ML20211N579

Amend 6 to License NPF-43,revising Tech Specs Re Primary RCS Activity Limits,Per Generic Ltr 85-19

ML20211N579
Person / Time
Site:	Fermi  (NPF-43-A-006, NPF-43-A-6)
Issue date:	02/20/1987
From:	Adensam E Office of Nuclear Reactor Regulation
To:
Shared Package
ML20211N585	List: ML20211N579 ML20211N594
References
GL-85-19, NUDOCS 8703020132
Download: ML20211N579 (12)

Text

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o UNITED STATES

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DETROIT EDISON COMPANY WOLVERINE POWER SUPPLY COOPERATIVE INCORPORATED DOCKET NO. 50-341 FERMI-?

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 6 License No. NPF-43 1.

The Nuclear Regulatory Comission (the Commission or the NRC) has found that:

A.

The application for amendment filed by the Detroit Edison Company (Deco),datedJune 27, 1986, complies with the standards and require-ments of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endanaerina the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; 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 w'ith 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 Specifica-tions as indicated in the enclosure to this license amendment; and paragraph 2.C.(2) of the Facility Operating License No. NPF-43 is hereby amended to j

read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 6, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. DECO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

0703020132 870220 DR ADOCK 0500 1

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. 3.

This amendment is effective.as of date of issuance.

FOR THE NUCLEAR REGULATORr heMISSION fft

[1I N

Elinor G. Adensam, Director BWR Project Directorate No. 3 Division of BWR Licensing

Enclosure:

Changes to the Technical Specifications Date of Issuance: February 20, 1987

ENCLOSURE TO LICENSE AMFNDMENT NO. 6 FACILITY OPERATING LICENSE NO. NPF-43 DOCKET NO. 50-341 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change. The corresponding over-leaf pages are also provided to maintain document completeness.

REMOVE INSERT 3/4 4-15 3/4 4-15(overleafpage) 3/4 A-16 3/4 4-16 3/4 4-17 3/4 4-17 3/4 4-18 3/A 4-18 (overleaf page)

B3/4 4-3 B3/4 A-3 B3/4 4-4 B3/4 4-4 (overleaf page) 6-15 6-15 (overleaf pagel 6-16 6-16 6-16a (pagination charpe only) l l

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.Q A

< g TABLE 3.4.4-1 REACTOR COOLANT SYSTEM E

i CHEMISTRY LIMITS i

s fu OPERATIONAL CONDITION CHLORIDES CONDUCTIVITY (pmhos/cm 925'C) pjj 1

1 0.2 ppm i 1.0 5.6 1 pH $ 8.6 2 and 3 1 0.1 ppm 1 2.0 S.6 5 pH $ 8.6 At all other times 1 0.5 ppm i 10.0 5.3 1 pH $ 8.6 1

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REACTOR COOLANT SYSTEM 3/4.4.5 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.5 The specific activity of the primary coolant shall be limited to:

a.

Less than or equal to 0.2 microcuries per gram DOSE EQUIVALENT I-131, l

and b.

Less than or equal to 100/E microcuries per gram.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4.

ACTION:

a.

In OPERATIONAL CONDITION 1, 2, or 3 with the specific activity of the primary coolant; 1.

Greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 but less than or equal to 4.0 microcuries per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval, or greater than 4.0 microcuries per gram DOSE EQUIVALENT I-131, be in at least HOT SHUTDOWN with the main steam line-isolation valves closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

I 2.

Greater than 100/E microcuries per gram, be in at least HOT SHUTDOWN with the main steam line isolation valves closed within l

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

In OPERATIONAL CONDITION 1, 2, 3, or 4, with the specific activity of the primary coolant greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of Item 4.a of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.

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FERMI - UNIT 2 3/4 4-16 Amendment No. 6

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

(Continued) c.

In OPERATIONAL CONDITION 1 or 2, with:

1.

THERMAL POWER changed by more than 15% of RATED THERMAL POWER in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> *, or 2.

The off gas level, at the delay pipe, increased by more than 10,000 microcuries per second in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during steady-state operation at release rates less than 75,000 microcuries per second, or 3.

The off gas level, at the delay pipe, increased by more than 15%

in one hour during steady-state operation at release rates greater than 75,000 microcuries per second, perform the sampling and analysis requirements of Item 4b of Table 4.4.5-1 until the specific activity of the primary coolant is restored to within its limit.

SURVEILLANCE REQUIREMENTS 4.4.5 The specific activity of the reactor coolant shall be demonstrated to be within the limits by performance of the sampling and analysis program of Table 4.4.5-1.

• Not applicable during the startup test program.

FERMI - UNIT 2 3/4 4-17 Amendment No. 6

TABLE 4.4.5-1 m

E PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM OPERATIONAL CONDITIONS E

TYPE OF MEASUREMENT SAMPLE AND ANALYSIS IN WHICH SAMPLE Z

AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED ro 1.

Gross Beta and Gamma Activity At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 1, 2, 3 Determination 2.

Isotopic Analysis for DOSE At least once per 31 days 1

EQUIVALENT I-131 Concentration 3.

Radiochemical for E Determination At least once per 6 months

• 1 4.

Isotopic Analysis for Iodine a) At least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 1* *,' 2 * *, 3 * *, 4 *

• whenever the specific

)

activity exceeds a limit, w

as required by ACTION b.

b) At least one sample, between 1, 2 w"

2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the change in THERMAL POWER or oft gas level, as required

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by ACTION c.

5.

Isotopic Analysis of an Off-At least once per 31 days 1

gas Sample Including Quantitative Measurements for at least Xe-133, Xe-135 and Kr-88

• 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 <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

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1. Until the specific activity of the primary coolant system is restored to within its limits l

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REACTOR COOLANT SYSTEM BASES 3/4.4.5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small fractions of the dose guidelines of 10 CFR 100.

The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters, such as site boundary location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 0.2 microcuries per gram DOSE EQUIVALENT I-131, but less than or equal to 4.0 micro-curies per gram DOSE EQUIVALENT I-131, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

In accordance i

with Generic Letter 85-19, the results of specific activity analyses in which primary coolant exceeds the limits of Specification 3.4.5 will be included in the Annual Report (Specification 6.9.1.5.d).

Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena.

A reduction in frequency of isotopic analysis following power changes may be permissible if justified by the data obtained.

l Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

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i FERMI - UNIT 2 B 3/4 4-3 Amendment No. 6

i REACTOR COOLANT SYSTEM BASES y

3/4.4.6 PRESSURE / TEMPERATURE LIMITS 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, and startup and shutdown operations.

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

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with i

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the design' assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients in the reactor vessel wall produce 1

thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.

These thermal induced compressive stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based on steady state conditions, i.e., no thermal stresses, represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

l The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes.the control-ling location.

The thermal gradients established during heatup produce tensile stresses which are already present.

The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.

Subsequently, for i

the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

(

The reactor vessel materials have been tested to determine their initial i

RT The results of these tests are shown in Table B 3/4.4.6-1.

Reactor opbtionandresultantfastneutron,Egreaterthan1MeV,irradiationwill cause an increase in the RT Therefore, an adjusted reference temperature, baseduponthefluence,phoho.rus content and copper content of the material in question, can be predicted using Bases Figure B 3/4.4.6-1 and the recommenda-tions of Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The pressure / tempera-ture limit curve, Figure 3.4.6.1-1, curves A', B' and C', includes predicted adjustments for this shift in RT for the end of life fluence.

NDT The actual shift in RT f the vessel material will be established period-NDT ically during operation by removing and evaluating, in accordance with ASTM E185-73 and 10 CFR 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the material specimens and vessel inside radius are essentially identical, the irradiated specimens can be used with confidence in predicting reactor vessel material transition temperature shift.

The operating limit curves of Figure 3.4.6.1-1 shall be adjusted, as required, on the basis of the specimen data and recommendations of Regulatory Guide 1.99, Revision 1.

FERMI - UNIT 2 8 3/4 4-4

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4 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

The program shall include the following:

7 1.

Training of personnel, 2.

Procedures for sampling and analysis, and 3.

Provisions for maintenance of sampling and analysis equipment.

d.

High Density Spent Fuel Racks A program which will assure that any unanticipated degradation of the high density spent fuel racks will be detected and will not compromise i

the integrity of the racks.

6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted.

i STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall

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be submitted following (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of l (-

fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit.

6.9.1.2 The startup report shall address each of the tests identified in Subsection 14.1.4.8 of the Final Safety Analysis Report and shall include a description of the measured values of the operating conditiorfs or characteristics 1.

obtained during the test program and a comparison of these values with design i

predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be in-cluded in this report.

6.9.1.3 Startup reports shall be submitted within (1) 90 days following comple-tion of'the startup test program, (2) 90 days following resumption or commence-ment of commercial power operation, or (3) 9 months following initial criticality, i

whichever is earliest.

If the startup report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption i

or commencement of commercial operation) supplementary reports shall be submitted at least every 3 months until all three events have been completed.

ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year.

The initial report shall be submitted prior to March 1 of the year following initial criticality.

FERMI - UNIT 2 6-15

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ADMINISTRATIVE CONTROLS ANNUAL REPORTS (Continued)

6. 9.1. 5 Reports required on an annual basis shall include:

A tabulation on an annual basis of the number of plant, utility, and a.

other personnel (including contractors) receiving exposures greater than 100 mrems/yr and their associated man-rem exposure according to work and job functions,* (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance

[ describe maintenance], waste processing, and refueling). The dose assignments to various duty functions may be estimated based on i

pocket or thermoluminescent dosimeters (TLD) dosimeters or film badge measurements.

Small exposures totalling less than 20% of the indi-vidual total dose need not be accounted for.

In the aggregate, at least 80% of the total whole-body dose received from external sources should be assigned to specific major work functions; and b.

Documentation of all challenges to main steam line safety / relief valves, and A summary of ECCS outage data including:

I c.

i 1.

ECCS outage dates and duration of outages, 2.

Cause of each ECCS outage, 3.

ECCS systems and components in the outage, and 4.

Corrective action taken.

d.

The reports shall also include the results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.5.

The following information shall be included:

(1) reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2) results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit (each result should include date and time of sampling and the radioiodine concentrations);

I (3) clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) graph of the I-131 and one other radiciodine isotope t

concentrations in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and I

(5) the time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

4

• This tabulation supplements the requirements of $20.407 of 10 CFR Part 20.

FERMI - UNIT 2 6-16 Amendment No. 6

2 ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS 6.9.1.6 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory Commissien, Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.,

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT i

6.9.1.7 Routine Annual Radiological Environmental Operating Reports ccvering the operation of the unit during the previous calendar year shall be sub.nitted prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality.

The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the< radiological environmental surveillance activities for the, report period, including a compar-ison as appropriate, with preoperational studies, with operational controls, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environraent. The reports shall also include the results of land use censuses required by Specification 3.12.2.

The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental J

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i Amen' i.ent No. 6 d

FERMI - UNIT 2 6-16a

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