ML20087B566

Amends 203 & 203 to Licenses DPR-32 & DPR-37,respectively, Increasing Authorized Core Power Level from 2441 Mwt to New Limit of 2546 Mwt

ML20087B566
Person / Time
Site:	Surry
Issue date:	08/03/1995
From:	Russell W NRC (Affiliation Not Assigned)
To:
Shared Package
ML20087B571	List: ML20087B566 ML20087B573
References
NUDOCS 9508080183
Download: ML20087B566 (34)

Text

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-t UNITED STATES

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j NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 2066M001

'+4.....,o i

VIRGINIA ELECTRIC AND POWER COMPANY l

DOCKET NO. 50-280 I

SURRY POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 203 License No. OPR-32 1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated August 30, 1994, as supplemented by letters dated February 6, 1995, February 13, 1995, February 27, 1995, March 23, 1995, March 28, 1995, April 13, 1995, April 20, 1995, April 28,1995, May 5,1995 and June 8,1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; 8.

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

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

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

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

t have been satisfied.

9500080183 950803 PDR ADOCK 05000280 P

PDR

j

, 2.

Accordingly, Facility Operating License No. DPR-32 conditions 3.A. and 3.N. are hereby amended to read as follows:

A.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2546 megawatts (thermal).

N.

Deleted by Amendment No. 203 3.

Further, 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-32 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 203, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

4.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days following completion of the Cycle 13/14 refueling outage, currently scheduled for October 1995.

FOR THE NUCLEAR REGULATORY COMMISSION William T. Russell, Director Office of Nuclear Reactor Regulation Attachments:

1 1.

Pages 3 and Sa of License 2.

Changes to the Technical Specifications Date of Issuance:

August 3, 1995 3

1

-32 3.

This license'shall be deemed to contain and is subject to the conditions specified in the followl Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30,

'on 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70, and is subject to all appucable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified '

below:

i A.

Maximum Power Level The licensee is authorized.to operate the facility at steady state reactor i

core power levels not in excess of 2546 megawatts (thermal).

f B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 203, are hereby incorporated in the license. The licensee i

shall operate the facility in accordance with the Technical Specifications.

i C.

Reoorts The licensee shall make certain reports in accordance with the l l

requirements of the Technical Specifications.

D.

Records i

The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.

E.

Deleted by Amendment 65 v

i Surry - Unit 1 Amendment No. 203

- Sa -

K.

Secondary Water Chemistry Monitoring Procram The licensee shall implement a secondary water chemistry monitoring propram to inhibit steam generator tube degradation. This program sha 1 include:

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1.

Identification of a sampling 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 chemistry conditions; and 1

6.

A procedure for identifying the authority responsible for the interpretation of the data, and the sequence and timing cf administrative events required to initiate corrective action.

L.

The licensee shall fully implement and maintain in effect all provisions of the Commission approved Nuclear Security Personnel Training and Quali3 cations Program, including amendments and changes made pursuant to 10 CFR 50.54(p). The approved Nuclear Security Personnel Training and Qualifications Program consists of a document withheld from public disclosure pursuant to 10 CFR 2.790(d) identified as "Surry Power Station Nuclear Security Personnel Training and Qualifications Program" dated September 15,.1980. The Nuclear Security Personnel Training and Qualifications Program shall be fully implemented in accordance with 10 CFR 73.55(b)(4), within 60 days of this approval by the Commission. All security personnel shall be qualified within two years of this approval.

M.

The design of the reactor coolant pump and steam generator supports may be revised in accordance with the licensee's submittals dated November 5, 1985 (Serial No.85-136), December 3,1985 (Serial No. 85-136A), and January 14,1986 (Serial No. 85-136C).

N.

Deleted by Amendment No. 203 l

4.

This license is effective as of the date of issuance,' and shall expire at midnight on May 25, 2012.

FOR THE ATOMIC ENERGY COMMISSION Original signed by A. Giambusso A. Giambusso, Deputy Director for Reactor Projects Directorata of Ucensing Enclosure Appendix A -

i Technical Specifications Date of issuance: May 25,1972 Surry - Unit 1 Amendment No. 203 y

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i UNITED STATES g

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NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 2006H1001

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VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 203 License No. DPR-37 1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated August 30, 1994, as supplemented by letters dated February 6, 1995,-February 13, 1995, February 27, 1995, March 23, 1995, March 28, 1995, April 13, 1995, April 20, 1995, April 28, 1995 and May 5, 1995 and June 8, 1995, complies with the standards and requirements of the. Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will oper' ate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

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

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

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

l l

1 l

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3

i

' 2.

Accordingly, Facility Operating License No. DPR-37 conditions 3.A. and 3.N. are hereby amended to read as follows:

A.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2546 megawatts (thermal).

1 N.

Deleted by Amendment 3.

Further, 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-32 is hereby amended to read as follows:

B.

Technical Soecifin=tigni The Technical Spicifications contained in Appendix A, as revised through Amendment No. 203, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

4.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMISSION I

William T. Russell, Director Office of Nuclear Reactor Regulation Attachments:

1.

Pages 3 and 6a of License 2.

Changes to the Technical Specifications Date of Issuance: August 3,1995 1

, of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A.

Maximum Power Level The licensee is authorized to operate the. facility at steady state reactor core power levels not in excess of 2546 megawatts (thermal).

B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 203, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Sp1K:ifications.

c.

Banar_tt The ficansee shall make certain reports in accordance with the requirernents of the Technical Specifications.

D.

Records The licenses shall keep facility operating records in accordance with the requirements of the Technical Specifications.

E.

Deleted by Amendment 54

]

s a

Surry - Unit 2 Amendment No. 203 i

- 6a -

K.

Secondary Water Chemistry Monitoring Program The licensee shall implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation. This program-sha:1 include:

1.

Identification of a sampling 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; J

4.

Procedure for the recording and management of data; 5.

Procedures defining corrective actions for off control

-l point chemistry conditions; and 6.

A procedure for identifying the authority responsible for the interpretation of the data, and the sequence and timing cf administrative events required to initiate corrective action.

J l

L The licensee shall fully implement and maintain in effect all provisions of the Commission-approved Nuclear Security Personnel' Training' and Qualifications Program, including amendments and changes made pursuant to 10 CFR 50.54(p). The approved Nuclear Security Personnel l

Training and Qualifications Program consists of a documet withheld from public disclosure pursuant to 10 CFR 2.790(d) identified as "Surry Power i

Station Nuclear Security Personnel Training and Qualifications Program

• dated September 15,1980. The Nudear Security Personnel Training and Qualifications Program shall be fully implemented in accordance with 10 CFR 73.55(b)(4), within 60 days of this approval by the Commission. All security personnel shall be qualified wkhin two years of this approval.

l M.

The design of the reactor coolant pump and steam generator supports may be revised in accordance with the licensee's submittals dated November 5, 1985 (Serial No.85-136), December 3,1985 (Serial No. 85-136A), and January 14,1986 (Serial No. 85 1360).

N.

Deleted by Amendment No.-203 l

4.

This license is effective as of the date of issuance, and shall expire e midnight on January 29,2013.

FOR THE ATOMIC ENERGY COMMISSION -

t Original signed by Roger Boyd/for A. Giambusso, Deputy Director for Reactor Projects Directorate of Uctensing Enclosure Appendix A -

Technical Specifications Date of issuance: January 29,1973 Surry - Unit 2 Amendment No. 203

4 ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 203 TO FACILITY OPERATING LICENSE NO. DPR-32 AMENDMENT NO. 203 TO FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NOS. 50-280 AND 50-281 Revise Appendix A as follows:

Remove Paaes Insert Paaes TS 1.0-1 TS 1.0-1 TS 2.1-3 TS 2.1-3 TS 2.1-4 TS 2.1-4 TS 2.1-5

.TS 2.1-5 TS Figure 2.1-1 TS Figure 2.1-1 TS 2.2-2 TS 2.2-2 TS 2.3-2 TS 2.3-2 1

TS 2.3-7 TS 2.3-7 TS 3.1-1 TS 3'.1-1 TS 3.1-3 TS 3.1-3 TS 3.1-16 TS 3.1-16 TS 3.1-17 TS 3.1-17 TS 3.1-17a TS 3.1-17a TS 3.3-5 TS 3.3 TS 3.6-2 TS 3.6-2 TS 3.6-4 TS 3.6-4 TS 3.6-5 TS 3.6-5 TS 3.7-26 TS 3.7-26 TS 3.8-4 TS 3.8-4 TS Figure 3.8-1 TS Figure 3.8-1 TS 3.10-7 TS'3.10-7 TS 3.12-12 TS 3.12-12 TS 4.1-10a TS 4.1-10a TS 4.4-3 TS 4.4-3 TS 5.2-3 TS 5.2-3 I

TS 1.0-1 1.0 DEFINITIONS The following frequently used terms are defined for the uniform interpretation of the specifications.

A.

RATED POWEFi A steady state reactor core heat output of 2546 MWt.

l B.

THERMAL POWER l

The total core heat transferred from the fuel to the coolant.

C.

REACTOR OPERATION I

I 1.

REFUELING SHUTDOWN l

When the reactor is subcritical by at least 5% Ak/k and Tavg s i

s140 F and fuel is scheduled to be moved to or from the reactor core.

2.

COLD SHUTDOWN When the reactor is suberitical by at least 1% Ak/k and Tavg s i

s200*F.

3.

INTERMEDIATE SHUTDOWN When the reactor is suberitical by at I, east 1.77% Ak/k and 200*F

< Tavg < 547"F.

4.

HOT SHUTDOWN I

When the reactor is suberitical by at least 1.77% Ak/k and Tavg is f

2 547"F.

i Amendment Nos. 203 and 203

d TS 2.13

- uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the DNB heat flux at a particular core location to the local heat flux, is indicative of the margin to DNB. The DNB basis is as -

follows: there must be at least a 95% probaNiity with 95% confidence that the minimum DNBR of the limiting rod during Condition I and 11 events is greater than or equal to the DNBR limit of the DNB correlation being used. The correlation DNBR limit is based on the entire applicable experimental data set to meet this statistical criterion.(1)

The curves of TS Figure 2.1-1 which show the allowable power level decreasing with increasing temperature at selected pressures for constant flow (three loop operation) represent limits equal to, or more conservative than, the-f loci of points of thermal power, coolant system everage temperature, and coolant system pressure for which the calculated DNBR is not less than the design DNBR limit or the average enthalpy at the exit of the vessel is equal to the saturation value. The area where clad integrity is assured is below these lines. The temperature limits are considerably more conservative than would be required if they were based upon the design DNBR limit alone but are such that the plant conditions required to violate the limits are precluded by the self-actuated safety valves on the steam generators. The effects of rod bowing are l

also considered in the DNBR analyses.

P Amendment Nos. 203 and 203

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TS 2.1-4 g +

TS Figure 2.1-1 is based on a 1.55 cosine axial flux shape and a statistical.

treatment of key DNBR analysis parameter uncertainties including an enthalpy rise hot channel factor which follows the following functional form: FAH(N) =

1.56 [1 + 0.3(1-P)] where P is the fraction of RATED POWER. The limits include margin to accommodate rod bowing.(1) TS Figures 2.1-2 and 2.1-3 are based on an FAH(N) of 1.55, a deterministic treatment of key DNS analysis parameter uncertainties, and include a 0.2 rather than 0.3 part power multiplier for the enthalpy rise nct ehennel factor. The FAH(N) limit presented in the unit-and reload-specific CORE OPERATING LIMITS REPORT is confirmed for each reload to be accommodated by the Reactor Core Safety Limits.

These hot channel factors are higher than those calculated at full power over the range between that of all control rod assemblies I

Amendment Nos. 203 and 203 -

TS 2.1-5 fully withdrawn to maximum allowable control rod assembly insedion. The control rod assembly insertion limits are covered by Specification 3.12.

Adverse power distribution factors could occur at lower power levels because additional control rod assemblies are in the core; however, the control rod assembly insertion limits as specified in the CORE OPERATING LIMITS REPORT ensure that the DNBR is always greater at partial power than at full power.

The Reactor Control and Protection System is designed to prevent any anticipated combination of transient conditions for Reactor Coolant System temperature, pressure and thermal power level that would result in a DNBR less than the design DNBR limit (3) based on steady state nominal operating power levels less than or equal to 100%, steady state nominal operating Reactor Coolant System average temperatures less than or equal to 573.0 F and a steady state nominal operating pressure of 2235 psig. For deterministic DNBR analysis, allowances are made in initial conditions assumed for transient analyses for steady state errors of +2% in power, +4 F in Reactor Coolant System average temperature and 30 psi in pressure. The combined steady state errors result in the DNB ratio at the start of a transient being 10 percent less than the value at nominal full power operating conditions.

For statistical DNBR analyses, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically.coch that there is at least a 95% probability that the minimum DNBR for the limiting r.1 is greater than or equal to the statistical DNBR limit. The uncertainties in the plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a statistical DNBR limit which must be met in plant safety analyses using values of input parameters without uncertainties. The statistical DNBR limit also Amendment Nos. 203 and 203

i TS FIGURE 2.1-1 REACTOR CORE THERMAL AND HYDRAULIC SAFETY LIMITS THREE LOOP OPERATION,100% FLOW 670.0 660.0 N

l 650.0 -

N N

2sespois N

640 0 N

N l ! 630.0 -

l P N

N K

Li N

N I

620 g0 "w

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g h 600.0 -

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T 590 0 580.0 -

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570.0 560.0 550.0 0

10 20 30 40 50 60 70 80 90 100 110- 120 Percent of Rated Thermal Power L

Amendment Nos. 203 and 203

,,n.

TS 2.2 2 The nominal settings of the power-operated relief valves at 2335 psig, the reactor high pressure trip at 2385 psig and the safety valves at 2485 l

psig are established to assure never reaching the Reactor Coolant System pressure safety limit. The initial hydrostatic test has been conducted at 3107 psig to assure the integrity of the Reactor Coolant System.

l 1)

UFSAR Section 4 2)

UFSAR Section 4.3 4

Amendment Nos. 203 and 203

. TS 2.3-2 t

(b)

High pressurizer pressure - s 2385 psig.

l l

(c)

Low pressurizer pressure - 21860 psig.

(d)

Over1emperature AT g

I-i 1+ t s i

AT5 AT,[K - K (1 t2s) (T-T) + K (P - P)- f(Al)]

g i

2 3

where I

AT, - Indicated AT at rated thermal power, 'F T - Average coolant temperature, 'F T =573.0*F l

P - Pressurizer pressure, psig i

P' - 2235 psig K - 1.135 i

K - 0.01072 2

K - 0.000566 3

g

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Al _qt

• A, where q and qb are the percent power in the top and bottom halves of l

b g

the core respectively, and q + qb s total core power in percent of rated l

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power f(Al) - function of Al, percent of rated core power as shown in Figure 2.31 t - 25 seconds i

t - 3 seconds 2

t I

(e) Overpower AT ts 3

AT s AT,[K - K (1.13s)T - Ks (T - T) - f(AI)]

4 5

i Amendment Nos. 203 and 203 s

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TS 2.3 :

- The low flow reactor trip protects the core against DNB in the event of a sudden

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loss of power to one or more reactor coolant pumps. The undervoltage reactor

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trip protects against a decrease in Reactor Coolant System flow caus' d by a j

e loss of voltage to the reactor coolant pump busses. The underfrequency reactor trip (opens RCP supply breakers and) protects against a decrease in Reactor Coolant System flow caused by a frequency decay on the reactor coolant pump busses. The undervoltage and underfrequency reactor trips are expected to occur prior to the low flow trip setpoint being reached for low flow events caused t

by undervoltage or underfrequency, respectively.

The accident analysis conservatively ignores the undervoltage and underfrequency trips and assumes reactor protection is provided by the low flow trip. The undervoltage and i

underfrequency reactor trips are retained as back-up protection.

The high pressurizer water level reactor trip protects the pressurizer safety 3

valves against water relief. Approximately 1154 ft of water corresponds to 92%

of span. The specified setpoint allows margin for instrument error (7) and transient level overshoot beyond this trip setting so that the trip function prevents the water level from reaching the safety valves.

1 The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents. The specified setpoint assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the Auxiliary Feedwater System.(7)

The specified reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal unit operations. The prescribed setpoint above which these trips are unblocked assures their availability in the power range where needed.

4 Above 10% power, an automatic reactor trip will occur if two or more reactor coolant pumps are lost. Above 50%, an automatic reactor trip will occur if any l

pump is lost or de-energized. This latter trip Amendment Nos. 203 and 203

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TS 3.1-1 3.1 REACTOR COOLANT SYSTEM b

Aoplicability Applies to the operating status of the Reactor Coolant System.

Objectives To specify those limiting conditions for operation of the Reactor Coolant System which must be met to ensure safe REACTOR OPERATION.

l

.i These conditions relate to: operational components, heatup and cooldown, leakage, reactor coolant activity, oxygen and chloride concentrations, minimum temperature for criticality, and Reactor Coolant System overpressure mitigation.

l A.

Ooerational Components Specifications

)

l 1.

Reactor Coolant Pumps a.

A reactor shall not be brought critical with less than three l

pumps, in non-isolated loops, in operation.

Amendment Nos. 203 and 203

.., - - -~

TS 3.1-3 i

When all three pumps have been idle for > 15 minutes, the first e.

pump shall not be started unless: (1) a bubble exists in the 1

pressurizer or (2) the secondary water temperature of each steam generator is less than 50 F above each of the RCS cold leg temperatures.

2.

Steam Generator A minimum of two steam generators in non-ir,olated !oop shall be OPERABLE when the average Reactor Coolar.t System temperature is greater than 350 F.

3.

Pressurizer Safety Valves a.

Three valves shall be OPERABLE when the head is on the reactor vessel and the Reactor Coolant System average temperature is greater than 350 F, the reactor is critical, or the Reactor Coolant System is not connected to the Residual Heat Removal System.

b.

Valve lift settings shall be maintained at 2485 psig 1 percent

• The as-found tolerance shall be 13% and the as left tolerance shall be 11%.

Amendment Nos. 203 and 203 i

.2.

a

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l TS 3.1-16 BA&la The specified limit provides protection to the public against the potential release of reactor coolant activity to the atmosphere, as demonstrated by the following analysis of a steam generator tube rupture accident in UFSAR Chapter 14.3.1.

l Rupture of a steam generator tube would allow radionuclides in the reactor coolant to enter the secondary system. The limiting case involves a double-ended tube rupture coincident with loss of the condenser and release of steam from the secondary side to the atmosphere via the main steam safety valves or atmospheric relief valves. This is assumed to continue for 30 minutes in the analysis. The operator will take action to reduce the primary side temperature to a value below that corresponding to the relief or safety valve setpoint. Once this is accomplished the valves can be closed and the release terminated.

i i

I 4

1 Amendment Nos. 203 and 203

.4 m

TS 3.1-17 Permitting startup and/or REACTOR OPERATION to continue for limited time periods with the reactor coolant's specific activity 21.0 Ci/cc but < 10.0_ pCi/oc DOSE -

EQUIVALENT l-131 accommodates possible lodine spiking phenomena'which may

)

occur following changes in THERMAL POWER. Although the analysis of a steam j

generator tube rupture initiated with primary coolant activity at the 10.0 pCi/cc transient i

limit shows offsite doses well within the 10 CFR 100 limits, operation at the transient i

limit is restricted to no more than 10 percent of the unit's yearly operating time to limit the risk of appreciable releases following a postulated steam generator tube rupture.

l The basis for the 500'F temperature contained in the Specification is that the saturation pressure corresponding to 500*F, i.e., 680.8 psia, is well below the pressure - l at which the atmospheric relief valves on the secondary side could be actuated.

The accident analysis examines two cases of iodine spiking. For the case with a pre existing lodine spike, the transient coolant activity limit of 10.0 pCi/cc is assumed. For the case of a concurrent spike, the initial activity is assumed to correspond to the steady state limit of 1.0 Ci/cc. The concurrent iodine spike is modeled with a conservative iodine appearance rate. Both cases show doses at the exclusion area and low population zone boundaries which are well within the 10 CFR Part 100 limits

)

and control room doses which are within the General Design Critorier (GDC) 19 guidelines.

I Measurement of E will be performed at least twice annually. Calculations required to determine E will consist of the following:

1.

E shall be the average (weighed in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, j

other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

2.

• A determination of the beta and gamma decay energy per disintegration of each nuclide determined in (1) shcve by applyiry known decay energies 'and-schemes.

3.

A calculation of 5 by appropriate weighing of each nuclide's beta and gamma 4

energy with its concentration as determined in (1) above.

Amendment Nos. 203 and 203

TS 3.1-17e DOSE EQUIVALENT l 131 shall be that concentration of I-131 (pCi/cc) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132,1-133,1-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be either: a) those listed in l

Table lli of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites", or b) Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Compliance i

with 10 CFR Part 50, Appendix 1."

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Amendment Nos. 203 and 203

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l TS 3.3-5 The accumulators (one for each loop) discharge into the cold leg of the reactor coolant piping when Reactor Coolant System pressure decreases below i

accumulator pressure, thus assuring rapid core cooling for large breaks. The line from each accumulator is provided with a motor-operated valve to isolate j

the accumulator during reactor start-up and shutdown to preclude the discharge

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of the contents of the accumulator when not required.

Accumulator Motor Ooerated Discharoe isolation Valvet

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' Unit No.1

' Unit No. 2 MOV 1865A MOV 2865A MOV 1865B MOV 28658 j

MOV 1865C MOV 2865C However, to assure that the accumulator valves satisfy the single failure criteria, they will be locked, sealed or otherwise secured open by de-energizing the valve motor operators when the reactor coolant pressure exceeds 1000 psig.

The operating pressure of the Reactor Coolant. System is 2235 psig'and l

accumulator injection is initiated when this pressure drops to 600 psia. ' De-l energizing the motor operator when the pressure exceeds 1000 psig allows sufficient time during normal startup operation to perform the actions required to de-energize the valve. This procedure will assure that there is an OPERABLE f

flow path from each accumulator to the Reactor Coolant System during POWER i

OPERATION and that safety injection can be accomplish'ed.-

The removal of power from the valves listed above will assure that the systems of which they are a part satisfy the single failure criterion.

i i

i.

4 i

Amendment Nos. 203 and 203

l TS 3.6-2 l

.i 2.

A minimum of 96,000 gallons of water shall be available in the protected l

condensate storage tank to supply emergency water to the auxiliary feedwater pump suctions. A minimum of 60,000 gallons of water shall be i

available in the protected condensate storage tank of the opposite unit to l

l supply emergency water to the auxiliary feedwater pump suction of that l

unit.

l 3.

All main steam line code safety valves, associated with steam generators in unisolated reactor coolant loops, shall be OPERABLE with lift settings l

as specified in Table 3.6-1 A and 3.6-18.

C.

Prior to reactor power exceeding 10%, the steam driven auxiliary feedwater pump shall be OPERABLE.

l D.

System piping, valves, and control board indication required for operation of the components enumerated in Specifications 3.6.B.1, 3.6.B.2, 3.6.B.3, and 3.6.C l

shall be OPERABLE (automatic initiation instrumentation associated with the opposite unit's auxiliary feedwater pumps need not be OPERABLE).

j i

E.

The specific activity of the secondary coolant system shall be s 0.10 pCi/cc l

f DOSE EQUIVALENT l-131. If the specific activity of the secondary coolant j

system exceeds 0.10 Ci/cc DOSE EQUIVALENT l-131, the reactor shall be shut down and cooled to 500'F or less within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after detection and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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Amendment Nos. 203 and 203 i

TS 3.6-4 The capability to supply feedwater to the generators is normally provided by the operation of the Condensate and Feedwater Systems. In the event of complete loss of electrical power to the station, residual heat removal would continue to be assured by the availability of either the steam driven auxiliary feedwater pump or one of the motor driven auxiliary feedwater pumps and the 110,000-gallon protected condensate storage tank. In the event of a fire or high energy line break which would render the auxiliary feedwater pumps inoperable on the affected unit, residual heat removci would continue to be assured by the availability of either the steam driven auxiliary feedwater pump or one of the motor-driven auxiliary feedwater pumps from the opposite unit. A minimum of two auxiliary feedwater pumps are required to be aperable* on the opposite unit to ensure compliance with the design basis accident analysis assumptions, in that auxiliary feedwater can be delivered via the cross-connect, even if a single active failure results in the loss of one of the two pumps.

The specified minimum water volume in the 110,000-gallon protected condensate storage tank is sufficient for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of residual heat removal following a reactor trip and loss of all offsite electrical power. It is also sufficient to maintain one unit at hot shutdown for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, followed by a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> cooldown from 547 F to 350 F (i.e., RHR operating conditions). If the protected condensate storage tank level is reduced to 60.000 gallons, the immediately available replenishment water in the 300,000-gallon condensate tank can be gravity-fed to the protected tank if required for residual heat removal. An attemate supply of feedwater to the auxiliary feedwater pump suctions is also available from the Fire Protection System Main in the auxiliary feedwater pump cubicle.

The five main steam code safety valves associated with each steam generator have a total combined capacity of 3,842,454 pounds per hour at their individual relieving l

pressure; the total combined capacity of all fifteen main steam code safety valves is 11,527.362 pounds per hour. The nominal power rating steam flow is 11,260,000 pounds per hour.

The combined capacity of the safety valves required by Specification 3.6 always exceeds the total steam flow corresponding to the maximum steady state power than can be obtained during three reactor coolant loop operation.

excluding automatic initiation instrumentation 1

Amendment Nos. 203 and 203 1

TS 3.6 5 The availability of the auxiliary feedwater pumps, the protected condensate storage tank, and the main steam line safety valves adequately assures that sufficient residual heat removal capability will be available when required.

The limit on steam generator secondary side iodine - 131 activity is based on limiting the inhalation dose at the site boundary following a postulated steam line break accident to a small fraction of the 10 CFR 100 limits. The accident analysis, which is performed based on the guidance of NUREG-0800 Section 15.1-5, assumes the release of the entire contents of the faulted steam generator to the atmosphere.

Amendment Nos. 203 and 203

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e TS 3.8-4 (3) assuring that environmental conditions will not preclude ' access to close the valves and 4) that this administrative or manual action will prevent the release of radioactivity outside the containment.

The Reactor Coolant System temperature and pressure being below 350'F and 450 psig, respectively, ensures that no significant amount of flashing steam will I

be formed and hence that there would be no significant pressure buildup in the l

containment if there is a loss-of-coolant accident. Therefore, the containment intomal pressure is not required to be subatmospheric prior to exceeding 350'F and 450 psig.

The allowable value for the containment air partial pressure is presented in TS Figure 3.8-1 for service water temperatures from 25 to 95'F. The RWST water l

[

shall have a maximum temperature of 45'F.

4 The horizontal limit line in TS Figure 3.8-1 is based on LOCA peak calculated l

pressure criteria, and the sloped line is based on LOCA subatmospheric peak i

pressure criteria.

i i

i Amendment Nos. 203 and 203

TS FIGURE 3.8-1 SURRY TECHNICAL SPECIFICATION CURVE MAX CONTAINMENT ALLOWABLE AIR PARTIAL PRESSURE INDICATION VS. SW i

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___-_.____1_-_-_-_

i TS 3.10-7 Upon each completion of core loading and installation of the reactor vessel head, specific mechanical and electrical tests will be performed prior to initial' criticality.

The fuel handling' accident has been analyzed based on the methodology outlined in Regulatory Guide 1.25. The analysis assumes 100% of the gap activity from the highest powered assembly is released after a 100-hour decay-period following operation at 2605 MWt.

Detailed procedures and checks insure that fuel assemblies are loaded in the proper locations in the core. As an additional check, the movable incore i

1 detector system will be used to verify proper power distribution. This system is j

capable of revealing any assembly enrichment error or loading error which could cause power shapes to be peaked in excess of design value.

)

References i

i UFSAR Section 5.2 Containment isolation UFSAR Section 6.3 Consequence Limiting Safeguards UFSAR Section 9.12 Fuel Handling System UFSAR Section 11.3 Radiation Protection UFSAR Section 13.3 Table 13.3-1 l

UFSAR Section 14.4.1 Fuel Handling Accidents'

{

FSAR Supplement:

Volume I: Question 3.2 l

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Amendment Nos. 203 and 203 E

i

. -, - + -,

TS 3.1212 1

3.

If more than one rod position indicator channel per group or two rod position indicator channels per bank are inoperable during control bank motion to achieve criticality or POWER OPERATION, then the unit shall be placed in HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

F.

DNB Parameters 1.

The following DNB related parameters shall be maintained within theirlimits during POWER OPERATION-l Reactor Coolant System Tavg s; 577.0*F Pressurizer Pressure 2 2205 psig Reactor Coolant System Total Flow Rate 2 273,000 gpm

a. The Reactor Coolant System Tavg and Pressurizer Pressure shall be verified to be within their limits at least once overy 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

I

b. The Reactor Coolant System Total Flow Rate shall be determined to be within its limit by measurement at least once per refueling cycle.

2.

When any of the parameters in Specification 3.12.F.1 has been i

determined to exceed its limit, either restore the parameter 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 POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

3.

The limit for Pressurizer Pressure in Specification 3.12.F.1 is not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED POWER per minute or a THERMAL POWER step increase in excess of 10% of RATED POWER.

Amendment Nos. 203 and 203

,~

TS 4.1 102

+

(3) 5 determination will be started when the gross gamma degassed activity of radionuclides with half-lives greater than 15 rniputes analysis indicates 1 10 Ci/cc. Routine sample (s) for E analyses shall only be taken after a minimum of 2 EFPD and 20 days of power operation have elapsed since reactor was last suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

(4)

If the fifteen minute degassed beta and gamma activity is 10% or more of the limit given in Specification 3.6.E, a DOSE EQUlVALENT l-131 analysis will be l

performed.

(5)

When reactor is critical and average primary coolant temperature 2: 350*F.

l (6)

Whenever the specific activity exceeds 1.0 pCi/cc DOSE EQUlVALENT l-131 or 100/5 pCi/cc and until the specific activity of the Reactor Coolant System is l

restored within its limits.

[

(7)

One sample between 2 & 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a THERMAL POWER change exceeding 15 percent of the RATED POWER within a one hour period provided the average primary coolant temperature 2350*F.

(8)

When the fifteen minute degassed beta and gamma activity is less than 10% of the limit given in Specification 3.6.E.

l Amendment Nos. 203 and 203

~ TS 4.4-3 i

The containment is designed-for a maximum pressure of 45 psig.

The containment is maintained at a subatmospheric air partial pressure consistent with TS Figure 3.8-1 depending upon the cooldown capability _of the Engineered Safeguards and will not rise cbove 45 psig for any postulated loss-j of-coolant accident.

The initial test pressure for the Type A test is 47.0 psig to allow for containment expansion and equalization. A review was performed to determine the effects j

of pressurizing containment above its design pressure of 45.0 psig. This review was based on the original containment test at 52 poig. During that test, the i

calculated stresses were found to be well within the allowable yield strength of the structural reinforcing bars, therefore performance of the Type A test at 47 psig will have no detrimental effect on the containment structure.

~

All loss-of-coolant accident evaluations have been based on an integrated l

containment leakage rate not to exceed 0.1% of containment volume per 24 hr.

l 1

The above specification satisfies the conditions of 10 CFR 50.54(o) which stated '

that primary reactor containments shall meet the containment leakage test requiremefits set foith in Appendix J.

The limitations on closure and leak rate for the containment airlocks are required to meet the restrictions on containment integrity and containment leak rate. Surveillance testing of the airlock seals provides assurance that the overall airlock leakage will not become excessive due to seal damage during the intervals between airlock leakage tests.

i i

References UFSAR Section 5.4 Design Evaluation of Containment Tests and Inspections of l

i Containment UFSAR Section 7.5.1 Design Bases of Engineered Safeguards Instrumentation UFSAR Section 14.5 Loss of Coolant Accident 10 CFR 50 Appendix J

" Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors" I

Amendment Nos. 203 and 203 a

w n--

n TS 5.2-3 ~

l a-b.

A high containment pressure isolation signal closes the automatic trip valves in all normally open lines penetrating the containment which are not required to be open to control containment pressure to perform an orderly reactor shut down without actuation of the consequence limiting safeguards in case of a small Reactor Coolant System leak.

A further rise in containment pressure, indicating a major loss-of-c.

coolant accident, produces a containment high high pressure isolation signal which closes all normally open lines which penetrate the containment which have not been closed by 2-b above.

d.

Isolation can be accomplished manually from the control in the Main Control Room if any of the automatic signals fail to actuate the above valves.

C.

Containment Systems t

1.

Following a loss-of-coolant accident, the Containment Spray Subsystems distribute at least 2,S00 gpm borated water spray containing sodium hydroxide for iodine removal within the

.i containment atmosphere. The Recirculation Spray Subsystems recirculate at least 3,000 gpm of water from the containment sump.

i Amendment Nos. 203 and 203

-