ML19254D474
| ML19254D474 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/11/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19254D472 | List: |
| References | |
| NUDOCS 7910260024 | |
| Download: ML19254D474 (6) | |
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UNITED STATES 8'
NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 53 TO FACILITY OPERATING LICENSE NO. DPR-33 AMENDMENT N). 48 TO FACILITY OPERATING LICENSE S0. DPR-52 AMENDMENT NO. 25 TO FACILITY OPERATING LICENSE NO. DPR-68 TENNESSEE VALLEY AUTH0RTTY BROWNS FERRY NUCLEAR PLANT, UNITS NOS. 1, 2 AND 3 DOCKET NOS. 50-259, 50-260 AND 50-296 1.0 Introduction 1.1 Count Rate Requirements for SRMs By letter dated July 20, 1979 (TVA BF.iP TS 126), the Tennessee Valley Authority (the licensee or TVA) requested changes to the Technical Spec-ifications (Appendix A) appended to Facility Operating Licenses Nos. DPR-33, DPR-52, and DPR-68 for the Browns Ferry Nuclear Plant, Units Hos. 1, 2 and 3.
The propoced a'mendments and revised Technical Specifications would (1) allow the count rate in the Source Rar Monitor (SRM) channels to drop r
below 3 counts per second (cps) when the entire reactor core is being removed or replaced and (2) would correct a typographical error in Section 3.10.A.4.d.
The present Technical Specifications require that a count rate of at least 3 cps be maintained whenever one or more fuel assemblies are present in the core. With respect to the second item, tbc limiting condition for operation (LCO) in Section 3.10.A.5.d reads:
"An appropriate number oT SRMs are available as defined in specification 3.10.A"; this should read "3.10.B", since the latter is the section on core monitoring which addresses '.he SRM requirements.
1.2 Respiratory Protection Program 9 011 August 25, 1977, the Conmission issued a generic letter addressed to toe licensee with respect to the respiratory protection program described in Sections 6.3.D.3, 6.3.D.4, 6.3.D.5 and Table 6.3. A of the Technical Specifications for eech of the three Browns Ferry units.
The letter called attention to the fact that on November 29, 1976, the Commission published in the FEDEPAL REGISTER an amended Section 20.103 of 10 CFR 20, which became effective on December 29, 1976. One effect of this revision is that in order to receive credit for limiting the inhalation of airborne radioactive material, respiratory protective equipment must be used as stipulated in Regulatory Guide 8.15.
Another requirement of the amended regulation is that licensees authorized to make allowance for use of respiratocy protective equipment prior to December 29, 1976, must bring the use of their respiratory protective equipment ini.o conformance with Regulatory Guide 8.15 by December 29, 1977.
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. The Browns Ferry Technical Specifications anticir.ated the above Amendment to Secticn 20.103 of 10 CFR 20; section 6.3.D.5 states:
"5.
These specifications with respect to the provision of 20.103 shall be superseded by adoption of proposed changes to 10 CFR 20, section 20.103, which would make this specification unneces sa ry. "
Ir, our letter of August 25, 1977, we advised TVA that "In view of the provisions of Section ' 3.D of your Technical Specifications, which require conformance witn 10 CFR 20, the fact that Section 20.103 no longer requires specific authorization to employ respiratory protective equipment, and the revocation provisions of subrection 6.3.D.5, we con-clude that the necessary amendment to your facility's Technical Spec-ifications can be effected by merely deleting Sections 6.3.D.3, 6.3.D.4, 6.3.D.5 and Table 6.3.A."
In the letter, we also advised TVA that " Based on the revocation pro-vision of your current specification on respiratory protection and in the absence of prior written objection from you, we will include deletion of this specification in an amendment of your Technical Specificnions approved after December 28, 1977.
No response to this letter is required".
This amendment will delete Sections 6.3.D.4 6.3.D.5 and Table 6.3.A in accordance with our letter of August 25, 1977.
There is no safety significance since these sections are in effect revoked by 10 CFR 20.103.
2.0 Discussion During any core alteration, and especially during core loading, it is necessary to monitor flux levels.
In this manner, even in the highl unlikely event of multiple errors, there is reasonable assurance that any approach to criticality would be detected in time to halt operations.
The minimum count rate requirement in the Technical Specifications accom-plishes three safety functions:
(1) it assures the presence of some neutron in the core, (2) it provides assurance that the analog portion of the SRM channels is operable, and (3) it provides assurance that the SRM detectors are close enough to the array of fuel assemblies to monitor core flux levels.
Unloading and reloading of the cntire core leads to some difficulty with this nir.. mum count rate requirement.
When only a small number of assem-blies are present within the core, the SRM count rate will drop below the minimum due to the small number of neutrons being produced, and due to attenuation of these neutrons in the water (and control blades) separating the fuel from the SRM detectors.
Past practice has been to connect temporary "dur. King" chambers to the SRM channels in place of the normal detectors, and to locate these detectors near the fuel.
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. Besides being operationally inconvenient, dunking chombers suffer from signal variations due to their lack of fixed geometry. Moreover, the use of dunking chambers incrases the risk of loose objects being dropped into the vessel.
3.0 Evaluation 3.1 Minimum Flux in the Core A multiplying nedium with no neutrons present forms the basis for an accident scena-io in which reactivity is gradually but inadvertently added until the medium is highly supercritical.
No neutron flux will be evident since there are no neutrons present to be multiplied. The introduction of some neutrons at this point would cause the core to undergo a sudden power b_ st, rather than a gradual startup, with no warning frcm the nuclear instrumentation.
This scenario !s of great concern when loadins ;esh fuel, but is of lesser concern for exposed fuel.
Exposed fuel continuously produces neutrons by spoataneous fission of certain plutonium isotopes, photofission, and some delayed neutron emission.
This neutron production in exposed fuel is nurmally great enough to meet the 3 cps minimum for a full core after a refueling outage with the lumped neutron sources removed.
Thus, there is assurance that a minimum flux level will be present as long as some exposed fuel is present.
We therefore find the proposed amer.dment to be acceptable from the point of view of minimum flux provided the words
" full core reload" in Specification 3.10.B.l.b.2 are interpreted to mean
" reload of fuel which has previously accumulated exposure in the reactor."
We do not find the amendment to be applicable to the loading of a new core containing only fresh fuei.
Such a loading must use lumped neutron sources and dunking chambers to meet the normal 3 cps minimum count rate.
On September 19, 1979, TVA proposed alternative wording to Section 3.10.B.2 of the Technical Specifications specifying that the minimum count rate requirement of <3 cps only applies when bath irradiated and fresh fuel is being loaded; this change satisfies the staff's concern above and is acceptable.
3.2 SRM Operability The amended Specifications 3.10.B. l.b.2 and 4.10.8 will requi a a functional check of the SRM channels by eneans of a neutron source prior.o beginning core alterations and at least every eight hours thereafter.
The required interval for other types of alterations is usually once per day.
This would be sufficient for core unloading and reloading except that the more exten-sive fuel hcndling operations involved load to a slightly greater possibility of SRM failure. We agree that a tripled test frequency is sufficient to cover this, and therefore find the eight hour interval to be acceptable.
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- 3.3 Flux Attenuation The four SRM detectors are located, one per quadrant, roughly half a core radius from the center. Although these are incore detectors and thus very sensitive when the reactor is fully loaded, they lose some of their effectiveness when the reactor is partially defueled and the detecto s are located some distance from the array of rem.ning fuel.
GE's spent fuel pool studies have shown(I) that 16 or more fuel assemblies (i.e., four or more control cells) mast be loaded together before criti-cality is possible.
In spiral (and most other) loading sequences in the Browns Ferry cores, an array containing four or more control cells will be at most two control cells (i.e., about two feet) away from an SRM detector. We have previgugly examined the sensitivity loss in such a case on another docket, (21 and found it to be at most one decade of sensitivity (i.e., about one fifth of the SRM's logarithmic scale).
We find this to be acceptable.
However, there are areas near the 90 and 180 sides of the Browr.s Ferry cores where it is possible to load fuel at a still greater distance from the nearest detector.
In the absence of quanti'.ative justification by the licensee, we cannot find this amendment accentable for all possible loading sequences. Therefore, we find this ameniment to be acceptable only for spiral unloading / reloading sequences, which we understand to be the only sequences the licensee plans to actually use.
By " spiral sequences," we mean any sequence in which the central control cell is lar unloaded and first reloaded, all fueled locations are contiguous, and no imbedded cavities or major peripheral concavities are permitted.
On September 19, 1979, TVA proposed alternative working to Section 3.10.B.2 which specifies that the less than 3 cps only applies when the core is loaded in a spiral sequence; this satisfies the staff's concern and is acceptable.
4.0 Environmental Considerations We have determined that these amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made thi:
determination, we have furthrr concluded that these amendments involve an action which is insignificant from the standpoint of environmental impact, and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statcment, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
5.0 Conclusion We have concluded that:
(1) because the amendments do not i volve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do rot involve a significant hazards con-sideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, 1"'9 014
- and (^-) such activities will be conducted in compliance with the Commission's regulatio6s and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.
Dated:
October 11, 1979
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REFERENCES 1.
General Electric Standard Safety Analysis Report, 251-GESSAR, Section 4.3.2.7, p. 4.3-27.
2.
" Safety Evaluation by the Office of Nuclea, Reactor Regulation Supporting Amendment No. 27 to Facility Operating License No. DPR-63," Docket No.
50-220, enclosed with letter, Tnomas A. Ippolito (NRC) to Donald P. Dise (Niagara Mohawk Power Corporation), dated March 2,1979.
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