall be withdrawn to equal to or greater than 320 steps. ,

- APPLICABILITY: MODES 1

cnd 2+ * *.

ACTION:

With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing,within I hour either:

a. . Fully withdraw the shutdown rod, or

b. Declare the shutdown rod to be inoperable and apply Specification 3.10.2.1.

SURVEILLANCE REQUIREMENTS Each shutdown rod shall be determined to be fully withdrawn:

a. Within 15 minutes prior to withdrawal of any rods in Control Bank A or 3 during an approach to reactor criticality, and

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

'See Special Test Ex:eptions Specification 3.24.2.

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i l

1 l

Amendment No. 97 l

3-17n

REACTIVITY C.' wTROL SYSTEMS CONTROL GROUP INSERTION LIMITS, FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.10.2.6 The control banks shall be limited in physical insertion as shown in Figure 3.10 1.

APPLICABILITY: MODES l' and 2* * * (Four Loop Operation).

l ACTION:

l With the control banks inserted beyond the above insertion limits, except for i

i surveillance testing:

a. Immediately initiate action to restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or l b. Immediately initiate action to reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the above figure, or 1

c. Be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS I

l The position of each control bank sha!! be determined to be within the insertion ilmits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the individual rod positions.

5ee Special Test Ex5ptions Specification 3.h.2.

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i REACTIVITY CONTROL SYSTEMS CONTROL GROUP INSERTION LIMITS THREE LOOPS OPER ATING LIMITING CONDITION FOR OPERATION 3.10.2.7 The control banks shall be limited in physical insertion as shown in Figure 3.10-2.

APPLICABILITY: MODES l' and 2* " (Three Loop Operation).

ACTION:

With the control banks inserted beyond the above insertion limits, except for surveillance testing: ,

a. Immediately initiate action to restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

b. Immediately initiate action to reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the above flgure, or

c. Be in at least HOT STANDBY within 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

SURVEILLANCE REQUIREMENTS The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by_ verifying the individual rod positions.

See Special Test Exceptions Specification 3.24.2.

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3.10 REACTIVITY CONTROL SYSTEMS BASES 3.10.1 BORATION -CONTROL 3.10.1.1, 3.10.1.2, and 3.10.1.3 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that:(1) the reactor can be made subcritical from all operating conditions,(2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tave. The most restrictive condition during MODES 1,2 and 3 occurs at end-of-cycle liYe (EOL), and is associated with a postulated steam line break accident and resulting RCS cooldown. In the accident analysis, a minimum SHUTDOWN MARGIN of 1800 pcm for four loop operation and 2600 pcm for three loop operation is assumed. Operation in MODE 3 with two operating reactor coolant pumps is bounded by the four loop steam line break analysis. Operation in MODE 3 with one operating reactor coolant pump and two OPERABLE reactor coolant loops (both loop stop valves open in each loop) is bounded by the three loop

- steam line breai< analysis. Because of the short time involved, the 2600 pcm SHUTDOWN MARCIN limit need not be applied to the closure of the cold leg loop stop valve.in order to restart the reactor coolant pumps from an initial four loop operation condition. The most restrictive condition in MODES 4 and 5 is associated with the boron dilution accident. In the analysis of this accident, a minimum SHUTDOWN ,

MARGIN of 2600 pcm in MODES 4 and 5 is required to control the reactivity transient. j I

Accordingly, the SHUTDOWN MARGIN requirements are based upon this limiting condition and are consistent with current design basis assumptions.

3.10.1.4 MODERATOR TEMPERATURE COEFFICIENT The limits on the moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the accident and transient analysis.

The MTC values of this specification are associated with a specific set of plant j conditions; measurement of MTC values at conditions other than those explicitly stated i with extrapolation to the specified conditions is acceptable. Correction factors shall ;

account for fuel and moderator temperature and boron concentration. l 4

3-17q Amendment No. 97

7,,

REACTIVITY CONTROL SYSTEMS

)

, j l

(

BASES MODERATOR TEMPERATURE COEFFICIENT (Continued) {

1 .

i l The Surveillance Requirement for measurement of the MTC at the beginning of the fuel f l cycle is adequate to, confirm that the MTC remains within its limits since this :

coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. l l

3.10.1.5 MINIMUM TEMPERATURE FOR CRITICALITY j This specification ensures that the reactor. will not be made critical with the Reactor Coolant System average temperature less than 5250F. This limitation is required to ensures (1) the moderator temperature coefficient is within it analyzed temperature range, (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the !

reactor vesselis above its minimum RTNOT temperature.

~

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4 i

l I

l 3-17r Amendment No. 97

REACTIVITY CONTROL SYSTEMS BASES

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

OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the bank average Analog Rod Position Indicator within + 16 steps provides assurance that the Digital Rod Position Indicator and the Analog Rod Position Indicator is operating correctly over the full range of indication.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met. Continued operation with an inoperable rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod integrity during continued operation. In addition, those safety analyses affected by an inoperable rod are reevaluated to confirm that the results remain valid during future operation.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with Tavg equal to or greater than 5250F and with four or three reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a Reactor trip at operating conditions.

Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied. l l

i I

l 3-17s

l l

3.31 CONTA2NMENT Applicability: Applies to the operating statur of reactor containment.

Objective: To insure containment integrity.

l Specification: A. Leakage The reactor shall not be critical if the l containment leakage exceeds 0.25 weight percent of the contained air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when extrapolated to 40 psig in accordance

, with Surveillance Standard 4.4.

B. Containment Integrity (1) Containment integrity shall be maintained whenever the reactor coolant system is above 300 psig and 200*F. The shutdown margin shall be greater than 2600pcm when the containment is open.

(2) Containment integrity shall not be '

I violated when the reactor vessel head is removed unless the reactor coolant system is borated to the refueling j boron concentration.

~

l (3) Positive reactivity changes shall not ./

be made by rod drive motion or boron dilution whe'ever n the containment integrity is not intact. p/

/

C. Internal Pressure 4,

The reactor shall not be critical if the containment interna? pressure exceeds 3 psig, or the internal vacuum exceeds 2.0 psig. j D. Air Recirculatlyn,Systy Three of the four air recirculation units shall be operable whenever the reactor is critical.

E. Containment Spray System The containment spray system shall be

. operable whenever the reactor is critical.

F. Containment Venting (1) Either the containment air particulate monitoring system or the containment purge exhaust system shall be available at all times when the reactor is critical for post accident hydrogen j

"*"[f"8' Arendment No. 97

. -l (2) Cntainme::t purg2 c:ptbility may be i rendered inoperable when the reactor

{

is critical by placing a blank flange on the 42-inch purge air exhaust penetration inside the reactor contain-ment for a period of seven days. If

, the blank flange cannot be removed within seven days, then the reactor shall be shut down within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

G. The containment isolation valves specified in Table 3.11-1 shall be operable while in Modes 1, 2, 3, and 4, or:

With one or more of the isolation valve (s) specified in Table 3.11-1 inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either; t (1) Restore the inoperable valve (s),to OPERABLE status within four hours, or (2) Isolate each affected penetration within four hours by use of at least one deactivated au.'omatic valve secured in the isolation position, or (3) Isolate each affected penetration within four hours by use of at least one closed manual valve or blind flange, or

, (4) Be in at least HOT STANDBY within the next six hours and in COLD SHUTDOW within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

H. Containment Isolation f

The Containment Isoletion actuation system shall be operable with the following trip

. setpoints:

Containment Pressure - H1 1 5 psig Pressurizer Pressure - LOW > 1700 psig

^

Basis: A containment leakage rate of 0.3 weight percent of the contained air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at an internal pressure of 40 psig under hypothetical accident conditions with 3 of 4 air recirculation units operating will maintain public exposure well below 10 CFR 100 values (see Section 10.4 of the FDGA).

The reactor coolant system conditions of 300 psig i

and 200*F assure that no steam will be formed and hence there would be no pressure buildup in ,

the containment if a reactor coolant system rupture were to occur.

I

{ 3-19 Anendnent No. 4, 97 i

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f 3.13,REJyELING

,Ag.(chdility: Applies to operating limitations during refueling operations, s / t LObjective: To insure that no incident could occur during refueling

'; operations that would affect public health and safety.

t Specf.fication: A. Radiation levels in'the Containment s'nd Fuel Storage j (( Building shall be monitored continuously.

,9 v B. Core suberitical neutron flux shall be continuously et monitored by at least two neutron monitors, each with continuous visual and audible indication s i3 available, whenever-core geometry is being changed.

5 Wen core geometry is not being changed, at least one

a. neutron flux monitor shall be in service.

C. (1) Wenever the water level in the refueling cavity is less than 23 feet above the flange of the

, reactor pressure vessel, two residual heat

/ !

removal loops shall be OPERABLE. With less than the required RHR loops OPERABLE, immediately initiate corrective action to return the required t RER loops to OPERABLE status as soon as possible.

Also, with less than the required depth of water

described above, suspend all operations. invc1ving i movement of fuel assemblies or control rods within 3

the reactor pressur.e vessel. .

y (2) At least one RHR pump and heat exchanger shall be

, in operation except that the residual heat removal

( system may be removed from operation for up to I 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during performance of

\ core alternations in the vicinity of the reactor pressure vessel bot legs. With less than one

[ 1; \ residual heat removal pump and heat exchanger in f 4 operation except as described above, suspend all f li operation involving any increase in reactor decay i heat load or a reduction in boron concentration of the reactor coolant system. Close all contain-i i. 1 ~

J q i ment penetrations providing direct access from the containment atmosphere to the outside atmosphere

s. within four hours.

D. During reactor vessel head removal and while loading

(/

t and unloading fuel from the reactor, the boron l concentration of all filled, unisolated portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure a keff less than or equal to 0.94.

i 3-23 Amendment No. 7, EE, 39, 97 l c

^

l

_ - ~ - . _ . _ - - - _ - _ - - - - . . l

E. ' On2 ch::rging pump ctpabla of injseting' barste'd water to.the reactor coolant shall be available at all-time when changes in core geometry are taking place.

F. Whenever new fuel is added to the reactor core,.a 1/M plot shall be maintained to verify the suberiticality of the core.

G. Direct communication between the control Room and the refueling cavity manipulator crane shall be available whenever changes in core geometry are taking place.

L Spent fuel casks shall_not be, handled above the spent .

fuel pool or its edge except as provided.in Section 3.13.I.

until such time as NRC has received and approved the spent fuel cask drop evaluation.

I. After April 23, 1980, a spent fuel cask may be brought i into the Spent Fuel Building and may be moved into.or over t1:e spent fuel pool a _ total of ten times in order to remove fuel from the pool for study at an off-site laboratory or to return the fuel from the laboratory to the pool. Movement of the spent fuel cask under the provisions of this paragraph is conditioned on compliance (by the licensee) with all commitments made by the 1980, licensee April in 23,its1980.

letters In to the.NRC.

addition, dated April 18,1.he and all fuel within spent fuel pool shall have decayed for at least 90 days-before a spent fuel cask is handled above the pool.

Basis: The equipment and general procedures to be utilized during refueling are discussed in the Facility Description and Safety Analysis. Detailed instructions will be available for_use by refueling personnel. These instructions, the above-specified precautions, and the design of-the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety. Whenever no change is being made in core geometry, one flux monitor is sufficient. This permits maintenance of the instrumentation.

Continuous monitoring of radiation levels (A above) and neutron flux provides immediate indication of an unsafe condition. The residual heat removal pump is used to maintain a uniform boron concentration. The refueling boron concentration indicated in Part D will keep the core suberitical even if all control rods were withdrawn from the core. For a core configuration of all rods in, an additional 0.05 k penalty is required to account for a heavy load crushiIhthecoreintoamorereactiveconfiguration. <

Weekly checks of refueling 3-24 Amendment No. M , 97

1 i

l Intentionally Left Blank i

i i

1 3-27 Artendment No. 97

I

\

t I

Intentionally Left Blank Amendment No. 74, 97

3.17 POWER DISTRIBUTION LIMITS 3.17.1 AXIAL OFFSET FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.17.1.1 The AXIAL OFFSET shall be maintained within the limits of Figures 3.17-la or b.

APPLICABILITY: MODE 1, above 40% of RATED THERMAL POWER.

ACTlON:

With the AXIAL OFFSET outside the Acceptable Operation Limits specified in the above figures, within 15 minutes initiate corrective action and continue the corrective action so that the Axial OFFSET is within limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 40% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS

a. The AXIAL OFFSET shall be determined to be within the Acceptable Operation Limits of Figures 3.17-1 a or b by monitoring the AXIAL OFFSET using at least two OPERABLE excore Power Range channels and applying,the excore/incore correlation on a continuous basis.

b. The excore/incore correlation sha!! be verified at least once per 31 EFPD and adjusted at least once per 92 EFPD using the results of the measurements obtained in accordance with Specification 3.17.2.

I

c. The excore/incore correlation shall be determined af ter each fuel loading or maior change in excore Power Range instrumentation prior to exceeding 80% of RATED THERMAL POWER.

d. The excore Power Range detectors shall be calibrated / correlated relative to the Movable Incore Detector System measurements within 7 days af ter completion of incore measurements.

3-30

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POWER vs OFFSET, 0-250 EFPD, FOUR LOOP l

I Amendment No. 97 l

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~~l i~~

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~~% OFFSET FIGURE 3.17-2B POWER vs OFFSET, 250-EOC, THREE LOOP Amendment No. 97 A~~

., .j POWER DISTRIBUTION LIMITS 3.17.2 LINEAR HEAT GENERATION RATE FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.17.2.1 All LINEAR HEAT GENERATION RATES (LHGRs) shall not exceed the following kilewatt per foot limits for cycle residency time:

~~a. Less Than 125 EFPD 14.3 kW/f t~~
~~b. 125 To 250 EFPD 14.5 kW/f t~~
~~c. Greater Than 250 EFPD But Less Than END-O?-CORE LIFE - 15.5 kW/f t i~~
APPLICABILITY: MODE .1, above 40% of RATED THERMAL POWER.
ACTION:
With the LHCR of any fuel rod exceeding the limits specified above, initiate corrective action within 15 minutes and continue corrective action so tha the LHGR is within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 40% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
SURVE!LLANCE REQUIREMENTS
~~1. LHGR's shall be determined to be within the above limits by a core power distribution measurement using the Moveable incore Detector System and in consideration of the factors listed in 2 below:~~
~~a. At least once per 31 EFPD,~~
~~b. Prior to THERMAL POWER exceeding 80% of RATED THERMAL POWER af ter each fuelloading, and~~
~~c. After reaching 100% of RATED THERMAL POWER and achieving equilibrium xenon conditions af ter each refueling.~~
l l
l i l i Amendment No. 23, 7f, 97
o c.
l.
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
~~2. Measured values of core power peaking factors used in determining LHGRs shall include the following allowances:~~
~~a. Normal power peaking * **,~~
~~b. Flux peaking augmentation factors (Power Spike)*, Figure 3.17-3~~
~~c. Measurement uncertainty of 1.05,~~
~~d. Statistical density factor of 1.012,~~
~~e. Engineering factor of 1.02,~~
~~f. Stack shortening / thermal expansion factor of 1.007, and~~
~~g. Power level uncertainty of 1.02.~~
~~ltems a. and b. are ' chosen at a core height to maximize the product.~~
~~Determined using tt)e thimble location which yields the higher total core peaking factor.~~
.I l
l J
3-31b Amendment No. 97 1
1
i POWER DISTRIBUTION LIMITS .
THREE LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.17.2.2 All LINEAR HEAT GENERATION RATES (LHGRs) shall not exceed the following kilowatt per foot limits for cycle residency time:
~~a. Less Than 125 EFPD 9.295 kW/f t~~
~~b. 125 To 250 EFPD 9.425 kW/f t~~
~~c. Greater Than 125 EFPD But Less Than END-OF-CORE LIFE 10.075 kW/f t APPLICABILITY: MODE 1, above 40% of RATED THERMAL POWER.~~
ACTION:
With the LHCR of any fuel rod exceeding the limits specified above, initiate corrective action within 15 minutes and continue corrective action so that the LHGR is within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 40% of RATED THERMAL POWER j within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
SURVEILLANCE REQUIREMENTS l
1. LHCR's shall be determined to be within the above limits by a core power distribution measurement using the Moveable Incore Detector i System and in consideration of the factors listed in 2 below: l
~~a. At least once per 31 EFPD,~~
~~b. Prior to THERMAL POWER exceeding 50% of RATED THERMAL POWER af ter each fuel loading, if the initial power ascension is j performed with three loops operating.~~
. c. After achieving 65% of RATED THERMAL POWER and achieving l equilibrium xenon conditions af ter each refueling, if the initial power ascension is performed with three loops operating. ;
I l
I Amendment No. 97 3.31 c
POWER DISTRIBUTION LIMITS
.SURVEILL ANCE REQUIREMENTS (Continued) ,
~~2. Measured values of core power peaking factors used in determining LHGRs shallinclude the fo!!owing a!!owances:~~
1
~~a. Normal power peaking * * *,~~
~~b. Flux peaking augmentat. ion factors (Power Spike)*, Figure 3.17-3~~
~~c. Measurement uncertainty of'l.05,~~
~~d. Statistical density factor of 1.012,~~
~~e. Engirwering factor of 1.02, .~~
~~f. Stack shortening / thermal expansion factor of 1.007, and~~
~~g. Power level uncertainty of 1.02.~~
~~ltems a. and b. are chosen at a core height to maximize the product.~~
~~Determined using the thimble. location which yields the higher total core peaking factor.~~
Amendment No, 97 3-31d l
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POWER DISTRIBUTION LIMITS 3'17 3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - Fyg FOUR LOOPS OPERATING LIMITING CONDITION FOR OPERATION 3.17.3.1 The NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR, F,N,H, shall-be limited by the following relationship FNH f 1.60[1.0 + 0.3 (1 -P)3 where: p= THERMAL POWER ,and RATED THERMAL POWER FyH, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR,is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.
APPLICABILITY: MODE 1 ACTION: With F[H outside of the above specified limit:
~~a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:~~
~~, 1.- Restore F 3 N H to within the above limit, or~~
~~2. Reduce THERMAL POWER to less than 65% of RATED THERMAL POY/ER and reduce the Nuclear Overpower Trip Setpoint to less than or equal to 74% of RATED THERMAL POWER.~~
b. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above limit, verify through incore flux mapping that FNH is restored to within the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; and
c. Identify and correct the cause of the out-of-limit condition prior to increasing THERM AL POWER above the reduced THERMAL POWER limit required by ACTIONS a.2. and/or b. above subsequent THERMAL POWER operation may proceed provided that F H is demonstrated, through incore flux mapping, to be within the a ove limit prior to l exceeding the following THERMAL POWER levels:
~~1. A nominal 50% of RATED THERMAL POWER,~~
~~2. A nominal 75% of RATED THERMAL POWER, and~~
3. Within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of attaining greater than or equal to 95% of i RATED THERMAL POWER.
Amendment No. 97
POWER DISTRIBUTION LIMITS SURVEILL ANCE REQUIREMENTS The NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR shall be determined to be within the above limit:
I
~~a. Prior to operation above 80% of RATED THERMAL POWER af ter I~~
~~each fuel loading, and ,~~
~~b. At least once per 31 Effective Full Power Days.~~
l Amendment No. 97 3-31f 1
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POWER DISTRIBUTION LIMITS THREE LOOPS OPERATING LIMITING CONDITION FOR OPERATION .,
3.17.3.2 The NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR, Fyg, shall be limited by the following relationship:
FMH f 1.64 1.0 + 0.3 (0.65 - P)'3 where: p, THERMAL POWER ,and RATED THERMAL POWER FgH, NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR,is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.
APPLICABILITY: MODE 1 ACTION: With F[g outside of the above specified limit:
~~a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:~~
~~1. Restore Fyg to within the cbove limit, or~~
~~2. Reduce THERMAL POWER to less than 20% of RATED THERMAL POWER and reduce the Nuclear Overpower Trip Setpoint to less than or equal to 25% of RATED THERMAL POWER,~~
b. Within 24 hows of initially being outside the above limit, verify through incore flux mapping that FNH is restored to within the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; and
c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTIONS a.2. and/or b. above-POWER operation may proceed providedhthat Fqsubsequent THERMAL H is demonstrated, through incore flux mapping, to be within the above limit prior to exceeding the following THERMAL POWER levels:
~~1. A nominal 50% of RATED THERMAL POWER 1~~
1 Amendment No. 97 3-31g
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1 1
POWER DISTRIBUTION LIMITS
$ SURVEILLANCE REQUIREMENTS ,
The NUCL' EAR ENTHALPY RISE HOT CHANNEL FACTOR shall be determined to be within the above limit l a. . Prior to operation above 50% of RATED THERMAL POWER af ter each fuel loading if the initial power ascension was performed with three loops operating. -
~~b. At least once per 31 Effective Full Power Days.~~
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Amendment No. 97 3-31h I
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POWER DISTRIBUTION LIMITS
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3.17'.4 QUADRANT POWER TILT RATIO LIMITING CONDITIONS FOR OPERATION 3.17.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.
APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER.
ACTION:
~~a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:~~
~~1. Calculate the QUADRANT POWER TILT RATIO at least once per I hour until either: !~~
a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or ,
b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
~~2. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:~~
a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Nuclear Overpower Trip Setpoint within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce j THERMAL POWER to less than 65% of RATED THERMAL POWER l within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Nuclear Overpower Trip f Setpoint to less than or equal to 74% of RATED THERMAL 1 POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and
{
~~4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION )~~
above 50% of RATED THERMAL POWER may proceed provided l that the QUADRANT POWER TILT RATIO is verified within its j limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified J accentable at 95% or greater RATED THERMAL POWER.
Amendment No. 97 3-311 i l
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t POWER DISTRIBUTION LIMITS LIMITING CONDITIONS FOR OPERATION ACTION (Continued)
~~b. .With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:-~~
~~1. Calculate the QUADRANT POWER TILT RATIO at least once per -~~
hour until either:
a) The' QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
~~2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0, within 30 minutes;~~
~~3. Verify that the QUADRANT POWER TILT RATIO is within its .~~
limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 65% of RATED THERMAL POWER within the ' I next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 'and reduce the Nuclear Overpower Trip S, tpoint to less than or equal to 74% of RATED THERMAL POWF 1 within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and 1
4. Identify and correct the cause of the out-of-limit condition prior to I increasing THERMAL POWER; subsequent POWER OPERATION .l above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its i limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified j acceptable at 95% or greater RATED THERMAL POWER. j i
c. With the QUADRANT POWER TILT , RATIO determined to exceed 1.09 l due to causes other than the misalignment of either a shutdown or l control rod: {
1
~~1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:~~
l a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or I
b) THERMAL POWiiR is reduced to less than 50% of RATED l THERMAL POWER.
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Amendment No. 97 3-31j
P POWER DISTRIBUTION LIMITS LIMITING CONDITIONS FOR OPERATION ACTION (Continued)
2. Reduce THERMAL POWER to less than 20% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Nuclear Overpower Trip Setpoint to less than or equal to 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; and
3. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 20% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified at 95% or greater RATED THERMAL POWER.
~~SURVEILLANCE REQUIREMENTS~~
a. The QUADRANT POWER TILT RATIO shall be determined within 7 days af ter completion of the incore power distribution measurements used to determine the initial excore/incore correlation af ter each fuelloading or major change in excore Power Range instrumentation,
~~b. The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by calculating the ratio at least once per 7 days.~~
c. The QUADRANT POWER TILT RATIO shall be determined to be within the limit when above 75% of RATED THERMAL POWER with one Power Range channelinoperable by using the Movable incore Detector System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
o i
l Amendment No. 97 3-31k
LIMITING CONDITION FOR OPER ATION 3.17.5 The following DNB-related parameters shall be maintained within the limits shown in Table 3.17-1:
~~a. Reactor Coolant System T Cold,~~
~~b. Pressurizer Pressure~~
~~c. Reactor Coolant System Flow. Rate APPLICABILITY: MODE 1.~~
l ACTION:
With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
SUR EILLANCE REQUIREMENTS
a. Each of the parameters of Table 3.17-1 shall be verified to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
~~b. The RCS total flow rate shall be determined by heat balance within seven EFPD of achieving RATED THERMAL POWER after a refueling.~~
~~c. The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.~~
Amendment No. 97 3-31L i
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'3.17 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuelintegrity.during Condition 1
- (Normal Operation) and 11(Incidents of Moderate Frequency) events bya' (1) maintaining r .'the minimum DNBR in the core greater than or equal to 1.30.during normal operation and in short-term. transients, and (2) limiting the' fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criterra. In .
addition, limiting the peak linear power density during Condition i events provides assurance that the initial conditions assumed for the LOCA analyses are' met and the ECCS Interim' acceptance criterion limit of 23000F peak cladding temperature is not-
~~exceeded.-~~
3.17.1 AXIAL OFFSET The AX1AL OFFSET specification provides continuous con'firmation of acceptable LINEAR HEAT GENERATION RATES (LHCR) during the time interval between incore measurements.
3.17.2 LINEAR HEAT GENERATION RATE Limiting the peak LINEAR HEAT GENERATION RATE (LHGR) during Condition I .!
events provides assurance that the initial condition assumed for LOCA analyses are met and the ECCS Interim acceptance criterion limit of 23000F peak cladding temperature 3 is not exceeded. These limits are based on a mirimum inlet temperature of 5360F at .l RATED THERMAL POWER.
3.17.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limit on the NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (F]g) ensures that the minimum DNBR limit is not exceeded The Fjg is measurable, but will normally only be determined periodically. This
. periodic surveillance is sufficient to insure that the limits are maintained provided:
~~a. The control rod insertion limits of Specification 3.10.2.6 and 3.10.2.7 are maintained, and~~
' b. The AX1AL OFFSET limits of Specification 3.17.1.1 and 3.17.1.2 are maintained.
The relaxation of F]H as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. The full power limits (1.60 for four loop, and 1.64 for three loop) include a 4% incore measurement uncertainty.
1 Amendment No. 97 4
3-31m
- = _ _ - - _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ .
POW $R DISTRIBUTION LIMITS 1
l BASES 1
. 3.17.4 QUADRANT POWER TILT RATIO I
j The QUADRANT POWER TII.T RATIO limit assures that the radial power distribution i
satisfie's the design values used in power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power l i
operation, The limit of 1.02, at which corrective action is required, provides DNB and LINEAR _
HEAT GENERATION RATE protection with x-y plane power tilts. A limit of 1.02 was f selected to provide an allowance for the uncertainty associated with the indicated power tilt.
The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on the limiting LHGR is reinstated by reducing the maximum allowed power s by 3% for each percent ci tilt.
For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.
The incore detector monitoring is done with at least one set of four quadrant symmetric thimbles.
I Amendment No. 97
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i POWER DISTRIBUTION LIMITS .
BASES 3.17.5 DNB PARAMETERS
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The limits on the DNB4eMttd parameters assure that each of the parameters are q, maintained within the normahteady state envelope of operation assumed in the 7,e I
-' transient and accident analises. The lirdts are consistent with the accident
'b, t/p assumptions. The iridicand values of $42.00F and 2000 psig correspond to analytical' i i.
'1
' limits of 544.loF and 1960 psia, respectively, with allowances for. measurement s,i , '
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uncertainty. The four and three loop flow rates are the values assumed in the analyses? '
t. The measured values are reduced by the appropriate measurement uncertainty prior to s comparison with the limits. The four loop measurement and uncertainty are based on a '
heat balance and the results are used to calibrate the RCS flow indicators. The three loop flow rate will be deemed acceptable if 0.76 times the measured four loop flow rate .; c-l' is greater than the three loop flow rate requirement. Previous flow rate measurements _I have shown that the three loop flow rate is 0.77 times the four loop flow rate with measurement uncertainties included.
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3.24 SPECIAL TEST EXCEPTIONS '
3.24.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.24.1 The SHUTDOWN MARGIN requirement of Specification 3.10.1.1 or 3.10.1.4 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated rod worth is available for trip insertion from OPERABLE control rod (s).
APPLICABILITY: MODE 2.
ACTION:
~~a. With any control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately <~~
initiate and continue boration at greater than or equal to 30 gpm of 14000 ppm boric acid solution or its equivalent until the
{
SHUTDOWN MARGIN required by Specification 3.10.1.1 or 3.10.1.4 '
. Is restored. s
b. With all control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 30 gpm of 14000 ppm boric acid solution or its equivalent until the SHUTDOWN MARGIN required by' Specification 3.10.1.1 or 3.10.1.4 is restored.
~~i SURVEILLANCE REQUIREMENTS I~~
a. The position of each control rod either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
b. Each control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 30% withdrawn position within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.10.1.1 or 3.10.1.4 o
Amendment No. 97 3-47
.SPECIAL TEST EXCEPTIONS .
3.24.2 PHYSICS' TESTS 1 LIMITING CO_NDITION FOR OPERATION
' 3.24.2 The limitations of Specifications 3.10.1.5, 3.10.1.6, 3.10.2.1, 3.10.2.5, 3.10.2.6 and 3.10.2.7 may be suspended during the performance of PHYSIC' TESTS provided:
~~a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,~~
~~b. The Reactor Trip Setpoints on the OPERABLE Power Range channels are set at less than or equal to 25% of RATED THERMAL POWER, and~~
~~c. The Reactor Coolant System lowest operating loop temperature (Tavg) is greater than or equal to 5150F.~~
APPLICABILITY: MODE 2.
ACTION:
~~a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the reactor trip breakers.~~
b. With a Reactor Coolant System operating loop temperature (Tavg) less than $150F, restore Ta to within its limit within 15 minutes or be in at least HOT STANYBY within the next 15 minutes. ,
~~SURVEILLANCE REQUIREMENTS~~
~~a. The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.~~
b. Each Power Range channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS TESTS.
~~c. The Reactor Coolant System temperature (Ta vg) shall be determired to be greater than or equal to 5150F at least once per 30 minutes during PHYSICS TESTS.~~
Amendment No. 97 3-48
SPECIAL TEST EXCEPTIONS 3.24.3 POSITION INDICATION SYSTEM - SHUTDOWN ,
.. LIMITING CONDITION FOR OPERATION i
3.24.3 - The limitations of Specification 3.10.2.3 may be suspended during the !
performance of individual shutdown and control rod drop time measurements provided;
~
~~a. Only one shutdown or control bank is withdrawn from the fully inserted position at a time, and~~
b. The analog rod position indicator is OPERABLE during the withdrawal of the rods.*
APPLICABILITY: MODES 3,4, and 5 during performance of rod drop time measurements.
ACTION:
With the Position Indication Systems inoperable, or more than one bank of rods withdrawn, immediately open the reactor trip breakers.
SURVE!LL ANCE REQUIREMENTS The above required Position Indication Systems shall be determined to be OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of and at least once per 24 ,
hours thereafter during rod drop time measurements by verifying the Bank Average Analog Rod Position Indication System and the Digital Rod Position Indication System agree within 16 steps.
This requirement is not applicable during the initial calibration of the Analog Rod Position Indication Syst'em provided:(1) Kegg is maintained less than or equal to 0.94, and (2) only one shutdown or control rod bank is withdrawn from the fully inserted position at one time.
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l Amendment No. 97 .
3-49 1
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3.24 SPECIAL TEST EXCEPTIONS BASES 3.24.1 SHUTDOWN MARGIN This Special Test Exception provides that a minimum amount of control rod worth is immediately available for reactivity control when tests are performed for control rod worth measurement. This special test exception is required to ,
permit the periodic verification of the actual versus predicted core reactivity condition occuring as a result of fuel burnup or fuel cycling operations.
3.24.2 PHYSICS TESTS !
i This Special Test Exception permits PHYSICS TESTS to be performed at less than or equal to 3% of RATED THERMAL POWER with the RCS Tave slightly lower than normally allowed so that the fundamental nuclear characteristics of the core and related instrumentation can be verified. In order for various characteristics to be accurately measured,it is at times necessary to operate ;
outside the normal restrictions of these Technical Specifications. For instance, to measure the moderator temperature coefficient at BOL, it is necessary to ,
position the various control rods at heights,which may not normally be allowed by Specification 3.10.2.6 or Specification 3.10.2.7 which in turn may cause the RCS Tavg to fall slightly below the minimum temperature of Specification ;
I 3.10.1.6.
3.24.3 POSITION INDICATION SYSTEM-SHUTDOWN l This Special Test Exception permits the Position Indication Systems to be l inoperable during rod drop time measurements. The exception is required since the data necessary to determine the rod drop time are derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage i is small compared to the normal voltage and, therefore, cannot be observed if i
the Position Indication Systems remain OPERABLE.
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i Amendment No. 97 3-50 l l
4.9 MAIN STEAM ISOLATION VALVES l
Applicability: Applies to periodic testing o'f the main steam .!
isolation valves.
Objective: To verify the ability of the main steam isolation valves to close upon signal.
i Specification: The main steam isolation valves will be tested cach quarter for movement of the valve disc through a distance of approximately one and one-ha'lf inches. This quarterly test may be conducted while the station is on lihe. Simul-taneous closure of all four valves within ten seconds shall be verified each cold shutdown if it had not been done in the' previous three months.
Basis: ihe main steam isolation valves serve to limit the reactor coolant syst,em cooldown rate and resultant reactivity insertion, whenever the moderator coefficient of reactivity is negative, following a main steam break incident. A closure time of ten seconds has been shown to yield acceptable results.
The time period between verification of the ten second closure requirement is based on ASME Boiler and Pressure Vessel Code Section XI, Article IW-3410(a)(b) and Westinghouse Standard :
Technical Specifications, for Pressurizer Water Reactors (5/15/76).
~~Reference:~~
(1) FDSA -- Section 8.1.
(2) ASME Boiler and Pressure Vessel Code Section XI, Article IW-3420(a) and (b).
(3) Standard Technical. Specification 4.0.5.
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Amendment No. U, 97 l 4-18 l
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ML20236S373

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2.1 INTRODUCTION

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ML20236S373

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2.1 INTRODUCTION

References:

References:

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mi C =

Reference:

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