Text

Amends 95 & 95 to Licenses NPF-72 & NPF-77,respectively, Changing TS Limits on RCS Dei
	ML20151T200
Person / Time
Site:	Braidwood
Issue date:	09/03/1998
From:	Stewart Bailey; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20151T204	List: ML20151T200; ML20151T207;
References
	NUDOCS 9809090290
	Download: ML20151T200 (13)
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,j NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 3000H001 s...../

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COMMONWEALTH EDISON COMPANY i

DOCKET NO. STN 50-456 j

d BRAIDWOOD STATION. UNIT No.1 l

AMENDMENT TO FACILITY OPERATING LICENSE 4

Amendment No. 95 1

License No. NPF-72 7

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-1.

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

i A.

The application for amendment by Commonwealth Edison Company (the licensee) t dated January 14,1998, as supplemented by letter dated July 17,1998, 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 4

i can be conducted without endangering the health and safety of the public, and (ii) i 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 i

security or to the health and safety of the public; and 1

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.

1' 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility i

j Operating License No. NPF 72 is hereby amended to read as follows:

9309090290 990903 PDR ADOCK 05000456 P

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2-i (2) Technical Soecifications The Technical Specifications contained in Appendix A as revised through i

Amendment No. 95 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this i

license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 30 days.

i FOR THE NUCLEAR REGULATORY COMMISSION Stewart N. Bailey, Proje r

Project Directorate 111-2 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: September 3, 1998 i

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I COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-457 BRAIDWOOD STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 95 i

License No. NPF-77 1.

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

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated January 14,1998, as supplemented by letter dated July 17,1998, 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 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:

l

. (2) Technical Specifications

- The Technical Specifications contained in Appendix A as revised through Amendment No. 95 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-72, dated July 2, 1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date if its issuance and shall be impicmented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION Stewart N. Bailey, Proje a

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Project Directorate ill-2 l

Division of Reactor Projects -Ill/IV

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Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical l

Specificatbns i

Date of issuance: September 3,1998 i-L I

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ATTACHMENT TO LICENSE AMENDMENT NOS. 95 AND 95 FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET NOS. STN 50 456 AND STN 50-457 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change. Pages marked with an asterisk are provided for convenience.

l Remove Panes Insert Paaes 3/4 4-27 3/4 4-27 3/4 4-28 3/4 4-28 3/4 4-29 3/4 4-29 3/4 4-30*

3/4 4-30*

3/4 4-31 3/4 4-31 B 3/4 4-5 B 3/4 4-5 B 3/4 4-6*

B 3/4 4-6*

B 3/4 4-7*

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REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY

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LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:

a.

Less than or equal to 1 microcurie per gram DOSE EQUIVALENT l-131",'and b.

Less than or equal to 100/5 microCuries per gram of gross radioactivity.

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

ACTION-MODES 1,2 and 3*:

a.

With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUlVALENT l-131" for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T less than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and y

b.

With the specific activity of the reactor coolant greater than 100/5 microCuries per gram, be in at least HOT STANDBY with T less than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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With T, greater than or equal to 500*F.-

"For Unit 1 through Cycle 7, reactor coolant DOSE EQUlVALENT l-131 will be limited to 0.05 microCuries per gram.

l BRAIDWOOD-UNITS 1 & 2 3/4 4-27 AMENDMENT NO. 95

l REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION l

ACTION (Continued) l MODES 1,2,3,4, and 5:

With the specific activity of the reactor co2iant greater than 1 microcurie per gram DOSE EQUIVALENT l-131* or greater than 100/E microCuries per gram, perform the sampling l

and analysis requirements of item 4.a) of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits.

SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by l

performance of the sampling and analysis program of Table 4.4-4.

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For Unit 1 through Cycle 7, reactor coolant DOSE EQUIVALENT l-131 will be limited to 0.05 microCurles per gram.

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i BRAIDWOOD-UNITS 1 & 2 3/4 4-28 AMENDMENT NO. 95

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'20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCi/ GRAM DOSE EQUlVALENT l-131*

For Unit 1 through Cycle 7. Reactor Coolant Specific Activity > 0.05 pCi/ Gram DOSE EQUIVALENT l-131.

l BRAIDWOOD. UNITS 1 & 2 3/4 4-29 AMENDMENT NO. 95

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TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AM ANALYSIS PROGRAM TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WICH SAMPLE 4

AND ANALYSIS FREQUENCY AM ANALYSIS REQUIRED l

1.

Gross Radioactivity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1* 2* 3* 4 Determination **

1 2.

Isotopic Analysis for DOSE EQUIVA-Once per 14 days 1

LENT l-131 Concentration i

3.

Radiochemical for E Determination ***

Once per 6 months

1 i

4.

Isotopic Analysis for Iodine a)

Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , If, 2#, 3#, 4#, 5#

Including I-131 I-133, and I-135 whenever the specific i

activity exceeds 1 4

pC1/ gram DOSE EQUIVALENT I-131****

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or 100/E pC1/ gras of gross radioactivity, i

and b)

One sample between 2 1, 2, 3 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change I

exceeding 15% of l

the RATED THERMAL POWER within a 1-hour period.

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BRAIDN000 - UNITS 1 & 2 3/4 4-30 AMEN 0 MENT NO. 69 G

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TABLE 4.4-4 (Continued)

TABLE NOTATIONS Until the specific activity of the Reactor Coolant System is restored within its limits.

Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the reactor coolant except for radior'uclides with half-lives less than 10 minutes and all radiolodines. The total specific activity shall be the sum of the degassed beta-gamma activity and the total of all identified gaseous activities in the sample within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the sample is taken and extrapolated back to when the sample was taken.

Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level. The latest available data may be used for pure beta-emitting radionuclides.

'" A radiochemical analysis for E shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radioiodines, which is identified in the reactor coolant. Thq., specific activities for these individual radionuclides shall be used in the determination of E for the reactor coolant sample.

Determination of the contributors to E shall be based upon these energy peaks identifiable with a 95% confidence level.

"" For Unit 1 through Cycle 7, reactor coolant DOSE EQUIVALENT l-131 will be limited to 0.05 l

microCuries per gram.

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BRAIDWOOD-UNITS 1 & 2 3/4 4-31 AMENDMENT No. 95

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t REACTOR COOLANT SYSTEM BASES j-OPERATIONAL LEAKAGE (Continued)

The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent j

intersystem LOCA. Leakage from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or i

failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structuralintegrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride l

limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structuralintegrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady-State Limits.

The Surveillance Requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.8 SPECIFIC ACTIVITY '

The limitations on the specific activity of the reactor coolant ensure that the resulting 2-hour doses at the SITE BOUNDARY will not exceed an appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-state reactor-to-secondary steam ganerator leakage rate of 1 gpm. For Unit 1 through Cycle 7, the limitations on the specific activity of the reactor coolant ensure that the

. resulting 2-hour off site doses will not exceed an appropriately small fraction of the 10 CFR Part 100 dose guideline values following a Main Steam Line Break accident in conjunction with an assumed steady-state primary-to secondary steam generator leakage rate of 150 gpd from each of the unfaulted steam generators and maximum site allowable primary-to-secondary leakage from l

the faulted steam generator. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Braidwood Station, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.

BRAIDWOOD-UNITS 1 & 2 B 3/4 4-5 AMENDMENT NO. 95

REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)

The sample analysis for determining the gross specific activity and I can exclude the radiciodines because of the low reactor coolant limit of 1 microcurie /

gram DOSE EQUIVALENT I-131, and becacse, if the limit is exceeded, the radiciodine level is to be determined every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If the gross specific activity level

.and radioiodine level in the reactor coolant were at their limits, the radiciodine contribution would be approximatt.ly 1L In a release of reactor coolant with a typical mixture of radioactiviny, the actua', radiciodine contribution would be about 20L The exclusion of radionuclide<, with half-lives less than 10 minutes from these determinations.Sas been inade for several reasons.

The first consideration is the difficulty to ider.tify short-lived radionuclides in a sample that.equires a significant time to collect, transport, and analyze.

The second consideration is the predictable delay time between the postulated release of radioactivity from the reactor coolant to its release to the environ-ment and transport to the SITE BOUNDARY, which is relatable to at least 30 minutes decay time. The choice of 10 minutes for.the half-life cutoff was made because of the nuclear characteristics of the typical reactor coolant radioactivity. The radionuclides in the typical reactor coolant have half-lives of less than 4 minutes or half-lives of greater than 14 minutes, which allows a distinct window for determination of the radionue.lides above and below a half-life of 10 minutes.

For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the SITE BOUNDARY under any accident condition.

Based upon the above considerations 4 r excluding certain radionuclides from the sample analysis, the allowable time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months between sample taking and completing the initial analysis is based upon a typical time necessary to perform the sampling, transport the sample, and perform the analysis of about 90 minutes. After 90 minutes, the poss count should be made in a reproducible geometry of sample and counter having reproducible beta or gamma self-shielding properties.

The counter should be reset to a reproducible efficiency versus It is not necessary to identify specific nuclides. The radio-energy.

chemical determination of nuclides should be based on multiple counting of the sample with typical counting basis following sampling of less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , about 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , about I day, about I week, and about 1 month.

Reducing T,yg to less than 500*F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves.

The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to BRAIDWOOD - UNITS 1 & 2 B 3/4 4-6 AMENDMENT NO. 10

REACTOR COOLANT SYSTEM f

j BASES i

SPECIFIC ACTIVITY (Continued) l

}

take corrective action.

Information obtained on iodine spiking will be used i

to assess the parameters associated with spiking phenomenon. A reduction in frequency of isotopic analyses following power changes may be permissible if j

justified by the data obtained.

i 3/4.4.9 PRESSURE / TEMPERATURE LIMITS J

j The temperature and pressure changes during heatup and cooldown are l

i limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G:

i i

1.

The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be j

limited in accordance with Figures 3.4-2a (3.4-2b) and 3.4-3a (3.4-3b)

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for the service period specified thereon:

i a.

Allowable combinations of pressure and temperature for specific 1

temperature change rates are below and to the right of the i

limit lines shown.

Limit lines for cooldown rates between those presented may be obtained by interpolation; and b.

Figures 3.4-2a (3.4-2b) and 3.4-3a (3.4-3b) define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2.

These limit lines shall be caleclated periodically using methods provided below, 3.

The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F, 4.

The pressurizer heatup and cooldown rates shall not exceed 100'F/hr and 200' F/hr respectively.

The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320'F, and 5.

System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requiressnts of ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the 1973 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel and Code.

BRAIDWOOD - UNITS 1 & 2 8 3/4 4-7 AMENDMENT NO. 30

ML20151T200

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