ML19289D753

Amend 8 to License NPF-2 Re Fuel Rod Bow Penalty,Eccs Subsystem,Spent Fuel assemblies,quarter-core Flux Maps, Secondary Water Chemistry & Pressurizer Heatup Rate

ML19289D753
Person / Time
Site:	Farley  (NPF-02-A-008, NPF-2-A-8)
Issue date:	02/13/1979
From:	Schwencer A Office of Nuclear Reactor Regulation
To:
Shared Package
ML19289D754	List: ML19289D753 ML19289D755 ML19289D756
References
NUDOCS 7903140327
Download: ML19289D753 (30)

Text

.

4 UNITED STATES j#

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'g NUCLEAR REGULATORY COMMisslON W AStHNGTON, D. C. 20555 g..

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ALABAMA POWER COMPANY _

DOCKET N0. 50-348 9

JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING I,ICENSE Amendment No. 8 i

License No. NPF-2 1.

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

A.

The applications for amendment by the Alabama Power Company (the licensee) dated November 4 and December 14, 1977 and August 9,1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission, rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the l

health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment i

7 90314 0 3cN7 O

. and paragraph 2.C(2) of Facility Operating License No. NPF-2 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 8

, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Speciff cations.

3.

Tne license is further amended by adding the following new paragraph 2.C.(3)(g):

Alabama Power Company shall implement a secondary water 2.C.(3)(g) chemistry monitoring program to inhibit steam generator tube degradation.

This program shall include:

1.

Identification of a sanpling schedule for the critical parameters and control points for these parameters; 2.

Identification of the procedures used to quantify parameters that are critical to control points; 3.

Identification of process sampling points; 4.

Procedure for the recording and management of data; 5.

Procedures defining corrective actions for off control point i

chemistry conditions; and 6.

A procedure identifying the authority responsible for the inter-pretation of the data and the sequence and timing of administrative events required to initiate corrective action.

4.

The license amendment is effective as of the date of issuance.

FOR THE NU LEAR REGULATORY COMMISSION duaw A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance:

February 13, 1979

ATTACHMENT TO LICENSE AMENDMENT iiO. 8 FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.

Pages B 2-2 3/4 2-10a 3/4 3-39 3/4 4-29 3/4 5-5 3/4 5-Sa (added) 3/4 7-11 3/4 7-12 3/4 7-13 B 3/4 3-2 B 3/4 3-3 B 3/4 3-4 B 3/4 5-2 B 3/4 5-3 (added)

B 3/4 7-3 5-5 6-19

s l

t 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE _

The restrictions of this safety limit prevent overheating'pf the l

fuel and possible cladding perforation which would result in the release Overheating of th'e fuel of fission products to the reactor coolant.

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 I

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

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temper ~-

i ature 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-uni' form -heat flux distri-butions.

The local DNB heat flux ratio DNBR, defined as t'he ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the 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 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, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

8 FARLEY - UNIT 1 B 2-1

SAFETY LIMITS BASET N

5 and a reference cosine with a peak of 1.55 f r axial pcwer shah $., o The curves are based on an enthalpy hot channel factor, F An N

at reduced power based on allowance is included for an increase in FAH the express, ion:

1.55 [1 + 0.2 (1-P)] [1-RBP(BU)]

F 1

H RBP(BU) =.01 (-1.1667 + 0.05833 4UT for Bu > 400 $

khere:

and RBP (BU) = 0 for BU 1 400 These limiting heat flux conditions are higher than those calculated fpr the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the When the limits of the f, (al) function of the Overtemperature trip.

' axial power imbslance is not within the tolerance, the axial power imbalance effect on the Overtemperature AT trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACT 0R COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the inte.grity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel pressurizer and the reactor coolant system piping and fittings are designed to Section III of the ASME Code for. Nucl. ear Pcwer Plant wnich permits a maximum transient pres.sure of 110% (2735 psig) of design pressure.

The Safety Limit of 27-35 psig is therefore consistent with_ the design criteria and associated code requirements.

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

of design pres-sure, to demonstrate integrity prior to initial operation.

FARLEY - UNIT 1 B 2 -2 Amendment No. 8

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I versus Region Average Burnup Average FARLEY - Unit 1 3/4 2-10a Amendment No.8

INSTRUMENTATION MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The movable incore detection system shall be OPERABLE with:

a.

At least 80% of the detector thimbles, b.

A minimum of 2 detector thimbles per core quadrant, and c.

Suf ficient movable detectors, drive, and readout equipment to map these thimbles.

APPLICABILITY:

When the movab'>& incore detection system is used Tor:

a.

Recalibration of the excore neutron flux detection system, b.

Monitoring the QUADRANT POWER TILT RATIO, or Mc3surement of F"H, F (Z) and F l

c.

9 xy ACTION:

With the movable incore detection system inoperable, do not use the system for the above applicable monitoring or calibration ~ functions...

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable, t

SURVEILLANCE REQUIREMENTS 4.3.3.2 The movable incore detection system shall be demonstrated OPERABLE by-normali' zing each detector output when required for:

a.

Recalibration of the excore neutron flux detection _ system, or b.

Monitoring the QUADRANT POWER TILT RATIO ~, or-N 3H, F (Z) and Fxy' c.

Measurement of F q

FARLEY - UNIT 1 3/4 3-39 Amendment No. 8

INSTRUMENTATION SEISMIC INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE.

I APPLICABILITY: At all times.

ACTION:

With one or more seismic monitoring instruments inoperable for a.

more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10' days outlining the cause of the malfunction and the plans _for restoring the instrument (s) to OPERABLE status, b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILL NCE REQUIREMENTS 4.3.3.3.1 Each of the above seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-4.

4.3.3.~3.2 Each of the above seismic monitoring instruments actuated during a seismic event shall be restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and a CHANNEL CALIBRATION performed within 5 days following the seismic event.

Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days describing the magnitude, frequency spectrum and resultant effect upon facility features important to safety.

, fARLEY - UNIT 1 3/4 3-40

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:

A maximum cooldown of 200 F in any one hour period, i

a.

A maximum heatup of 100 F in any one hour period, and b.

A maximum spray water temperature differential of 320 F.

c.

i APPLICABILITY: At all times.

ACTION:

With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least. HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

1 SURVEILLANCE REQUIREMENT 5 The pressurizer temperatures shall be determined to be within 4.4.9.2 the limits at least once per hour during system heatup or cooldown.

The spray water temperature differential shall be determined to be withi.n the limit at least once.per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during auxiliary spray operation.

FARLEY - UNIT 1 3/4 4-29 Amendment No. 8

i i,

REACTOR _ COOLANT SYSTEM 3/4.4;10~ STRUCTURAL INTEGRITY S

ASME CODE CLASS 1, 2 and 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 components 1

shall be maintained in accordance with Specification 4.4.10.1.

i l

APPLICABILITY: ALL MODES ACTION:

With the structural integrity of any ASME Code Class 1 component (s) a.

not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50 F above the minimum temperature required by NDT considerations.

With the structural integrity of any ASME Code Cluss 2 component (s) b.

not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant. System temperature above 200*F.

With the structural integrity of any ASME Code Class 3 component (s) c.

not conforming to the above requirements, restore the structural integrity of the affected components to within its limit or isolate the affected component (s) from service.

'd.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.4.10.1 The structural integrity of ASME Code Class 1, 2 and 3 com-ponents shall be demonstrat~ed; FARLEY - UNIT I 3/4 4-30

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

A visual inspection of the containment sump and verifying 2.

that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens,

etc.) show no evidence of structural distress or corrosions, At least once per 18 months, during reactor shutdown and with-e.

in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of maintenance on or stroking operation of the following manual valves, verify that these valves are in the proper position for injection.

Valve Number _

CVC-V-8991 A/B/C CVC-V-8989 A/B/C CVC-V-8996 A/B/C CVC-V-8994 A/8/C f.

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

Verifying that each automatic valve in the flow path 1.

actuates to its correct position on a safety injection test signal.

Verifying that each of the following pumps start auto-2.

matica11y upon receipt of a safety injection test signal:

a)

Centrifugal charging pump b)

Residual heat removal pump By verifying th,a.t each of the following pumps develops a g.

discharge pressure (after subtracting suction pressure) on recirculation flow when tested pursuant to specification 4.0.5 :

1.

Centrifugal charging pump >_2458 psig.

2 Residual heat removal pump >_136 psig.

h.

Prior to entry into Mode 3 from Mode 4, verify that the mechanical stops on low lead safety injection valves RHR.HV 603 A/B are intact.

FARLEY - UNIT 1 3/4 5-5 Amendment No. 7,8

1

.t EMERGENCY' CORE COOLING SYSTEMS _

SURVEILLANCE REQUIREMENTS (Continued) i.

A flow balance test shall be conducted during shutdown to confirm the following minimum injection flow rates following completion of HPSI or LPSI system modifications that alter system flow characteristics:

HPSI System - Single Pump LPSI System - Single Pump 3,193 gpm (each injection leg) 3,3981 gpm (total injection) i i

i e

FARLEY - UNIT:1 3/4 5-Sa Amendment No. 8

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EMERGENCY CORE COOLING SYSTEMS _

ECCS SUBSYSTEMS - T,yg < 350 F LIMITING CONDITION FOR OPERATION t

3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

One OPERABLE centrifugal charging pump, b.

One OPERABLE residual heat removal heat exchanger, c.

One OPERABLE residual heat removal pump, and d.

An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: MODE 4.

ACTION:

a.

With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

b.

With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System Tavg less than 350 F by use of alternate heat removal methods c.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date.

SURVEILLANCE REQUIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.

FARLEY - UNIT 1 3/4 5-6

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