ML20002A377

Amend 43 to License DPR-3,incorporating Provision in Tech Specs Re Requirement for Operation W/Refueled Core 13

ML20002A377
Person / Time
Site:	Yankee Rowe
Issue date:	08/25/1977
From:	Goller K Office of Nuclear Reactor Regulation
To:
Shared Package
ML20002A376	List: ML20002A375 ML20002A377 ML20002A378 ML20002A379
References
NUDOCS 8011140676
Download: ML20002A377 (45)

Text

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b UNITED STATES i-

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NUCLEAR REGULATORY COMMISSION (f'

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YANKEE ATOMIC-ELECTRIC COMPANY

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DOCKET NO. 50-29 E

YANKEE NUCLEAR POWER STATION (YANKEE-R0WE)

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AMENDMENT TO FACILITY OPERATING LICENSE

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Amendment No. 43 License No. DPR-3 7

L 1.

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

A.

The application for amendment oy Yankee Atomic Electric Company (the licensee) dated January 6,1977, as supplemented March 11; April 13; May 2; June 30; July 7,14 and 15, Augurt 1, 4, 5, 8, 9 and 22,1977, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confomity with the application, the provisions of the act, and the rules and regulations of

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the Commission, C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common 1

defense and security or to the health and safety of the public; and i.

t E.

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amencment, and paragraph 2.C.(2) of Facility Operating License No. DPR-3 is hereby amended to read as follows:

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"( 2 ) Techniral Soccifications 9 :;.g The Technical Specifications contained in Appendix A,

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as revised through Amendment No. 43, are hereby

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incorporated in the license. The licensee shall operate the facility in accordance with the Technical d

Specifications."

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This license amendment is $fTective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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Karl R. Goller, Assistant Director for Operating Reactors

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Division of Operating Reactors

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Attachment:

Changes to the Technical Specifications Date of Issuance:

August 25, 1977 i

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ATTACHMENT TO LICENSE AMENDMENT NO. 43 F:

FACILITY' LICENSE NO. DPR-3 bh.==

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DOCKET NO. 50-29 j

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i2 2 Revise Appendix A as follows:

Remove the.following pages and insert identically numbered pages:

I E5 2-1 / 2-2 2-3 / 2-4 B 2-1 / B 2-2 i

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3/4 1-23 / 2/4 1-24 3/4 1-25 / 3/4 1-26 3/4 1-27 / 3/4 1-28 3/4 1-29 3/4 2-1 / 3/4 2-2

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3/4 2-3 / 3/4 2-4 3/4 2-5 / 3/4 2-6 3/4'2-7 / 3/4 2-8 3/42-9-/3/42-10

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3/4 2-11 / 3/4 2-12 3/4 2-13.

3/4 5-1 / 3/4 5-2 3/4 5-7 / 3/4 5-8 B 3/4 1-3 / B 3/4 1-4 B 3/4 1-5 B 3/4 2-1 / B 3/4 2-2

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B 3/4 2-3 i

B 3/4 5-1 / B 3/4 5-2 5-1 / 5-2 Revised area is indicated by marginal lines. Overleaf pages are provided 'for convenience.

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i 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

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2.1 SAFETY LIMITS T

REACTOR CORE

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2.1.1 The combination of THERMAL POWER, Main Coolant System pressure, and the

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i limits shown in Figures 2.1-1 and 2.1-2 for 4 and 3 loop operation, respectively.

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APPLICABILITY: MODES 1 and 2.

ACTION

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Whenever the point defined by the combination of the highest operating loop cold leg temperature and THERMAL POWER has exceeded (is above and to the right or) the appropriate Main Coolant System pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MAIN COOLANT SYSTEM PRESSURE 2.1.2 The Main Coolant System pressure shall not exceed 2735 psig.

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

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I-1 MODES I and 2 Whenever the Main Coolant System pressure has exceeded 2735 esig, 1

I be in HOT STANDBY with the Main Coolant System pressure within i

its limit within i hour.

MODES 3, 4 and 5.

T Whenever the Main Coolant System pressure has exceeded 2735 psig, j

reduce the Main Coolant System pressure to within its limit within-5 minutes.

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70 80 90 100 110 120 130 Indicated Reactor Power, Percent REACTOR CORE SAFETY LIMIT - ALL LOOPS IN OPERATION FIGURE 2.1_1 YANKEE-ROWE 2-2 Amendment No. 43 h h.; f,. $,1

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500 50 60 70 80 00 100 110 Indicated isactor Power, Percent REACTOR CORE SAFETY LIMIT - 3 LOOPS IN OPEPATION C

FIGURE 2.1-2 il YANKEE-ROWE 2-3 Amendment,.!c.

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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l-L i:.

2.2 ' LIMITIhG SAFETi CVtTFM SETTINGS

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2.2.1 _The reactor protective system instrumentation setpoints shall be E

set consistent with the Trip Setpoint values shown in -Table 2.2-1.

APPLICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

With a reactor protective system instrumentation trip setpoint less

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conservative than the value shown in the Trip Setpoint column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored y;:..

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to OPERA 3LE status with its trip setpoint adjusted consiste,nt with the

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Trip Setpoint value.

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2.1 SAFETY LIMITS BASES

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The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the main 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 sur-face temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures becat ie of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and main coolant temperature

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and pressure have been related to DNB through the W-3 correlation. The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio.of the heat flux that would cause ONB at a particular. core lccation to the local heat flux, is indicative of the margin t'o'DNB'.

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The minimum vc.lue of the ONBR 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 con-fidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Main Coolant System pressure and cold leg temperaturg.for which the minimum DNER is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

B~ecause of flow instability, DNS may occur prematurely should the core exit quality become too great. The limiting core exit quality for preventing flow instability is taken conservatively as 0.08.

The limiting hot channel factors used 1n determining the thermal limit curves are higher than those calculated at full power for the range from all control rods fully Mithdrawn to maximum allowable control rod insertion.

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SAFETY LIMITS

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afial power sh$pe.1.80; and a reference cosine with a peak of 1.44 for F of 2.76; F, of

' These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion.

2.1.2 MAIN COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Main Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the main coolant from reaching the containment atmosphere.

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The reactor pressure vessel, pressurizer and pumps are designed to Section VIII of the ASME Boiler and Pressure Vessel Code for Nuclear Power Plant, including all addenda through 1956, which permits a maximum transient pressure of 110%, 2735 psig, of design pressure.

Pressure relief devices must be provided that will prevent pressure from exceeding 110 percent o' the design pressure.

The Main Coolant System piping and valves are designed to ANSI (formerly ASA) Standards, Power Piping Code, Section 531.1, 1955 Edition, and B16.5,1957 Edition, respectiv.'y, which allows the design to be based on normal operating pressure and tempera ure and also allows exceeding the design conditions for periods of time. The stress level can be increased 15 percent above~the Code l

allowable design value for not more than 10 percent of the design life and uo to 20 percent above the allowable for up to 1 percent of the design life. Since normal plant operating pressure is 2000 psig, there

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is no c:nflict with either design condition. The setting of the Main Coolant System safety valves could allow pressure to increase to 2560 psig during a transieat. The amount of time this condition is expected te exist is well witnin the allowances of B31.1. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code recuirements.

The entire Main Coola-. System was hydrotested at 3435 psig,138%

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

YANKEE.;;WE B 2-2 Amendment No. 43

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REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE C0h7ROL RODS CONTROL ROD OPERABILITY

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LIMITING CONDITION FOR OPERATION

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M:1 3.1.3.1 All control rods which are inserted in the core shall be OPERABLE and positioned within + 8 inches (indicated position) of every other rod in their group.

APPLICABILITY: MODES 1* and 2*

ACTION:

With one or more control rods inoperable due to being immovable a.

as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

With more than one control rod inoperable or misaligned from b.

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any other rod f('its ' group by more than + 8 inches (indicated position), be in at least HOT STANDB) within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, With one control rod inoperable or misalianed from any other c.

rod in its' group by more than 2,8 inches (indicated position-),

POWER OPERATION may continue provided that within one hour eitner:

1.

The rod is restored to OPERABLE status within the above alignment requirements, or 2.

The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

POWER s OPERATION may then continue provided that:

a)

An analysis of the potential ejected rod worth i

is pefferred withtm 3 days and the rod worth is determired to be < 0.88% ao at zero power and

< 0:5% 4o at RATED THERMAL POWER for the remainder of the fuel cycle, and

'See Special Test Exceptions 3.10.2 and 3.10.4 YANKEE-ROWE 3/4 1-23 Amendment No. 44 43 l

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REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) b)

The SHUTDOWN MARGIN requirement of. Specification

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3.1.1.1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and e=.=

c)

A power distribution map is obtained from the movable incore detectors and F and F are verified to be within their likits wikNin-72 hours.

d)

The THERMAL POWER level is reduced to < 75% of THER*AL POWER allowable for the Main Coolant pump combination within one hour and within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the Power Range and Intermediate Power Range Neutron Flux high trip setpoint b reduced to < 108% of the 75% of allowable THERMAL POWER, or e)

The remainder of the rods in the group with the inoperable rod are aligned to within + 8 inches

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of the inoperable rod within one hour while maintaining the rod sequence and insertion limits of Figures 3.1-1 and 3.1-2; the THERMAL POWER level shall be restricted pursuant.to Specification 3.1.3.5 during subsequent operation.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each control rod shall be detemined to be within the limit by verifying the individual tod positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.1.3.1.2 Each control rod not fully inserted shall be determined to be OPERABLE by movement of at least 4 inches in any one direction at least once per 31 days.

4.1.3.1.3 The maximum reactivity insertion rate due to withdrawal of

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the highest worth control rod group shall be detemined not to exceed 1.5 x 10' ak/k per second at least once per 18 months.

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REACTIVITY CONTROL SYSTEMS POSITION IN'DICATOR CHANNELS LIMITING CONDITION FOR OPERATION 3.1.3.2 All control rod primary and secondary position indicator channeis shall be OPERAELE and capable of determining the control rod positions within + 3 inches.

APPLICABILITY: MODES 1 and 2.

ACTION:

With a maximum of one primary rod position indicator channel a.

per group inoperable either:

1.

Determine the position of the non-indicating rod (s) in-directly by the movable incore detectors at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and immediately after any motion of the non-indicating rod which exceeds 8 inches in one direction since the last determination of the rod's position, or 2.

Reduce THERMAL' POWER to < 50% of THERMAL. POWER allowable

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for the main coolant pump combination within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, b.

With a maximum of one secondary position indicater per group inoperable either:

1.

Verify that all primary rod position.idica'.:rs for the affected group are OPERABLE, or 2.

Reduce THERMAL POWER to < 50% of THERMAL POWER allowable for the main coolant pump combination within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1. 3. 2 Each rod position indicator chann'el shall be determined to be OPERABLE by verifying the primary position indication system and the secondary position indicator channels agree within 3 inches at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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ROD DROP TIME R;

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E=i LIMITING CONDITION FOR OPERATION

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3.1.3.3 The individual control rod drop time from the fully withdrawn r==

position shall be < 2.5 seconds from loss of stationary gripper coil

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voltage to 6-inch coil entry with:

T,yg >,.515'F, and l

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All main coolant pumps operating.

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APPLICABILITY: MODES 1 and 2 n2=

1 ACTION:

Wi:.h the drop time of any control rod determined to exceed

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4 the above limit, restore the rod drop time to wi' thin the above

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3 b.

With the rod drop times within limits but determined with 3 main coolant pwr.ps operating, operation may proceed provided THERMAL POWER is restricted to < 75% of RATED THERMAL POWER.

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The provisions of Specification 3.0.4 are not applicable.

i SURVEILLANCE RE0VIREMENTS

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..hu 4.1.3.3 The rod drop time of control rods shall be demonstrated through li measurement prior to reactor criticality:

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For all rods following each removal of the reactor vessel

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For specifically affected individual rods following any main-

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which would affect the drop time of those specific rods, and

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At least once 'p'er 18 months.

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REACTIVITY CONTROL SYSTEMS SHUTDOWN RCD INSERTION LIM'T

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3.1.3.4 All shutdown rods : hall be withdrawn to at least 87 inches.

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APPLICABILITY: MODES 1* and 2*#-

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ACTION:

With a maximum of one shutdown rod not withdrawn to within the limit.

except for surveillance testing pursuant to Specification 4.1.3.1.2,

-ee within one hour either:

a.

Withdraw the rod to within the limit, or

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' Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE REOUIREMENTS

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" 1.3.4 Each shutdown rod shall be determined to be withdrawn to within the limit:

a.

Within 15 minutes prior to withdrawal of any rods in regulating groups C and A during an, approach to reactor criticality, l

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Atleastonceper4hou'rsthereafter.

• See Special Test Exceptions 3.10.2 and 3.10.4 iWi th K,f f > 1.0.

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CONTROL ROD INSERTION LIMITS'

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LIMITING CONDITION FOR'0PERATION

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3.1.3.5 The control groups shall be limited in physical insertion as

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shown in Figure 3.1-1.

APPLICABILITY: MODES 1* and 2*!.

ACTION:

With the control groups inserted beyond the above insertion limits.

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except for surveillance testing pursuant to Specification 4.1.3.1.2, _

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either:

a.

Restore the control groups to within the limits within

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two hours,.or 5

b.

-Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the x.

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group position using the above figure, or

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c.

Se in at '! east HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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SURVEILLANCE RE0VIREMENTS

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4' 4.1.3.5 The position of each control group shall be determined to be

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within the insertion limits at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

"See Special Test Exceotions 3.10.2 and 3.10.4

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Group A J

Group C I

Control Rod Position, Inches Withdrawn t

' Allowable THERMAL Pcwer based on the main coolant _oump combination in operation.

FIGURE 3.1 1 YANKEE-ROWE _

3/4 1-29 Amendment No. 43

(

~

=.=--

POWER DIST'RIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued) 4.2.1.2 The below factors shall be included in the calculation of peak full power LHGR:

Heat flux power peaking factor, F"q, measured using incore a.

instrumentation at a power > 10%.

i b.

Effect of inserting the control group from its position at the as shown in time of measurement to its insertion limit, F 7 Figure 3.2-2.

The rod insertion limit is shown in Figure 3.1-1.

The multiplier for xenon redistribution is :s function of core c.

lifetime as given in Figure 3.2-3.

In addition, if.ontrol rod Group C is inserted below 75 inches, allowable power may not be l regained until power has been at a ' educed level defined below for at least twenty four hours w control rod Group C l

between 75 and 90 inches.

Reduced power = allowable fraction of full power times

' multiplier given in Figure 3.2-4.

Exceptions:

1.

If the rods are inserted below 75 inches and power does not go below the reduced power calculated above, hold at the lowest attained power level for at least twenty four hours with control rod Group C between l 75 and 90 inches before returning to allowa'ble power.

2.

If the rods are inserted below 75 inches and zero power is held for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, no reduced power level need be held on the way to the allowable fraction of full power.

d.

Shortened stack height factor,1.009.

e.

Measurement uncertainty, 1.05.

f.

Power level uncertainty,1.03.

l g.

Heat flux engineering factor, F,1.04, q

b.

Core average. linear heat generation rate at full power, l

I 4.40 kw/ft.

YANKEE-ROWE 3/4 2-2 Amendment No. 43

=

.7

+.

y m

.. ~,

,p.

p y

.-,.-,,y

_~

(~

4 W

o

[

3/4.2 POWER DISTRIBUTION LIMITS I

PEAX LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 1

l 3.2.1 The peak linear heat generation rate (LHGR).shall not exceed the

~

limits of Figure 3.2-1 during steady state operation for the first 225 equivalent full power days (EFPD's) operation in Cycle XIII.*

APPLICABILITY: MODE 1 l

ACTION:

With the peak LHGR exceeding the limits of Figure 3.2-1; i

i a.

Within 15 minutes reduce THERMAL POWER to not more than that f

fraction of the THERMAL POWER allowsble, for each fuel type

~'

and for the main coolant pump combination in operation, as ex-pressed below:

Fraction of THERMAL POWER =

Limitino LHGR i

Peak Full Power LHGR

=

i j

b.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> reduce the Power Range and Intermediate Power Range Neutron Flux high trip setpoint to < 108", of the fraction of THERMAL POWER allowable for the main coolant pump combination.

SURVEILLANCE REOUIREMENTS i

i 4.2.1.1 The peak LHGR shall be determined to be witnin the limits of i

Figure 3.2-1 using incore instrumentation to obtain a power distribution map:

a.

Prior to initial operation above 75% of RATED THERMAL POWER after each fuel loading, and t

b.

At least once per 1,000 EFPH.

l c.

The provisions of Specification 4.0.4 are not applicable.

,.,m.

I

' Operation af ter 225 EFPD's into Cycle XIII may proceed only after a revised

~

Figure 3.2-1 based on appropriate LOCA analyses has been approved by the c

NRC. Operation in the 3-Loop mode. is not permitted until appropriate LOC

analyses 'for this mode have been approved by the NRC.

YANKEE-ROWE 3/4 2-1 Arendment No. 43 i

.m

=

POWER DISTRIBUTION LIMITS

=

SURVEILLANCE REQUIREMENTS (Continued) 4.2.1.2 The below factors shall be included in the calculation of peak full power LHGR:

Heat flux power peaking factor, F, measured using incore a.

instrumentation at a power > 10%.

b.

Effect of inserting the control group from its position at the as shown in time of measurement to its insertion limit, F y

Figure 3.2-2.

The rod insertion limit is shown in Figure 3.1-3 1.

The multiplier for xenon redistribution is a function of core c.

lifetime as given in Figure 3.2-3.

In addition, if control rod Group C is inserted below 75 inches, allowable power may not be l regained until power has been at a reduced level defined below for at least twenty four hours with control rod Group C l

between 75 and 90 inches.

Reduced power = allowable fraction of full power times multiplier given in Figure 3.2-4 Exceptions:

1.

If the rods are inserted below 75 inches and power does not go below the reduced power calculated above, hold at the lowest attained power level for at least twenty four hours with control rod Group C between l j

75 and 90 inches before returning to allowa~ble power.

2.

If the rods are inserted below 75 inches and zero power is held for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, no reduced power level need be held on the way to the allowable fraction of full power.

d.

Shortened stack height factor,1.009.

Measurement uncertainty,1.05.

=

e.

f.

Power level uncertainty,1.03.

e g.

Heat flux engineering factor, F",1.04.

q h.

Core average linear heat generation rate at full power, l

4.40 kw/ft.

YANKEE-ROWE 3/4 2-2 Arendment No. 43 e

w, n

y n

r POWER DISTRIBUTION LIMITS

~

SURVEILLANCE REOUIREMENTS (Continued) 4.2.1.3 At least once per 1000 EFPH the following limits shall be determined by calculation not to be exceeded at RATED THERMAL POWER:

Hottest channel exit coolant temperature < 602 F, and l

a.

b.

Maximum clad surface temperature in hottest channel < 637'F.

l

Z..

YANKEE-RC'4E 3/4 2-3 Amendment.o.

43

.__ ___.... _ _ _. _. _ =.....

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M 30 Fresh and Exposed Exxon Fuel

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P Core XIII Fuel Exposure (EFPD) k FIGURE 3.2-1

+w Core XIII tilowable Peak Rod LHGR Versus Exposure i

i

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GROUP c ROD POSITION, INCHES WITHDRAWN v n A..w..... C s

v y i 1

F G Measure::. cat FIGURE 3.2-2 Factor F as a Function of Rod Insertion YANKEE-ROWE 3/4 2-5 Amendment No. 43 h

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43 90RmiIGlNAl

b 2""

C :

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aamod pa:npag sof Jat[d p tnw YA N KE E-P C'a*E 3/4 2-7 knendment No. XI43 PMR ORIGINAL

..w-.,

c

+,

g

_fi ~

g.;

f

=

ppr r 2

L+

^J','.'

POWER DISTRIE',3,44 LIMITS is HEAT FLUX HOT CHK.NEL FACTOR-F

= :

q

_ LIMITING CONDITION FOR OPERATION 4

1

)

3.2.2 F shall be limited by the folicwing relationships:

q 4 ~ [2.76] for P > 0 3 F

P 1

i F

< [5.52] for P,$ 0.5 l

q where P = THERMAL POWER "4

RATED THERMAL POWER 1

APPLICABILITY: MODE 1 4

ACTION:

With F exceeding its limit:

q Reduce THERFAL POWER at least If. for each 1% F exceeds the a.

limit within 15 minutes and similiarly reduce Ehe Power Range

..) Intermediate Powei *ange Nuetron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

e b.

Identify and correct the cause of the out of limit condition i

prior to increasing THERFAL POWER above the reduced limit required by a, above; THERMAL POWER may then be increased 4

provided F is demonstrated through incere mapping to be 4

witnin its 11mit.

)

+ -

I L

t YANKEE.ROWE 3/4 2-8 Arendment flo. 43 l::

,,a-

~

=:-

POWER DISTRIBUTION LIMITS

\\

==?4

a

'FE SURVEILLANCE REOUIREMENTS

. l=(;

.*.".h".'.

4.2.2.1 F shall be determined to be within its limit by:

q a.

Using the n...:..le incere detectors to obtain a power distribution map:

l.

Prior to initial operation above 75% of RATED THERMAL

,e POWER after each fuel loading, and 2.

At least once per 1000 Effective Full Power Hours, b.

Increasing the measured F component of the power distri-bution map by 4% to accou8t for engineering tolerances, l

increasing the value by 5% to account for measurement un-l certainties, and further increasing the value by 3% to account for fuel densification.

3 c.

The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 When F is measured pursuant to Specification 4.10.2.2, an over-all measured F 9shall be obtained from a power distribution map and in-creased by 4% fo account for engineering tolerances, increased by 5% to a

account for measurement uncertainity and further increased by 3% to account for fuel de~nsification.

l

l.,

5 YANKEE-ROWE

- 3/42-9 g

.

~

.,.~,

- - - i

L; l.-

'."-l A l

POWER DISTRIBUTION LIMITS NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - F H LIMITINC CONDITION FOR OPERATION

l, 3.2.3 F

shall be limited by the following relationsnip:

H F

< l.80 [1.0 + 0.2 (1-P)]

l H

where P = THERMAL POWERRATED THERMAL POWER, and F = 1.86 for Gulf United fuel and 1.75 for EXXON Nuclear fuel APPLICABILITY: MODE 1

~

ACTION.

With F exceeding its limit:

t.H l

a.

Reduce THERMAL POWER to less than 50% of RATED THERFAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range and Intermediate Power Range Neutron Flux-high trip setpoints to < 55% of RATED THERFAL POWER within the next a hours, Demonstrate through in-core mapping that F"Ymit or reduce is within its b.

limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the t THERFAL POWER to less than 5% of RATED THEPFAL POWER withi."

the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and c.

Identify and correct the cause of the out of limit condition prior to increasing THERFAL POWER above the reduced limit required by a or b, abovq; subsequent POWER OPERATION may proceed provided that F',

is demonstrated through in-core mapping to be within its II..it at a nominal 50% of RATED THERFAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERFAL POWER prior to exceeding this THERFAL power and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERMAL POWER.

=

YANKEE-ROWE 3/4 2-10 Amendment No. 43

~^

'n POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS 4.2.3.1 F,I shall be determined to be within its limit by using the mov-I g

able incore detectors to obtain a power distribution map:

Prior to operation above 75% of RATED THERMAL POWER after each a.

fuel loading, and b.

At least once per 1000 Effective Full Power Hours.

The provisions of Specification 4.0.4 are not applicable.

c.

I The measured F.

of 4.2.3.1 above shall be increased b; 5%

4.2.3.2 for measurement uncertaTNty.

YANKEE-ROWE 3/4 2-11 Amendment No. 43

l L.

f.h, I

POWER DISTRIBUTION LIMITS DNB PARAMETERS E

^:if:,.

LIMITING CONDITION FOR OPERATION

1. 3..

3.2.4 The following DNS related parameters shall be maintained within the limits shown on Table 3.2-1:

=

a.

Main Coolant System Inlet Temperature.

i..

b.

Main Coolant System Pressure c.

Main Coolant System Total Flow Rate APPLICABILITY: MODE 1 ACTION:

... g

=

With any of the above parameters exceeding its limit, restore the param-eter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

q=

_ SURVEILLANCE REOUIREMENTS 4.2.4.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.4.2 The Main Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 months.

~=:

=

YANKEE-ROWE 3/4 2-12 v

-y

.J TABLE 3.2-1 E

DilB PARAMETERS M

I, LIMITS E

4 Loops in 3 Loops in PARAMETER Operation Operation Main Coolant System Inlet Temperature

< 515"F

< 515"F Main Coolant System Pressure

> 1950 psig*

> 1950 psig*

Main Coolant System Total 6

6 Flow Rate

>38.3 x 10 lb/hr

>29.9 x 10 lb/hr M

O 4

Y U

• Limit not applicable during elther a TilERMAL POWER ramp increase in excess of 5% RATED TilERMAL POWER per minute or a TilERMAL l'0WER step increase in excess of 10% RATED islERML POWER.

g.

II ea i

k I

s, 4

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=e; 3/4.5 EMERGENCY CORE COOLING SYSTEMe. (ECCS)

ACCUMULATOR

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LIMITING CONDITION FOR OPERATION t

3.5.1 The low pressure safety injection accumulator shal. Se OPERABLE with:

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a.

Isolation valves SI-MOV-1 and SI-TV-608 open, b.

A minimum useable contained borated. vater volume of 700 cubic feet of borated water, equivalent to an indicated level of f

261" in the accumulator.

l c.

A minimum baron concentration of 2200 PPM, d.

An accumulator nitrogen cover-pressure of less tha,715 l

1 psig.

The nitrogen supply system with three supply pressure regola cing e.

valves set at 473+ 10 psig and at least:

l 1.

Sixteen' 48' cubic foot nitrogen bottles 3,1390 psig, or 2.

Seventeen 48 cubic foot nitrogen bottles > 1340 psig, or 3.

Eighteen 48 cubic foot nitrogen bottles 3,1294 psig, f.

Two OPERABLE low level venting system, and g.

Timers set to cperate between 6.4 and 7.1 seconds.

l APPLICABILITY: MODES 1, 2, 3* 4* and 5*

ACTION:

a.

With the accumulator inoperable, except as a result of a closed isolation valve or as a result of one Tnoperable l

pressure regulating valve or one inoperable low level venting system, restore the inoperable accumulator to OPERABLE status within.15 minutes or be in at least HOT SHUTDOWN with main coolant pressure < 1000 psig within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. -

4 b.

With the accumulator inoperable due to one isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within one hour and be in at least

=

HOT SHUTDOWN with main coolant pressure < 1000 psig within the

-next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

' Main cociant pressure 3,1000 psig.

YANKEE-ROWE 3/4 5-1 Amendment No. 43 w

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g EMERGENCY CORE COOLING SYSTEMS p

LIMITING CONDITION FOR OPERATION (Continued)

With one inoperable supply pressure regulating valve or with one c.

inoperable low level venting system, restore the inoperable regulating valve or venting system to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT STANDBY within one hour and be in at least HOT SHUTDOWN witP main coolant pressure < 1000 psig within the following 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; SURVEILLANCE RE0VIREMENTS 4.5.1 The accumulator shall be demonstrated OPERABLE:

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

1.

Verifying the contained borated water volume, accumulator nitrogencover-pressure and the supp1v pressure-regulator discharge pressure, and 2.

Verifying that accumulator isolation valves SI-MOV-1 and SI-TV-608 are open.

b.

At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase of > 1% by verifying the boron concentration of the accumulator solution, At least once per 31 days by verifying the nitrogen supply c.

bottle pressures.

d.

At least once per 18 months during shutdown by:

1.

Verifying GPERABILITY of the nitrogen supply system by observing operation of each regulating valve.

l 2.

Verifying OPERABILITY of each low level venting system when the level switch column water level is lowered; at the same time, verify closure of SI-MOV-1 and SI-TV-608, 3.

Verify actuation of each accumulator time delay relay.

The acceptable rundown time is 6.75 + 0.34 seconds.

4 Verify that SI-TV-604, SI-TV-605 and SI-TV-606 open when NS-S0V-46 is energized and again.when NS-50V-47 is energized.

l YANKEE-ROWE 3/4 5-2 Amendment No. 43 e

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EMERGENCY CORE COOLING SYSTEMS

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SURVEILLANCE REQUIREMENTS (Continued) 7.

Verifying that ECCS recirculation subsystem each pair of i

redundant valves and purification pumps are aligned to

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receive electrical power from separate OPERABLE busses.

8.

Verifying that each ECCS long term hot leg injection sub-system charging pump is aligned to receive electrical power from an OPERABLE bus.

j 9.

Verifying that the long term hot leg injection flow metering instrument is OPERABLE by observing charging il flow rate at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

3 By a visual inspection which verifies that no loose debris c.

(rags, trash, clothing, etc.) is present in the containment

.I wF'ch could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This 4

visual inspection shall be performed:

1.

For all accessible areas of the containment prior to establishing containment integrity, and 2.

Of the areas affected within containment at the comple-tion of each containment entry when containment -integrity i

is established.

4 j

d.

At least once per 18 months by visual inspection of the con-l tainment sump and verifying that the subsystem suction inlets i

are not restricted by debris and that the sump components

'(trash racks, screens, etc.) show no evidence of structural distress or corrosion.

}

At least once per 18 months, during shutdown, by:

e.

i 1.

Cycling each power operated (excluding automt.ic) valve i

in the flow path through at least one ;omplete cycle of full travel.

2.

Verifying that valve CS-MOV-532 actuates to its correct j

position on a safety injection signal.

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EMERGENCY CORE COOLING SYSTEMS t-SURVEILLANCE REOUIREMENTS (Continued) 3.

Verifying that each of the following pumps start auto-matically upon receipt of a safety injection signal:

a)

High pressure safety injection (HPSI) pump b)

Low pressure safety injection (LPSI) pump 4

Verify the proper positioning of the HPSI throttle valves SI-V-571, SI-V-672, SI-V-673 and SI-V-674 at the first COLD SHUTDC'JN following Cycle XIII startup and at least once per 36 months thereafter by developing a back pressure of 875 psig in the high pressure safety injection header witn two HPSI pumps operating as follows:

a)

Pressure to the sucticn of the HPSI ptmps to be 170+,

10 psi.

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b)

LPSI flow is isolated.

c)

Injection flow is to one loop with the other icops isolated by closing the appropriate injection gate valves CS-MOV-536, CS-MOV-537, CS-MOV-538, and CS-MOV S39.

a d)

The flow to the injection icops shall not be less than 200 gpm.

e)

The above test shall be repeated to include the opera-tion of all HPSI pumps.

5.

Verifying that two Icw pressure safety injection pumps develop a combined flow > 2180 cpm. Test every LPSI pump at least once per 36 months.

6.

Verifying that each charging pump stops automatically upon receipt of a safety injection signal.

7.

Verifying that the long term hot leg injection ficw meter-ing instrument is OPERABLE by performing a CHANNEL CALIBRATION.

S.

Verifying that each valve listed in Specification 4.5.2.b.*

is in its normally open position.

YANKEE-RC'4E 3/4 5-8 Amendment No. 43

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3/4.1 REACTIVITY CONTROL SYSTEMS BASES

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3/4.1.2 BORATION SYSTEMS The baron injection system ensures that negative reactivity control is.available durita each mode of _ facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths from the borated water sources to the charging pumps, 4) boric acid gravity flow connection, 5) associated heat tracing systems, and 6) power supply from OPERABLE 480 volt busses.

flow paths.

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

The boration capability is sufficient to provids a SHUTDOWN MARGIN from all operating conditions of 5.0% ak/k after xenon decay and cooldown to 200*F. The maximum boration capability requirement occurs at EOL from fall power equilibrium xenon conditions and requires 776 gallons of 22,000 ppm borated water from the boric acid mix tank or 9192 gallens of 2200 ppm borated water from the safety injection tank.

With the Main Coolant System temperature below 200*F, except during refueling, one charging pump is acceptable without single failure con-sideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting positive reactivity change in the event the single charging pump becomes inoperable.

The boren capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 7% ak/k after xenon decay and cooldown from 200*F

t. 140*F. This condition requires either 1059 gallons of 22,000 ppm borated water from the boric acid mix tank or 13,218 gallons of 2200 ppm borated water from the safety injection tank.

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-4 3/4.1 REAC'fVITY CONTROL SYSTEMS BASES d

3/4.1.3 MOVABLE CONTROL RODS The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN PARGIN is maintained. (3) the individual control rod worth and hot channel factors used for the accident analyses are not exceeded, and (4) limit the potential effects of a rod ejection accident. OPERABILITY of the con-trol rod position indicators is required to determine control rod pos-1 itions and thereby ensure compliance with the control rod alignment and insertion limits.

For control rod misalignment up to 8 inches from every other control rod in its group and for all control rod groups within their m.

insertion limits, the hot channel factors will be well within the design limits of:

Fq1 2.76 H 1 1.80 F

The limit applies to three loop operation, in which case the power

cordinate is rescaled to 100% of rated three loop power.

This ensures that the induced peaking will not lead to worse thermal conditions than for four loop operation since the flow to power ratio is greater for three loop operation.

If a control rod is misaligned the hot channel factors, potential ejected rod worth, and SHUTDOWN MARGIN will be s.howrr to be within design limits and reactor power will be reduced. The requirements that no more than one inoperable control rod is allowed and that the shutdown nargin is maintained ensures that th'e reactor can be brought to a safe shutdown condition at any time.,

The ACTION statements which permit limited variations from the aaricrequirementsareaccomWn'fedbyadditionalrestrictiorswhich ensure that the original criter'.., are met. Misalignment of a rod re-quires ejected rod worth and SHUTDOWN MARGIN to be within limits and a restriction in THERMAL p0WER; either of these restrictions provide 1ANKEE-ROWE B 3/4 1-4

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REACTIVITY CONTROL SYSTEMS

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BASES

?/4.1.3 MOVABLE CONTROL RODS (Continued) assurance of fuel rod integrity during continued operation.

The reac-tre tivity worth of a misaligned rod is limited for the remainder of the i

fuel cycle to prevent exceeding the assumptions used in the accident

analysis, i

The maximum rod drop time restriction is consistent with the as-sumed red drop time used in the accident analyses. Measurement with i

T,y t 515'F and with all main coolant pumps operating ensures that the l

mea lured drop times will be representative of insertion times exper-ienced during a reactor trip at operating conditions.

Control rod positions and ">PERABILITY of the rod position indica-tors are required to be verified on a nominal basis of once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

These verification frequencies are adequate for assuring that the ap-plicable LCO's are satisfied.

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ANKET.-ROWE B 3/4 1-5 Amendm(qt No. 43 Y

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3/4.2 POWER DISTRIBUTION LIMITS BASES The s m if'. cations of this section provide assurance of fuel integrity cariag Condition I (Normal Operation) and II (Incidents of Moderate Fres;. y) events by:

(a) maintaining the minimum DNBR in the y

core > 1.30 during normal operation and in short term transients, and (b) ITmiting the fission gas release, fuel pellet temperature & cladding mechanical properties to within assumed design criteria.

3/4.2.1 PEAK LINEAR HEAT GENERATION RATE Limiting the peak linear heat generation rate (LHGR) during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS accepcance criteria limit of 2200'F is not exceeded.

When operating at constant power, all rods out, with equilibrium xenon, power peaking in the Yankee Rowe core decreases monotonically as a function of cycle burnup.

This has been verified by both calculation and measurement on Yankee cores and is in accord with the expected j

behavior in a core that does not contain burnable poison. The all-rods-out power peaking measured prior to exceeding 75% of RATED THERMAL P0riER af ter each fuel loading thus provides an upper bound on all-rods-cut power peaking for the remainder of that cycle. Thereaf ter the measured power peaking shall be checked every 1000 equivalent full power hours and the latest measured value shall be used in the computation. The only effects which can increase peaking beyond this value would be control red inser 'on and xenon transients and these are accounted for

~

in calcuisting i;ak thGR.

The core is stable with respect to xenon, and any xenon transi which may be excited are rapidly damped.

The xenon multiplier in Figure 3.2-3 was selected to conservatively account for transients which can result from control rod motion at full power.

The limits on power level and control rod position following control rod insertion were selected to prevent exceeding the maximu, allowable linear heat generation rate limits in Figure 3.2-1 within t% first few hours following return to power after the insertion.

With Yankee's highly damped core, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> hold allows sufficient time for the initial xenon maldistribution to acco:nodate itself to the new power distribution.

The restriction on controi rod location during these 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the return to allowable fraction of full power will not cause additional redistribution due to rod motion.

YANKEE-ROWE B 3/4 2-1

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3.l 3/4.2 POWER' DISTRIBUTION LIMITS BASES (Continued)

After 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> at zero power, the average xenon concentration has decayed to about 20% of the full power concentration.

Since the xenon concentrations are so low, an increase in power directly to maximum allowable power creates transient peaking well below the value imposed by the xenon redistribution multiplier. Thus, any increase in power peaking due to this operation is below the value accounted for in the calculation of the LHGR.

These conclusions are based on plant tests and on calculations per-formed with the SIMULATE three dimensional nodal code used in the analysis of Core XI (reference cycle) described in proposed Change No.115, dated March 29,1974.

'~

3/4.2.2 and 3/4.2.3 HEAT r' LUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux and ethalpy hot channel factors ensure that

1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limi,t.

Each of these hot channel factors are measurable but will normally only be determined periodically as specified in Specification 4.2.2.1 a'nd 4. 2. 3.1.

This periodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided:

a.

Control rods in a single group move together with no individual rod insertion differing by more than _+ 8 inches from any other rod in the group.

b.

Control rod groups are secuenced with overlapping groups as described in Specification 3.1.3.5.

The control rod insertion limits of Specification 3.1.3.5 is c.

maintained.

N The relaxation in F as a function of THERMAL POWER allows changes iqtheradialpowershaphgfor all permissible rod insertion limits.

F will be maintained with its limits provided conditions a thru c aH a5cve, are maintained.

When an F measurement is taken, experimental error, engineering tolerance and fuel densification must be allowed for.

5% is the appropriate allowant.e for a fuil core map taken with the incore detec*.or flux mapping system. 4% is the appropriate allowance for engineering tolerance, and 3% is the appropriate allowance for fuel densification.

YANKEE-ROWE B 3/4 2-2 Amendment No. 43 l

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3/4.2 p0WER DISTRIBUTION LIMITS BASES (Continued)

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-When F is measured, experimental error must be allowed for and 5%

'=i is the apprchriate allowance for'a full core. gap taken with the incore 1

detection system. The specified limit for F; also contains a 6%

alloy lance for uncertainties which mean that n$rmal operation will in F H

• 1.80.

The 6% allowance is based on the following considerations:

l f

f Abnormal perturbation in tge radial power shape, such as from a.

rod misalighment, effect F more directly than F.,

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b.

Although rod movement has a direct influence upon limiting F to withi,n its limits, such control is not readily available Eo l

limit F'3g, and l

c.

Errors in prediction for cntrol power shape detected during startup physics tests can be compensated #or in F by restric t-g ing axial flux distributions. This compensation for F is

_s aH less readily available.

3/4.2.4 DNB PARAMETERS The limits on the DNS related parameters assure that each of the t

parameters are maintained within the normal steady state envelope of operation assumed in the transient & accident analyses. The limits are consistent with the accident analysis assumptions and have been an-alytically demonstrated adequate to maintain a minimtm DNBR of 1.30 throughout each analyzed transient. The Main Coolant System inlet temperature assumed in the analyses is 519'F, conservatively 4'F in l

excess of the limit to allow for uncertainty in plant measurement. The Main Coolant System pressure assumed in the analysis is 1925 psig, conservatively 25 psig less than the limit.to allow for uncertainty in plant measurement. The assumed operating deadband of + 50 psig is applied to the nominal 2000 psig limit, yielding a minimum operational limit of 1950 psig.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through i

instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes & other expected transient operation. The 18 month periodic measurement of the Main

=-

i Coolant System total flow rate is ade,uate to detect flow degradation &

.. 1 ensure correlation of the flow indica + 'on channels with measured flow

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such that the indicated percent flow will provide sufficient terifi-cation of flow rate on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

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YANKEE-ROWE-B 3/4 2-3 Amendment No. 43 p

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3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATOR The OPERABILITY of the acuumulator ensures that a sufficient volume of

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borated water will be immediately forced into the reactor core through each of the cold legs in the event the Main Coolant System pressure falls below the pressure of the accumulator. This initial surge of water into the core provides the initial cooling mechanism during large Main Coolant System pipe ruptures.

The limits on accumulator volume, baron concentration and pressure ensure that the assumptions used for accumulator injection in the acci-dent analysis are met. A minimum useable water volume of 700 cubic feet require accumulator water volume to be at least 850 cubic feet.

ine accumulator power operated i:alation valve fails to mee,t single failure criteria and removal of power to the valve is required.

The limits for operation with the accumulator inoperable for any reason except an isolation valve closed or pressurization system inoperable minimizes the time exp~osure.of the plant to a LOCA event occurring concurrently which may result in unacceptable peak cladding tempera-tures.

If a closed isolation valve cannot be immediately opened, the full capability of the accumulator is not available and prompt action is required to piace the reactor in a MODE where this capability is not I

required.

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of three independent ECCS safety injection subsystems, the recirculation subsystem, and the long term hot leg injection sub-system ensur9s that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss o'f one safety in-

]i jection subsystem, one purification pump and one fixed speed charging pump through any single failure cons.ideration. Two safety injection subsystems operating in conjunction with the accumulator tre capable of supplying sufficient core cooling to limit the peak cladding temper-atures within acceptable limits for all postulated break sizes ranging from the double ende'd break of.t,he largest Main Coolant System cold leg pipe downward.

In addition,'the recirculation and long term hot leg injection subsystems provide long term core cooling and boron mixing capability during the accident recovery period.

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EMERGENCY CORE COOLING SYSTEMS BASES

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,ECCS SUBSYSTEMS (Continued)

With the Main Coolant System temperature and pmssure below 330*F, and 1000 psig, respectively, one OPERABLE ECCS safety injection subsystem, a recirculation subsystem and a long tenn hot leg injection subsystem with only one pump per subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor, the decreased probability of a LOCA and the limited core

+

cooling requirements because of the negligible energy stored in the primary coc,lant under these conditions.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that, at a minimum, the assumptions used ir. the safety analyses are met and that subsystem OPERABILITY is maintained.

Complete system tests cannot be performed when the reactor is operating because of their inter-relation with operating systems. The method of assuring operability of these systems is a combination of complete system tests performed during refueling shutdowns and monthly tests of active system components (pumps and valves) during reactor operation.

The test interval is based on the judgement that more frequent testing would not significantly increase reliability.

The subsystems power operated valves fail to meet single failure criteria and removal of power to the valves is required.

3/4.5.4 SAFETY INJECTION TANK The OPERABILITY of the Safety Injection Tank (SIT) as part of the ECCS ensures that a sufficient supply of borated water is available for injectirn by the ECCS in the event of a LOCA.

The limits on SIT minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation co~oling flow to the core, and

2) the reactor will remain subcritical in the cold condition following mixing of the SIT and the Main Coolant System water volumes with all control rods inserced except for the most reactive control assembly.

These assumptions are consistent with the LOCA analyses, which is based on allowing a minimum of 77,000 gallons to be injected by the safety injection subsystems before the recirculation is manually established.

LOCA analyses show that an injection of 77,000 gallons is sufficient to limit core temperatures and containment pressure for the full spectrum of pipe ruptures.

This leaves up to 40,000 gallons in the SIT as reserve.

The boron concentration of 2200 ppm is the highest value assumed in any

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YANKEE-ROWE B 3/4 5-2 H

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i 5.0 DESIGN FEATURES

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EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1-1.

LOW POPULATION ZONE 5.1. 2 The low population zone shall be as shown in Figure 5.1-2.

1 5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel spherical shell having the following design features:

a.

Nominal inside diameter = 125 feet.

b.

Minimum thickness of steel shell

• 7/8 inches, c.

Net free volume = 850,000 cubic feet.

DESIGN PRES 5URE AND TEMPERATURE 5.2.2 The reactor containment is designed and shall be maintained for a maximum internal pressure of 34.5 psig and a temperature of 249'F.

5.3 REACTOR CORE FUEL ASSEMBLIES 2

5.3.1 The reactor core shall contain 76 fuel assemblies with each fuel asser.bly containing 230 or 231 fuel rods clad with Zircaloy-4 Each fuel rod shall have a nominal active fuel length of 91 inches.

Each fuel assembly shall contain a maximum total weight of 234 kilograms uranium. Reload fuel is identical in physical design to the Core XII EXXON fuel and shall have a maximum enrichment of 4.0 weight percent U-235.

YANKEE-ROWE' 5-1 Amendment tio. 43 i

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