ML15113A164

Amends 139,139 & 136 to Licenses DPR-38,DPR-47 & DPR-55, Respectively,Updating Tech Spec Ref to FSAR to Ensure Consistency W/Updated FSAR

ML15113A164
Person / Time
Site:	Oconee
Issue date:	05/30/1985
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Duke Power Co
Shared Package
ML15113A165	List: ML15113A164 ML15113A166
References
DPR-38-A-139, DPR-47-A-139, DPR-55-A-136 NUDOCS 8506210445
Download: ML15113A164 (33)

Text

0 RE J NGITE UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-269 OCONEE NUCLEAR STATION, UNIT NO. I AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 139 License No. DPR-38

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

A. The application for amendment by Duke Power Company (the licensee) dated February 13, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No. DPR-38 is hereby amended to read as follows:

3.1 Technical Specifications The Technical Specifications contained in Appendices A and*B, as revised through Amendment No.139 are hereby incorporated -in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

9 506210445 850530 PDR ADOCK 05000269 PDR

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3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION J hn F. Stolz, Chi f

( perating Reactors Branch #4

ivision of Licensing

Attachment:

L Changes to the Technical Specifications Date of Issuance: May 30, 1985

• 1 A1&

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-270 OCONEE NUCLEAR STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 139 License No. DPR-47

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

A. The application for amendment by Duke Power Company (the licensee) dated February 13, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No. DPR-47 is hereby amended to read as follows:

3.8 Technical Specifications The Technical Specifications contained in Appendices A and-B, as revised through Amendment No.139 are hereby incorporated-in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION J n F. Stolz, Chie erating Reactors Branch #4 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: May 30, 1985

UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 DUKE POWER COMPANY DOCKET NO. 50-287 OCONEE NUCLEAR STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.136 License No. DPR-55

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

A. The application for amendment by Duke-Power Company (the licensee) dated February 13, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-55 is hereby amended to read as follows:

3.B Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 136 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

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3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION o

F. Stolz, Chief ating Reactors Branch #4 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: "lay 30, 1985

ATTACHMENTS TO LICENSE AMENDMENTS AMENDMENT NO. 139 TO DPR-38 AMENDMENT NO. 139 TO DPR-47 AMENDMENT NO. 136 TO DPR-55 DOCKETS NOS. 50-269, 50-270 AND 50-287 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment numbers and contain vertical lines indicating the area of change.

Remove Pages Insert Pages 1-2 1-2 1-3 1-3.

2.2-1 2.2-1 2.3-4 2.3-4 3.1-2 3.1-2 3.1-4 3.1-4 3.1-9 3.1-9 3.1-11 3.1-11 3.1-12 3.1-12*

3.1-13 3.1-13 3.1-16 3.1-16 3.1-18 3.1-18 3.2-2 3.2-2 3.3-6 3.3-6 3.6-3 3.6-3 3.13-1 3.13-1 3.16-1 3.16-1 4.3-1 4.3-1 4.4-5 4.4-5 4.4-6 4.4-6*

4.5-12 4.5-12 4.6-3 4.6-3 4.7-1 4.7-1 4.8-1 4.8-1 4.9-1 4.9-1 5.2-1 5.2-1*

5.2-2 5.2-2 5.3-1 5.3-1

• Overleaf page included for document completeness.

1.2.7 Refueling Operation An operation involving a change in core geometry by manipulation of fuel or control rods when the reactor vessel head is removed.

1.2.8 Startup The reactor shall be considered in the startup mode when the shutdown margin is reduced with the intent of going critical.

1.3 OPERABLE A system, subsystem, train, component or device shall be considered OPERABLE when it is capable of performing its intended safety functions. Implicit in this definition shall be the assumption that all essential auxiliary equip ment required in order to assure performance of the safety function is capable of performing its related support function(s). Auxiliary equipment includes but is not limited to normal or emergency electrical power sources, cooling and seal water, instrumentation and controls, etc.

If either the normal or emergency power to system, subsystem, train, component or device is not avail able it is considered OPERABLE for the purpose of satisfying the requirements of its applicable Limiting Condition for Operation, provided:

(a) the alternate power source is available, and (b) the redundant system is operable.

1.4 PROTECTIVE INSTRUMENTATION LOGIC 1.4.1 Instrument Channel An instrument channel is the combination of sensor, wires, amplifiers and output devices which are connected for the purpose of measuring the value of a process variable for the purpose of observation, control and/or protection.

An instrument channel may be either analog or digital in nature.

1.4.2 Reactor Protective System The reactor protective system is shown in Figures 7.2-1 and 7.2-4 of the FSAR.

It is that combination of protective channels and associated circuitry which forms the automatic system that protects the reactor by control rod trip.

It includes the four protective channels, their associated instrument channel inputs, manual trip switch, all rod drive protective trip breakers and acti vating relays or coils.

1.4.3 Protective Channel A protective channel as shown in Figure 7.2-1 of the FSAR (one of three or one of four independent channels, complete with sensors, sensor power supply units, amplifiers and bistable modules provided for every reactor protective safety parameter) is a combination of instrument channels forming a single digital output to the protective system's coincidence logic.

It includes a shutdown bypass circuit, a protective channel bypass circuit and reactor trip module and provision for insertion of a dummy bistable.

Amendments Nos.

139, 139, & 136 1-2

1.4.4 Reactor Protective System Logic This system utilizes reactor trip module relays (coils and contacts) in all four of the protective channels as shown in Figure 7.2-1 of the FSAR, to pro vide reactor trip signals for de-energizing the six control rod drive trip breakers.

The control rod drive trip breakers are arranged to provide a one out of two times two logic.

Each element of the one out'of two times two logic is controlled by a separate set of two out of four logic contacts from the four reactor protective channels.

1.4.5 Engineered Safety Features System This system utilizes relay contact output from individual channels arranged in three analog sub-systems and two two-out-of-three logic sub-systems as shown in Figure 7.3-1 of the FSAR. The logic sub-system is wired to provide appropriate signals for the actuation of redundant Engineered Safety Features equipment on a two-of-three basis for any given parameter.

1.4.6 Degree of Redundancy The difference between the number of operable channels and the number of channels which when tripped will cause an automatic system trip.

1.5 INSTRUMENTATION SURVEILLANCE 1.5.1 Trip Test A trip test is a test of logic elements in a protective channel to verify their associated trip action.

1.5.2 Channel Test A channel test is the injection of an internal or external test signal into the channel to verify its proper output response; including alarm and/or trip initiating action where applicable.

1.5.3 Instrument Channel Check An instrument channel check is a verification of acceptable instrument performance by observation of its behavior and/or state; this verification includes comparison of output and/or state of independent channels measuring the same variable.

1.5.4 Instrument Channel Calibration An instrument channel calibration is a test, and adjustment (if necessary), to establish that the channel output responds with acceptable range and accuracy to known values of the parameter which the channel measures or an accurate simulation of these values. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip and shall be deemed to include the channel test.

Amendments Nos.139

, 139, & 136 1-3

2.2 SAFETY LIMITS -

REACTOR COOLANT SYSTEM PRESSURE Applicability Applies to the limit on reactor coolant system pressure.

Objective To maintain the integrity of the reactor coolant system and to prevent the release of significant amounts of fission product activity.

Specification 2.2.1 The reactor coolant systeni pressure shall not exceed 2750 psig when there are fuel assemblies in the reactor vessel.

2.2.2 The setpoint of the pressurizer code safety valves shall be in accordance with ASME, Boiler and Pressurizer Vessel Code,Section III, Article 9, Summer 1967.

Bases (1)

The reactor coolant system serves as a barrier to prevent radionuclides in the reactor coolant from reaching the atmosphere.

In the event of a fuel cladding failure, the reactor coolant system is a barrier against the release of fission products. Establishing a system pressure limit helps to assure the integrity of the reactor coolant system. The maximum transient pressure allowable in the reactor coolant system pr.T1ure vessel under the ASME code,Section III, is 110% of design pressure.

The maximum transient pressure allowable in the reactor coolant system piping, valves, and fittings under USAS Section B31.7 is 110% of design pressure. Thus, the safety lim of 2750 psig (110% of the 2500 psig design pressure) has been established.

The settings, the reactor hgh pressure trip (2300 psig) and the pressurizer safety valves (2500 psig) have been established to assure never reaching the reactor coolant system pressure safety limit. The initial hydrostatic test was conducted at 3125 psig (125% of design pressure) to verify the integrity of the reactor coolant system. Additional assurance that the Reactor Coolant pressure does not exceed the safety limit is provided by setting the pressurizer electromatic relief valve at 2450 psig.

REFERENCES (1) FSAR, Section 5 (2) FSAR, Section 5.2.3.10.1 (3) FSAR, Section 5.2.2.3, Table 5.4-7 (4) FSAR, Section 5.4.6, Table 5.1-1 Amendments Nos.

139, 139, &136 2.2-1

Shutdown Bypass In order to provide for control rod drive tests, zero power physics testing, and startup procedures, there is provision for bypassing certain segments of the reactor protection system. The reactor protection system segments which can be bypassed are shown in Table 2.3-1A. Two conditions are imposed when 2.3-1B 2.3-1C the bypass is used:

1. By administrative control the nuclear overpower trip setpoint must be reduced to a value <5.0% of rated power during reactor shutdown.

2. A high reactor coolant system pressure trip setpoint of 1720 psig is automatically imposed.

The purpose of the 1720 psig high pressure trip setpoint is to prevent normal operation with part of the reactor protection system bypassed.

This high pressure trip setpoint is lower than the normal low pressure trip setpoint so that the reactor must be tripped before the bypass is initiated. The over power trip setpoint of <5.0% prevents any signi ficant reactor power from being produced when performing the physics tests.

Sufficient natural circulation (5) would be available to remove 5.0% of rated power if none of the reactor coolant pumps were operating.

Single Loop Operation Single loop operation is permitted only after the reactor has been tripped and is subject to the limitations set forth in Specification 3.1.8. The RPS trip setting limits and permissible instrument channels bypasses will be confirmed prior to single loop operation.

REFERENCES (1) FSAR, Section 15.2.1 (2) FSAR, Section 15.7.1 (3) FSAR, Section 15.8.1 (4) FSAR, Sections 15.3.1, and 15.3.3 (5) FSAR, Section 15.6.3 Amendments Nos. 139 139

& 136 2.3-4

0

0.

Bases The limitation on power operation with one idle RC pump in each loop has been imposed since the ECCS cooling performance has not been calculated in ac cordance with the Final Acceptance Criteria requirements specifically for this mode of reactor operation. (1) A time period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allowed for operation with one idle RC pump in each loop to effect repairs of the idle pump(s) and to return the reactor to an acceptable combination of operating RC pumps. The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for this mode of operation is acceptable since this mode is expected to have considerable margin for the peak cladding temperature limit and since the likelihood of a LOCA within the 24-hour period is considered very remote.

A reactor coolant pump or low pressure injection pump is required to be in operation before the boron concentration is reduced by dilution with makeup water. Either pump will provide mixing which will prevent sudden positive reactivity changes caused by dilute coolant reaching the reactor. One low pressure injection pump will circulate the equivalent of the reactor coolant system volume in one-half hour or less.

(2)

The low pressure injection system suction piping is designed for 300aF and

370 psig; thus the system with its redundant components can remove decay heat when the reactor coolant system is below this temperature.

(3,4)

One pressurizer code safety valve is capable of preventing overpressurization when the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat.

(5) Both pressurizer code safety valves are required to be in service prior to criticality to conform to the system design relief capabilities. The code safety valves prevent overpressure for a rod withdrawal accident at hot shutdown.

(6) The pressurizer code safety valve lift setpoint shall be set at 2500 psig +/-1% allowance for error and each valve shall be capable of relieving 300,000 lb/hr of saturated steam at a pressure no greater than 3% above the set.pressure.

REFERENCES (1) FSAR, Section 5.1.2.3 (2) FSAR, Section 6.3.3.2, and Tables 5.3-1, 5.4-2, 5.4-3, 5.4-6, 5.4-7, 5.4-8, and 6.3-2 (3) FSAR, Sections 5.4.7-1, and 9.3.3.2.3 (4) FSAR, Sections 5.4.7.4, and 6.3.3.2 (5) FSAR, Sections 5.2.3.10.4, and 5.4.6 (6) FSAR, Sections 5.2.3.7, and 15.2.3 Amendments Nos. 139, 139, & 136 3.1-2

Bases -

Units 1, 2 and 3 All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, startup and shutdown operations, and inservice leak and hydrostatic tests.

The various categories of load cycles used for design purposes are provided in Table 5.2-1 of the FSAR.

The major components of the reactor coolant pressure boundary have been analyzed in accordance with Appendix G to 10 CFR 50. Results of this analysis, including the actual pressure-temperature limitations of the reactor coolant pressure boundary, are given in BAW-1699 and BAW-1691.

The Figures specified in 3.1.2.1, 3.1.2.2 and 3.1.2.3 present the pressure temperature limit curves for normal heatup, normal cooldown and hydrostatic tests respectively. The limit curves are applicable up to the indicated effective full power years of operation. These curves are adjusted by 25 psi and 10aF for possible errors in the pressure and temperature sensing instruments. The pressure limit is also adjusted for the pressure differ ential between the point of system pressure measurement and the limiting component for all operating reactor coolant pump combinations.

The cooldown limit curves are not applicable to conditions of off-normal operation (e.g., small LOCA and extended loss of feedwater) where cooling is achieved for extended periods of. time by circulating water from the HPI through the core.

If core cooling is restricted to meet the cooldown limits under other than normal operation, core integrity could be jeopardized.

The pressure-temperature limit lines shown on the figures specified in 3.1.2.1 for reactor criticality and on the figures referred to in 3.1.2.3 for hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 for reactor criticality and for inservice hydrostatic testing.

The actual shift in RTNDT of the beltline region material will be established periodically during operation by removing and evaluating, in accordance with Appendix H to 10 CFR 50, reactor vessel material irradiation surveillance specimens which are installed near the inside wall of this or a similar reactor vessel in the core region, or in test reactors.

The limitation on steam generator pressure and temperature provide protection against nonductile failure of the secondary side of the steam generator. At

.metal temperatures lower than the RT of +60aF, the protection against nonductile failure is achieved by limling the secondary coolant pressure to 20 percent of the. preoperational system hydrostatic test pressure. The Amendments Nos. 139, 139, & 136 3.1-4

appropriate by on site diffusion measurements using SF6 (sulfur hexafluoride) as a gas tracer.

The combined gamma and beta whole body dose from a semi-infinite cloud is given by:

Dose (Rem) = 1/2[EA-V*X/Q*(3.7 x 1010 dps/Ci) - (1.33 x 1011 Rem/N'eV/m')1 Dose (Rem) = 0.246-E*A-V-X/Q Amax (pc/cc) = (Dose)max 0.5 0.246*E*V*X/Q 0.246 x E x 78.25 x 1.16 x 10 A

(pc/cc) = 224/E max Where A = Reactor coolant activity (pCi/ml = Ci/m 3)

V = Reactor coolant volume at 580'F leaked into secondary system (2763 ft 3 =

78.25 m3 )

X/Q = Atmospheric dispersion coefficient at site boundary for a two hour period (1.16 x 10 4 sec/m')

E = Average beta and gama energies per disintegration (HeV) corrected to operating temperature and pressure.

Calculations required to determine Z will consist of the following:

1. Quantitative measurement of the specific activity (in units of pCi/cc) of radionuclides with half lives longer than 30 minutes, which make up at least 95% of the total activity in reactor coolant samples.

2. A determination of the average beta and gamma decay energies per disinte gration for each nuclide, measured in (1) above, by utilizing known decay energies and decay schemes (e.g., Table of Isotopes, Sixth Edition, March 196S).

3. A calculation of E by the average beta and gamma energy for each radionuc lide in proportion to its specific activity, as measured in (1) above.

RE FREN CE FSAR, Section 15.9 Amendments Nos.

139 139 6

135 3.1-11

S:'ec if ic t ion 3.1.5.1 If the

entr 2 ion of o

'xyg.

in th primary coolant exce L. C!

durin, pow.-r operation, correctiv. action shall be initiated :i eight hours to return o-ygen levels to < 0.1 ppm.

3.1.5.2 If the concentration of chloride in the prim.ary cool3nt exceed.4s U.15 ppm during pover operation, corrective action shall be iritiateu eight hours to return chloride levels to < 0.15 pp-n.

3.1.5.3 If the concentration of fluorides in the primary coolant e:ce;:ds 0.1 ppm follo:ing modifications-or repair to the primary syste:: involv i2 welding, corrective action shall be initiated within eight hours to return fluoride levels to < 0.15 p.>m.

3.1.5.4 If the concentration limits of oxygen, chloride or fluoride in 3.1.5.

3.1.5.2 and 3.1.5.3 above are not restored within 24 houi:s the rm t' shall be placed in a hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thar2 after.

If the normal operational limits are not restored within an additional 24-hour period, the reactor shall be placed in a cold s down condition within 24-hours thereafter.

3.1.5.5 If the oxygen concentration and the chloride or fluoride concentra tion of the primary coolant system individually exceed 1.0 pp.

the reactor shall be immediately brought to the hot shutdown corditio.

using normal shutdown procedure and action is to be taken i!7.Fdiatclx to return the system to within normal operation specifications.

If normal operating specifications have not been reached in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the reactor shall be brought to a cold shutdown condition using normal procedure.

Bases By maintaining the chloride, fluoride and oxygen concentration in the rea:

coolant within the specif ications, the integritV of the reactor coolntC s.:

is protected against Dotential stress corrosion attack.

(1,2)

The oxygen concentration in the reactor coolant system is nor::.ily e::ctad t be below detectable limits since dissolved hydrogen is used-when the reactor is critical and a residual of hydrazine is used when the reactor is sacritical to control the o:<vgen.

The requireomnt that the oxvgen cocentratio 0 ot exceed 0.1 ppm. during power operation is added assurance thac stress corrsi.

cracks will not occur.

If the oxygen, chloride, or fl1uoride limits are exceeded, measures can be taken to correct the cond tion (e.g.,

switch to the spare demineralir.'r replace the ion exchange resin, increase the hydrocen :oncentrat ion ii makeup tank, czc.)

and further becaus;e of the Li7e dependent nature L: auv adverse effects arising from chlori de or oxgCen cohcentrations in c::cS 2:

the limits, it is unnecessary to shutdo.,:n immediately, since the condition2 a

be corrected.

3.1-12

d. Total reactor coolant system leakage rate is periodically determined by comparing indications of reactor power, coolant temperature, pressurizer water level and letdown storage tank level over a time interval.

All of these indications are recorded.

Since the pressurizer level is main tained essentially constant by the pressurizer level controller, any coolant leakage is replaced by coolant from the letdown storage tank resulting in a tank level decrease.

The letdown storage tank capacity is 31 gallons per inch of height and each graduation on the level recorder represents 1 inch of tank height.

This inventory monitoring method is capable of detecting changes on the order of 31 gallons. A 1 gpm leak would therefore be detectable within approximately one half hour.

As described above, in addition to direct observation, the means of detecting reactor coolant leakage are based on 2 different principles, i.e., activity, sump level and reactor constant inventory measurements.

Two systems of different principles provide, therefore, diversified ways of detecting leakage to the reactor building.

The upper limit of 30 gpm is based on the contingency of a complete loss of station power. A 30 gpm loss of water in conjunction with a complete loss of station power and subsequent cooldown of the reactor coolant system by the turbine bypass system (set at 1,040 psia) and steam driven emergency feedwater pump would require more than 60 minutes to empty the pressurizer from the com bined effect of system leakage and contraction. This will be ample time to restore electrical power to the station and makeup flow to the reactor coolant system.

REFERENCES FSAR Sections 11.5.1, and 5.2.3.10.3 Amendments Nos. 139, 139, & 136 3.1-16

D. Dissolved boron concentration This correction is for any difference in boron concentration, if required, between zero and full power. Since the moderator coefficient is more positive for greater dissolved boron concen trations, the sign of the correction depends on whether boron is added or removed. The correction is 0.16 x 10-6 La /APPM. (The magnitude of the correction varies slightly with boron concen tration; the value presented above, however, is valid for a range in boron concentrations from 1000 to 1400 ppm.)

E. Control rod insertion This correction is for the difference in control rod worth (%

Ak/k) in the core between zero and full power. The correction is 0.17 x 10- 4a m/%Ak/k, where the sign for rod worth change is negative for rod insertion, because the moderator coefficient is more negative for a larger rod worth in the core.

F. Isothermal to distributed temperature The correction for spatially distributed moderator temperature has been found to be less than or equal to zero. Therefore, zero is a conservative correction value for distributed effects.

G. Azimuthal :xenon stability Before commercial operation a test will be performed to verify that divergent azimuthal xenon oscillations do not occur.

REFERENCES (1) FSAR, Section 15 (2) FSAR, Sections 4.2, and 4.3.2 (3) FSAR, Section 15.14 Amendments Nos.

139, 139, & 136 3.1-18

Bases The high pressure injection system and chemicat addition system provide con trol of the reactor coolant system boron concentration.)

This is normall accomplished by using any of the three high pressure injection pumps im ser es with a boric acid pump associated with either the boric acid mix tank or the concentrated boric acid storage tank. An alternate method of boration will be the use of the high pressure injection pumps taking suction directly from the borated water storage tank.(2)

The quantity of boric acid in storage in the concentrated boric acid storage tank or the borated water storage tank is sufficient to borate the reactor coolant system to a 1% Akik subcritical margin at cold conditions (70F) with the maximum worth stuck rod and no credit for xenon at the worst time in core life.

The current cycles for each unit were analyzed with the most limiting case selected as the basis for all three units.

Since only the present cycles were analyzed, the specifications will be re-evaluated with each reload. A minimum of 1020 ft3 of 8,700 ppm boric acid in the concentrated boric acid storage tank, or a minimum of 350,000 gallons of 1835 ppm boric acid in the borated water storage tank (3) will satisfy the requirements.

The volume re quirements include a 10% margin and, in addition, allow for a deviation of 10 EFPD in the cycle length. The specification assures that two supplies are available whenever the reactor is critical so that a single failure will not prevent boration to a cold condition.

The required amount of boric acid can be added in several ways.

Using only one 10 gpm boric acid pump taking suction from the concentrated boric acid storage tank would require approximately 12.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to inject the required boron. An alternate method of addition is to inject boric acid from the borated water storage tank using the makeup pumps.

The required boric acid can be injected in less than six hours using only one of the makeup pumps.

The concentration of boron in the concentrated boric acid storage tank may be higher than the concentration which would crystallize at ambient conditions.

For this reason, and to assure a flow of boric acid is available when needed, these tanks and their associated piping will be kept at least 10*F above the crystallization temperature for the concentration present.

The boric acid concentration of 8,700 ppm in the concentrated boric acid storage tank cor responds to a crystallization temperature of 77*F and therefore a temperature requirement of 87*F.

Once in the high pressure injection system, the concen trate is sufficiently well mixed and diluted so that normal system temperatures assure boric acid solubility.

.REFERENCES (1)

FSAR, Sections 9.3.1, and 9.3.2 (2)

FSAR, Figure 6.0.2I (3)

Technical Specification 3.3 Amendments Nos. 139

,139 136 3.2-2

Three hundred and fifty thousand (350,000) gallons of borated water (a level of 46 feet in the BWST) are required to supply emergency core cooling and reactor building spray in the event of a loss-of-core cooling accident. This amount fulfills requirements for emergency core cooling. The borated water storage tank capacity of 388,000 gallons is based on refueling volume require ments. Heaters maintain the borated water supply at a temperature above 50aF to lessen the potential for thermal shock of the reactor vessel during high pressure injection system operation. The boron concentration is set at the amount of boron required to maintain the core 1 percent subcritical at 701F without any control rods in the core. The minimum value specified in the tanks is 1835 ppm boron.

It has been shown for the worst design basis loss-of-coolant accident (a 14.1 ft2 hot leg break) that the Reactor Building design pressure will not be exceeded with one spray and two coolers operable. (4) Therefore, a maintenance period of seven days is acceptable for one Reactor Building cooling fan and its associated cooling unit provided two Reactor B'uilding spray systems are operable for seven days or one Reactor Building spray system provided all three Reactor Building cooling units are operable.

Three low pressure service water pumps serve Oconee Units I and 2 and two low pressure service water pumps serve Oconee Unit 3. There is a manual cross connection on the supply headers for Unit 1, 2, and 3. One low pressure service water pump per unit is required for normal operation. The normal operating requirements are greater than the emergency requirements following a loss-of coolant accident.

Prior to initiating maintenance on any of the components, the redundant component(s) shall be tested to assure operability. Operability shall be based on the results of testing as required by Technical Specification 4.5.

The maintenance period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is acceptable if the operability of equipment redundant to that removed from service is demonstrated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to removal. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period prior to removal is adequate to permit efficient scheduling of manpower and equipment testing while ensuring that the testing is performed directly prior to removal. The basis of acceptability is the low likelihood of failure within a clearly defined 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> following redundant component testing.

REFERENCES (1) ECCS Analysis of B&W's 177-FA Lowered-Loop NSS, BAW-10103, Babcock &

Wilcox, Lynchburg, Virginia, June 1975.

(2) Duke Power Company to NRC letter, July 14, 1978, "Proposed Modifications of High Pressure Injection System".

(3) FSAR, Section 9.3.3.2 (4) FSAR, Section 15.14.5 Amendments Nos.

139, 139, & 136 3.3-6

When containment integrity is established, the limits of 10CFR100 will not be exceeded should the maximum hypothetical accident occur.

REFERENCE FSAR, Section 3.8 Amendments Nos.

139

, 139

, & 136 3.6-3

3.13 SECONDARY SYSTEM ACTIVITY Applicability Applies to the limiting conditions of secondary system activity for operation of the reactor.

Objective To limit the maximum secondary system activity.

Specification The iodine-131 activity in the secondary side of a steam generator shall not exceed 1.4 pCi/cc.

Bases For the purpose of determining a maximum allowable secondary coolant activity, the activity contained in the mass released following a loss of load accident is considered. As stated in FSAR Section 15.8.2, 148,000 pounds of water is released to the atmosphere via the relief valves. A site boundary dose limit-of 1.5 rem is used.

The whole body dose is negligible since any noble gases entering the secondary coolant system are continuously vented to the atmosphere by the condenser air ejector, thus, in the event of a loss of load incident there are only small quantities of these gases which would be released.

1-131 is the significant isotope because of its low MPC in air and because the other iodine isotopes have shorter half-lives, and therefore, cannot build up to significant concentrations in the secondary coolant, given the limitations on primary system leak rate and technical specification limiting activity.

One-tenth of the contained iodine is assumed to reach the site boundary, making allowance for plateout and retention in water droplets. 1-131 is assumed to contribute 70% of the total thyroid dose based on the ratio of 1-131 to the total iodine isotopes given in Table 11.2-2 of the FSAR.

The maximum inhalation dose at the site boundary is then as follows:

Dose (rem) = Ci* V*B*DCF *(0.1)* X/Q C = Secondary coolant activity (2.0 pCi/cc 1-131 equivalent)

V = Secondary water volume released to atmosphere (90 M3 )

.B = Breathing rate (3.47 x 10 4 m3 /sec)

X/Q = Ground level release dispersion factor (1.16 x 10 sec/m 3)

DCF = 1.48 x 106 rem/Ci 0.1 = Fraction of activity released The resultant dose is 1.15 rem compared to the Radiation Protection Guide of 1.5 rem for an annual individual exposure in an unrestricted area.

Amendments Nos. 139, 139, & 136 3.13-1

3.16 HYDROGEN PURGE SYSTEM Applicability Applies to the Reactor Building Hydrogen Purge System.

Objective To define the conditions necessary to assure the availability of the Reactor Building Hydrogen Purge System.

Specification If the Reactor Building Hydrogen Purge System should become inoperable, it shall be'restored to an operable status within 7 days or the Oconee Units shall be shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Bases The hydrogen purge system is composed of a portable purging station and a portion of the Penetration Room Ventilation System. The purge system is operated as necessary to maintain the hydrogen concentration below the control limit. The purge discharge from the Reactor Building is taken from one of the Penetration Room Ventilation System penetrations and discharged to the unit vent. A suction may be taken on the Reactor Building via isolation valve PR-7 (Figure 6.0-5 of the FSAR) using the existing vent and pressurization connections.

The analysis to determine the effect on the incremental doses at the site boundary, resulting from purging hydrogen from the Reactor Building following a postulated LOCA, requires that the purge be started at 460 hours0.00532 days <br />0.128 hours <br />7.60582e-4 weeks <br />1.7503e-4 months <br /> (19.2 days) following the LOCA to limit hydrogen concentration to 4% by volume.

If the Hydrogen Purge System is determined to be inoperable, the requirement to restore the system to an operable status within seven days will provide reasonable assurance of its availability in the event of a LOCA.

Amendments Nos. 139, 139, & 136 3.16-1

4.3 TESTING FOLLOWING OPENING OF SYSTEM Applicability Applies to test requirements for Reactor Coolant System integrity.

Objective To assure Reactor Coolant System integrity prior to return to criticality following normal opening, modification, or repair.

Specification 4.3.1 When Reactor Coolant System repairs or modifications have been made, these repairs or modifications shall be inspected and tested to meet all applicable code requirements prior to the reactor being made critical.

4.3.2 Following any opening of the Reactor Coolant System, it shall be leak tested at not less than 2200 psig prior to the reactot-being made critical.

4.3.3 The limitations of Specification 3.1.2 shall apply.

Bases Repairs or modifications made to the Reactor Coolant System are inspectable and testable under applicable codes. The specific code and edition thereof shall be consistent with 10 CFR 50.55.

REFERENCE FSAR, Section 5 Amendments No.

139,139

, & 136 4.3-1

When co.tainment integrity is established, the overall containment leak rate of 0.25 weight percent of containment air at 59 psig will assure that the limits of IOCFRIOO will not be exceeded should the maximum hypothetical accident occur.

In order to assure the integrity of the containment, periodic testing is performed at reduced pressure, 29.5 psig. The permissible leakage rate at this reduced pressure has been established from the initial integrated leak rate tests in conformance with 10CFR5O, Appendix J.

The containment air locks (i.e., Personnel Hatch and Emergency Hatch) are tested on a more frequent basis than other penetrations. The air locks are utilized during periods of time when containment integrity is required as well as when the reactor is shutdown. Proper verification of door seal integrity is required to ensure conCainment integrity. Because the door seals are recessed, damage from tools due to air lock entry is improbable; however, a leak test of the outer door seals has been shown to be an acceptable alternative to the full hatch test to ensure air lock integrity.

REFERENCE (1) FSAR, Sections 3.8.1.7.4, 6.2.4, and 14 Amendments Nos.

139 139 6

136 4.4-5

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4.5.4 Low Pressure Injection System Leakage Applicability Applies to Low Pressure Injection System leakage.

Objective To maintain a preventive leakage rate for the Low Pressure Injection System which will prevent.significant off-site exposures.

Specification 4.5.4.1 Acceptance Limit The maximum allowable leakage from the Low Pressure Injection System components (which includes valve stems, flanges and pump seals) shall not exceed two gallons per hour.

4.5.4.2 Test During each refueling outage, the following tests of the Low Pressure In jection System shall be conducted to determine leakage:

a. The portion of the Low Pressure Injection System, except as specified in (b), that is outside the containment shall be tested either by use in normal operation or by hydrostatically testing at 350 psig.

b. Piping from the containment emergency sump to the low pressure injection pump suction isolation valve shall be pressure tested at no less than 59 psig.

c. Visual inspection shall be made for excessive leakage from components of the system. Any excessive leakage shall be measured by collection and weighing or by another equivalent method.

Bases The leakage rate limit for the Low Pressure Injection System is a judgment value based on assuring that the components can be expected to operate with out mechanical failure for a period on the order of 200 days after a loss of coolant accident. The test pressure (350 psig) achieved either by normal system operation or by hydrostatically testing, gives an adequate margin over the highest pressure within the system after a design basis accident.

Similarly, the pressure test for the return lines from the containment to the Low Pressure Injection System (59 psig) is equivalent to the design pressure of the containment.

The dose to the thyroid calculated as a result of this leakage is 0.76 rem for a two-hour exposure at the site boundary.

REFERENCE FSAR, Sections 15.15.4, and 6.3.3.2.2 Amendments Nos.

139,

139, & 136 4.5-12

e 0

4.6.10 Annually, a one hour discharge service test at the required maximum load shall be made on tbe instrument and control batteries, the Keowee batteries, and the switching station batteries.

4.6.11 Monthly, the operability of the individual diode monitors in the Instru ment and Control Power System shall be verified by imposing a simulated diode failure signal on the monitor.e 4.6.12 Semiannually, the peak inverse voltage capability of each auctioneering diode in the 125 VDC Instrument and Control Power System shall be measured and recorded.

Bases The Keowee Hydro units, in addition to serving as the emergency power sources for the Oconee Nuclear Station, are power generating sources for the Duke system requirements. As power generating units, they are operated frequently, normally on a daily basis at loads equal to or greater than required by Table 8.1-1 of the FSAR for ESF bus loads. Normal as well as emergency startup and operation of these units will be from the Oconee Unit I and 2 Control Room. The frequent starting and loading of these units to meet Duke system power requirements assures the continuous availability for emergency power for the Oconee auxiliaries and engineered safety features equipment.

It will be verified that these units will carry the equipment of the maximum safeguards load, within 25 seconds, including instrumentation lag, after a simulated re quirement for engineered safety features. To further assure the reliability of these units as emergency power sources, they will be, as specified, tested for automatic start on a monthly basis from the Oconee control room. These tests will include verification that each unit can be synchronized to the 230 kV bus and that each unit can energize the 13.8 kV underground feeder.

The interval specified for testing of transfer to emergency power sources is based on maintaining maximum availability of redundant power sources.

Starting a Lee Station gas turbine, separation of the 100 kV line from the remainder of the system, and charging of the 4160 volt main feeder buses-are specified to assure the continuity and operability of this equipment. The one hour time limit is considered the absolute maximum time limit that would be required to accomplish this.

REFERENCE FSAR, Section 8 Amendments Nos.

139 139

& 136 4.6-3

4.7 REACTOR CONTROL ROD SYSTEM TESTS 4.7.1 Control Rod Trip Insertion Time Test Applicability Applies to the surveillance of the control rod trip insertion time.

Objective To assure the control rod trip insertion time is within that used in the safety analyses.

Specification The control rod insertion time shall be measured at either full flow or no flow conditions as follows:

a.

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

b.

For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and

c.

For all rods at least once following each refueling outage.

The maximum control rod trip insertion time for an operable control rod drive mechanism, except for the Axial Power Shaping Rods (APSRs), from the fully withdrawn position to 3/4 insertion (104 inches travel) shall not exceed 1.66 seconds at reactor coolant full flow conditions or 1.40 seconds for no flow conditions. For the APSRs it shall be demonstrated that loss of power will not cause rod movement. If the trip insertion time above is not met, the rod shall be declared inoperable.

Bases The control rod trip insertion time is the total elapsed time from power interruption at the control rod drive breakers until the control rod has completed 104 inches of travel from the fully withdrawn position. The specified trip time is based upon the safety analysis in FSAR Chapter 15.

A rod is considered inoperable if the trip insertion time is greater than the specified allowable time.

REFERENCES (1) FSAR, Section 15 (2) Technical Specification 3.5.2 Amendments Nos.

139, 139 136 4.7-1

4.8 MAIN STEAM STOP VALVES Applicability Applies to the main steam stop valves.

Objective To verify the ability of the main steam stop valves to close upon signal and to verify the leak tightness of the main steam stop valves.

Specification 4.8.1 Using Channels A and B, the operation of each of the main steam stop valves shall be tested during each refueling outage to demon strate a closure time of one second or less in Channel A and a clo sure time of 15 seconds or less for Channel B.

4.8.2 The leak rate through the main steam stop valves shall not exceed 25 cubic feet per hour at a pressure of 59 psig and shall be tested during each refueling outage.

Bases The main steam stop valves limit the Reactor Coolant System cooldown rate and resultant reactivity insertion following a main steam line break accident.

Their ability to promptly close upon redundant signals will be verified during each refueling outage. Channel A solenoid valves are designed to close all four turbine stop valves in 240 milliseconds. The backup Channel B solenoid valves are designed to close the turbine stop valves in approximately 12 seconds.

Using the maximum 15 second stop valve closing time, the fouled steam generator inventories and the minimum tripped rod worth with the maximum stuck rod worth, an analysis similar to that presented in FSAR Section 15.13 (but considering a blowdown of both steam generators) shows that the reactor will remain sub critical after reactor trip following a double-ended steam line break.

The main stop valves would become isolation valves in the unlikely event that there should be a rupture of a reactor coolant line concurrent with rupture of the steam generator feedwater header. The allowable leak rate of 25 cubic feet per hour is approximately 25 percent of total allowable containment leakage from all penetrations and isolation valves.

REFERENCES (1) FSAR, Section 10.3.4, and 15.13 (2)

Technical Specification 4.4.1 Amendments Nos.

139,139

, & 136 4.8-1

4.9 E=EGENCY7EED:E PLY? AND VAL'.E PE IODIC TESTING A1%.)li cabil1i tv ADplies to the periodic testing of the turbine-driven and motor-driven emergency fesdwater pu-ps and associated valves.

Obiective To verify that the emergency feedwater pumps and associated valves are operable.

Snecification 4.9.1 PuM Test The turbine-driven and motor-driven feedwater pumps shall be operated on recirculation to the upper surge tank for a minimum of one hour in accordance with the requirements of Specification 4.0.4.

4.9.2 Valve Test Automatic valves in the emergency feedwater flow path will be determined to be operable in accordance with the requirements of Specification 4.0.4.

4.9.3 System Flow Test Prior to Unit operation above 25% Full Power following any modifications or repairs to the emergency feedwater system which could degrade the flow path and at least once per refueling cycle, the emergency feedwater system shall be given either a manual or an automatic initiation signal.

4.9.4 Acceptance Criteria These tests shall be considered satisfactory if control board indication and visual observation of the equipment demonstrates that all components have operated properly.

In addition, during operation of the System Flow Test (Item 4.9.3 above), flow to the steam generators shall be verified by control room indication.

Bases.

The monthly testing frequency is sufficient to verify that the emergency feed water pumps are operable. Verification of correct operation is made both from the control room instrumentation and direct visual observation of the pumps.

The parameters which are observed are detailed in the applicable edition of the ASMi Boiler and Pressure Vessel Code,Section XI.

The System Flow Test verifies correct total system operation following modifications or repairs.

REFERENCE (1) FSAR, Section 10.4.7.4 Amendment Nos.

139, 139,& 136 4.9-1

5.2 CONTAINMENT Specification The containment for this unit consists of three systems which are the reo:1rtoP building, reactor building isolation system, and penetration roo vent i ilation system.

5.2.1 Reactor Buildin&

The reactor building completely encloses the reactor and its assOciated reactor coolant system.

It is a fully contionuus r.

inforced concrete struclture in the shape of a cylindiC wiLh I

shallow domed roof and flat foundation slab.

The cylindric portion is prestressed by a post tensioning systemt.,i conisiist kl horizontal and vertical tendons.

The dome has a-it w

pose tensioning system. The structure can withstand rhe t 3

horizontal and 3 vertical tendons in the cylinder wall or adjacent tendons in the dome without loss of function.

T12 foundation slab is conventionally reinforced with high str:ngth reinforcing steel.

The entire structure is lined with 1/4" welded steel plate to provide vapor tightness.

The internal volume of the reactor building is approximately 1.91.

x 106 cu. ft.

The approximate inside dimensions are: diiater--

116'; height--208 1/2'.

The approximate thickness of the concrete forming the buildings are: cylindrical wall--3 3/4'; dome---3 1/4';

and the foundation slab--S 1/2'.

The concrete containment structure provides adequate bieooi:=.

shielding for both normal operation and accident situations.

Design pressure and temperature are 59 psig and 2S.60F, r.speetively.

The reactor building is designed for an external at.:seric pressure of 3.0 psi greaccr than the internal. '21u*r:;:1ie.

.L113 greater than the differential pre7sure of 2.5 5nig that cou id developed if the build ing is sealed with an int

nal, of 120 0 F with a barometric pressure of 29.0 inche:; of i i ndth building is subsequently cooled to an internal tumpratura 01.

800 F with a concurrent rise in barometric pressure to 31.0 imche:;

of Hg.

Since the building is designed for this precsure i

ential, vacuum breakers are not required.

Penetration assemblies are seal welded to the reactor build..K liner.

Access openings, electrical penetrations, and fuel for tube covers are equipped with double seals.

Reacto bu idin::

purge penetrations and reactor building atmosphere sampli.::

penetrations are equipped with double valves having r seating surfaces. (1) 5.2-1

The principal design basis for the structure is that it he capable of withstanding the internal pressure resulting from a loss of coolant accident, as defined in FSAR Section 15 with no loss of integrity.

In this event the total energy contained in the water of the reactor coolant system is assumed to be released into the reactor building through a break in the reactor coolant piping.

Subsequent pressure behavior is determined by the building volume, engineered safety features, and the combined influence of energy sources and heat sinks.

5.2.2 Reactor Building Isolation System Leakage through all fluid penetrations not serving accident consequence-limiting systems is to be minimized by a double barrier so that no single, credible failure or malfunction of an active component can result in loss-of-isolation or intolerable leakage.

The installed double barriers take the form of closed piping systems, both inside and outside the reactor building and various types of isolation valves.

(2) 5.2.3 Penetration Room Ventilation System This system is designed to collect, control, and minimize the release of radioactive materials from the reactor building to the environment in post-accident conditions.

It may also operate intermittently during normal conditions as required to maintain satisfactory temperature in the penetrations rooms.

When the system is in operation, a slight negative pressure will be main tained in the penetration room to assure inleakage.

(3)

REFERENCES (1)

FSAR Sections 6.2.1, and 6.2.3 (2)

FSAR Section 6.2.3 (3)

FSAR Section 6.5.1.1 Amendments-Nos.

139, 139, & 135 5.2-2

5.3 REACTOR Specification 5.3.1 Reactor Core 5.3.1.1 The reactor core contains approximately 93 metric tons of slightly enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods.

The reactor core is made up of 177 fuel assemblies, all of which are prepressurized with Helium. (1) 5.3.1.2 The fuel assemblies shall form an essentially cylindrical lattice with an active height of 144 in. and an equiva ent diameter of 128.9 in. (2) 5.3.1.3 There are 61 full-length control rod assemblies (CRA) and 8 axial power shaping rod assemblies (APSR) distributed in the reactor core as shown in FSAR Figure 4.3-3. The full-length CRA and the APSR shall conform to the design described in the FSAR or reload report.

(1) 5.3.1.4 Initial core and reload fuel assemblies and rods shall conform to design and evaluation described in FSAR or reload report and shall not exceed an enrichment of 3.5 percent of U-235.

5.3.2 Reactor Coolant System 5.3.2.1 The design of the pressure components in the reactor coolant system shall be in accordance with the code requirements. (3) 5.3.2.2 The reactor coolant system and any connected auxiliary systems exposed to the reactor coolant conditions of temperature and pres sure, shall be designed for a pressure of 2,500 psig and a tem perature of 650aF. The pressurizer and pressurizer surge line shall be designed for a temperature of 670*F.

(4) 5.3.2.3 The maximum reactor coolant system volume shall be 12,200 ft3.

REFERENCES (1) FSAR, Section 4.2.2 (2) FSAR, Section 4.3.1, and Table 4.3.1 (3) FSAR, Section 5.2.3.1 (4) FSAR Section 5.2.1 Amendments Nos. 13 9

'139 6 135 5.3-1