ML20058A315
| ML20058A315 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/16/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20058A306 | List: |
| References | |
| NUDOCS 7812050023 | |
| Download: ML20058A315 (9) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTIf;G AMENDMENT NO. 44 TO FACILITY OPERATING LICENSE NO. DPR-33
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AMENDMENT NO. 40 TO FACILITY OPERATIt'G LICENSE NO. DPR-52 AMENDMENT NO.17 TO FACILITY OPERATING LICEllSE NO. DPR-68 i
i TENNESSEE VALLEY AUTHORITY 1
f BROWNS FERRY NUCLEAR PLANT, UNITS NOS. 1, 2 AND 3 DOCKET NOS. 50-259, 50-260, AND 50-296 1.0 Introduction By letter dated August 11,1978 (TVA BFNP TS 114), the Tennessee Valley Authority (the licensee or TVA) requested changes to the Technical Specifications (Appendix A) appended to Facility Operating Licenses Nos. DPR-33, DPR-52 and DPR-68 for the Browns Ferry Nuclear Plant, Units Nos.1, 2 and 3.
The proposed amendments and revised j
Technical Specifications would (1) permit the averaSe power range monitor (APRM) system to be inoperable in the refuel mode, provided the source range monitors (SRMs) are connected to give a non-coincidence, 4
J high flux scram and (2) in the refuel and shutdown modes only, permit i
less than three intermediate range monitors (IRMs) per trip channel to be operable-provided at least four IRMs (one in each core quadrant) 4 are connected to give a non-coincidence, high flux scram. The present Technical Specifications require that a minimum of three IRMs per trip channel be operable at all times (i.e., shutdown as well as startup and operation).
l The reason for this request is to allow the interchange of the fission chambers in the current APRM system with reduced radiation exposure to the operating personnel and with reduced handling and movement of fuel.
This can be achieved by removing many LPRMs simnultaneously 'rather than i
in sequence. The sequential removal would leave the APRM system operable but the simultaneous removal would not.
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1 In a separate letter dated August 2,1978 (TVA BFNP TS 112), TVA requested five changes to the Technical Specifications, all of which are administrative in nature.
The changes would:
(1) clarify an ambiguious portion of the Technical Specifications related to the rod l
block monitor system, (2) remove reference to an obsolete 1968 version of an ASTM procedure, (3) modify the list of snubbers that are required to be operable, (4) change one of the four locations from which milk samples are routinely collected and (5) remove a specification for additional test of secondary containment that only applied to the first operating cycle for each Browns Ferry unit.
2.0 Discussion As described in Section 7.5 of the Final Safety Analysis Report (FSAR) for the Browns Ferry Nuclear Plant (BFMP), the Neutron Monitoring System consists of six major subsystems:
(a) the Source Range Monitor (SRM) subsystem, (b) the Intermediate Range Monitor (IRM) subsystem, (c) the Local Power Range Monitor (LPRM) subsystem, (d) the Average Power Range Monitor (APR!!) subsystem, (e) the Rod Block Monitor (RBM) subsystem and (f) the Traversing In-Core Probe (TIP) subsystem.
The IRM subsystem conitors neutron flux fror the upper portion of the SRM range to the lower portion of the Power Range Monitoring Subsystems.
The IRM system normally consists of eight noveable miniature chambers with two such chanbers in each core quadrant. No more than one of the IRMs in each quadrant may be bypassed. The eight IRM channels are divided into two IR!! sub-systems and at least one IRM from each sub-system must reach 120/125 of full scale to initiate a reactor scram. The IRM system is nominally designed for protection in the startup mode and analyses (FSAR, Section 14.5.3) have been perfortred showing that the system adequately prevents fuel damage due to rod withdrawal errors postulated to occur during startup.
The APRM subsystem provides a continuous indication of average reactor power frcn a few percent to 1255 of rated reactor power. The subsystem has six APEM channels, each of which uses input signals from a number of LPRM channels. Three APRM channels are associated with each' of the trip systems of the Reactor Protection System.
T'he APRM system which consists of a number of stationary fission chambers -
i dispersed thmughout the core, is normally required _ to be operable in j
the refuel mode with a high flux scram setpoint corresponding to 15%
rated power.
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e 3-D cauu the / F:'1 response is actually the ccm bined response of a number of it.divice.1 fission chu bers located throughout the core, the APRM prirv rily p.v. ' des protect ic.n for core-wide trant ice.t power increases which might occur in the run t. ode (above 15' rated power).
Also, in thc startup rir! the AP'L". Travii. b.cl.up prctection to the IR" system acains c loc.:li;:: d powcr increases which night reselt from postulated rod.:ithdrve i e r ro rs.
Alitet.nh the RM syster as described above it. required by the current leci nical S p'ci ficattor'. to be operable in both the shutacen and refuel rx ;e, no s p. ci fic event b..s been an;lyzed in the plant FSAR uhich tci:( s cre-Ut for a
- .a a " initia t t d by tiic 11:". systen, with a given set point er nu!.: cr of by; med ins ti c. n t r.. Sinilarly, the APh", is requii t J to c.p.. '. t e n o n i l l, in tlc refuel i.ade, but no ti Lnsient or accidcnt takir.g c. J '. t f o r.w /J i." i r,i t i.t + d s crn'.. L t d po s t ul a t c d t o oc c ur i n the iis'l We*6 bee r ani.l y:.:t d i t. the Flout FSAR. As oiscost.ed in the
.tiori ehich tc'. iou.. there is only om event whi h the staf f rcn c....t pos t ulate - nr..el.s, an open. tor Lyp.assing the interlocks and withdrawing
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,usuit di re::tiy ir, posi: i u react.vi t y ins t etions, inciocing cent roi r od iecoval crrar during ref acling and fuel assertiy insertion error during refueline!.
Section 7.6 of tM FSAP. c'escribes the refuelina interlocks th:t prevent an inadvert -nt criticality during refueling oper atior.s and t ha + are der.i r:ntd te back up procedural core recctivitt con t r ol: du it g refuelin; opm t ;cus. Sc ction 3.10 of t he P c.sent. feriy Gt clece Plant Technical Specifice ians lists the rcstricticris that apply durinn cere alterations to ensure th::t core reactivity is within the capability of the control rods ar.d to prevent criticality during refecling.
! '" !": r.c.Je switch is in P.IfUD. Onl/ o r. r' Control rod Can be with$.*
5".;ti:n e a seccnd rc.! initictc c red bicck thc reby preveatin; i; a t W a.:.! ef icra t ha n on e ra f a '. a t i m.-
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- in W rtent criticality. ii'e r.uclL.r ch.2 ruc terit t ics o f the c arc a.,se that the re..ctor is su' c ritical cs en c' ;n tt.e hichMt torth centrol r.
ir. fally. i:'uren.
Refueling preced.n s ai c written to avoid situath in t.hich ie.dwrient cri'ic ili t is r.0 i ble. The cor bir.o tien of reiucifi-i, tulccl.s for c en t: 01 re :i' e.' the re f uc li re pl a t f.,r.r. pre..
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i During certain periods, it is desirable to perform maintenance on two control rods and/or control rod drives at the same time. The maintenance is performed with the mode switch in the " refuel" position to provide the refueling interlocks no.mally available during refueling operations.
In order to withdraw a second control rod after withdrawal of the first rod, it is necessary to bypass the refueling interlock on the first control rod which prevents more than one -control rod from being withdrawn at the same time. The present Technical Specifications permit bypassing the refueling interlock with the requirement that an adequate shutdown margin be demonstrated or that all remaining control rods have their directional control valves electrically disarmed to ensure that inadvertent criticality cannot occur during this maintenance, The adequacy of the shutdown margin is verified by demonstrating that i
the core is shut down by a margin of 0.38 percent Ak with the strongest operable control red fully withdrawn, or that at least 0.38% Ak shutdown margin is available if the remaining control rods have had their direc-tional control valves disarmed.
Disarming the directional control valves does not inhibit control rod scram capability.
3.0 Evaluation 3.1 APRM-IRM Systems We have reviewed the plant Technical Specifications and the nuclear design characteristics of the fuel. We have concluded that a local criticality during shutdown or refueling operations could only occur through violation of tecnnical specifications such as an operator error in withdrawing a control rod for maintenance, adjacent to a previously withdrawn rod.
Although such operator errors are not likely to occur, they are not impossible. We have therefore considered the' applicant's request for proposed modifications to the SRM, IRM and APRM systems in terms of the impact on the protection against postulated local criticality which could occur while the mode selection switch is in the refuel or shutdown positions.
1 The most severe test of the adequacy of the modified IRM and SRM systems would be the withdrawal (for maintenance) of t control rod near the edge of the reactor core face adjacent to a previously j
withdrawn rod. Because the proposed Technical Specifications allow one IRM in each core quadrant to be bypassed, the IRM nearest the pair of withdrawn rods was assumed to be bypassed, i
Because the modified IRM system would initiate a reactor scram when any IRM reaches the trip set point, the modified system will actuate a scram at an earlier time during the wi*hdrawal of the seco'nd eod than would the normal system. The normal sys,em would require trips in each IRM subsystem.
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We conclude that the redundant independent IRM instruments connected to give non-coincident scrams provide better protection against fuel damage due to a localized power increase than.does the ApRM system with its 15% scram setpoint. Beacuse the IRM instruments are independent in the modified IPM system, the IRM will be its own backup. The IRM scram setpoint will be 120/125 of the lowest IRM scale which correspords to very los flux levels. Although the flux level at the second nearest IRM (the backup IRM) would be low throughout the rod withdrawal event, it will be high enough to scram the reactor at a lo,eer flux level than with the present arrangenent using the APRM monitors. We therefore, conclude that the litensee's proposal for the IRM system modification results in a system that is more sensitive to possible operator errors during core modifications than is the present arrangement and therefore the proposed modification is acceptable.
In addition, the SRM system would be connected to scram the reactor at a level of 5 x 105 counts per second. Although the SRM is not considered safety grade equipment, the licensee has proposed to provide the SRM 1
scram function, and we believe this is desirable as an additional backup.
to the IRM system.
A concern which was raised during the NRC review was what technique (s) will be provided to assure that the reconfiguration of tSe SRM's and IRM's I
to the non-coincidence trip rode is in fact accerplishen prior to removing the APRM protection. By letter dated November 13,1978, the licensee has agreed to the followino administrative controls. The pracedures related to naintenance cf detectors (" Browns Ferry Nuclear Plant-Instrument Maintenance Instructions") will be reviewed, and revised as necessary, to include:
(1) a specific reference to the Technical Specification Table 3.1. A and associated Notes 21 and 22, which indicate i
that the SRM's/lRM's must be re-configured to provide non-coincidence high flux scram protection, and (2) a specific procedural step _which requires that verification will be made that the appropriate shorting links have been renoved prior to maintenance on IRM/LPRM detectors.
These controls provide adequate assurance that the reconfiguration of the SRMs ard IRMs will be accomplished prior to removing the APRM protection.
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4 Due to the interwoven design of the shorting link system, clarification of the notes to Table 3.1. A is needed.
The following sentence should be added to Note 21: "The removal of eight (8) shorting links is required to provide non-coincidence high-flux scram protection from the Source Range -
Moni to rs".
The following sentence should be added to Note 22: The removal of four (4) shorting links is required to provide non-c incidence high-flux scram protection from the IRM's".
As is proposed by the licensee for Unit No. 3, the Technical Speci fications for Units Nos. I and 2 sho scram setpoint is < 5 x 10gid include in Note 21 to Table 3.1.A that the CPS.
To summarize, we find that the modification TVA has proposed for the Brcwns Ferry IRM systems is acceptable.
The modified systems will be more sensitive to the flux perturbations resulting from the worst postulated transient than the present arrangement.
Furthermore, as discussed previously, the redundant and independent 1RM instruments which will comprise the modified IRM systems -will provide protection against inadvertent criticality in the refuel mode equivalent to or_ better than the present APRM system.
Inoperability of the APRM with the modified IRM in place is therefore acceptable for the refuel mode.
t As described in the " Discussion" above, Section 3.10 of the Technical Specifications includes restrictions on withdrawal of control rods during core alterations. As an additional backup to the neutron monitoring instrumentation, we have propostd, and the licensee has accepted, an addition to the surveillance requirements in Section-4.10 of the Technical Specifications to require that-no control rod may be-withdrawn for maintenance until two licensed operators have confirmed that there is no fuel in the cell controlled by the particular control rod or that all immediately adjacent control rods are fully inserted and electrically disarmed.
This requirement, in conjunction with the more sensitive IRM system, will insure that there is no possibility 3
1 of inadvertent criticality during core modifications.
In summary we conclude that the proposed changes to the licensee's
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Technical Specifications do not involve an increase in the probability i
of a transient or accident but in fact should reduce the consequences of such events.
The proposed changes do not involve a reduction in" safety margin. No change in a safety limit or a safety limit margin is involved.
We therefore conclude that the proposed changes to the Browns Ferry Technical Specifications with respect to the APRM and IRM systems are acceptable and do not involve a significant hazards censideration.
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3.2 Snubbers Table 3.6.H of the Browns Ferry Technical Specifications contains a list of " Shock Suppressors (snubbers)" that are required ;o be operable to protect the primary coolant system or other safety related components.
Section 3.6.H.6 of the Technical Specifications states that:
Snubbers may be added to safety-related systems without prior license amendment to Table 3.6.H provided that a revision to Table 3.6.H is included with j
a subsequent license amendment request". TVA proposes to add three snubbers to Table 3.6.H on the Fire Protection System. They also propose tu delete the two snubbers that were formerly on the control rod drive (CRD) line since the CRD return line has been capped at the reactor vessel and rerouted to the reactor water cleanup return line as part of the modifications to reduce the potential for cracking in the CRD return line.
The line-and thus the snubbers-are no longer present in the system.
I TVA also proposes to delete four snubbers from Table 3.5.H on the condensatt bypass line, since this line is a non-critical system (i.e., not classificd as a safety-related systen) and failure of this by-pass line will not cause damage to a critical system. We conclude that the proposed changes to Table 3.6.H are acceptable.
3.3 ASTM Procedure Section 4.9. A.3 of the Technical Specifications requires that aL sample of diesel fuel shall be analyzed once a month and that the quality shall be within the acceptable limits specified in an obsolete 1968 i
version of ASTM procedure D975.
This ASTM procedure is under revision.
I In lieu of referring to the specific version of the ASTM procedure (which is subject to the periodic revisions) TVA has proposed to change the Technical Specifications to read:
"The quality shall be within the acceptable limits specified in Table 1 of the latest revision to ASTM 0975 and logged".
Since the most recent revision to this standard method of analysis reflects the current best judgement of the country's experts who are on the various ASTM committees, the most recent edition of the standard is the one that should be used as the " referee method" rather than the edition in effect when the plant was under construction. We conclude that the proposed change to the Technical Specification is acceptable.
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3.4 Rod Block Monitors Control rod block functions are provided to prevent excessive control rod withdrawai so that the safety limit minimum critical power ratio is not violated. Two rod block monitor (RBM) channels are provided. The current Technical Speci fications and the Bases therefore (Section 3.2.C.2) j state that:
"The minimum number of operable instrument channels r
specified in Table 3.2.C for the Rod Block Monitor may be reduced by.
one in one of the trip systems for maintenance and/or testing, provided that this condition does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any thirty day period".
TVA proposes to relocate this requirement in the Technical Specifications, adding it as part of " Note 7" to Table 3.2.C and rewording it to be more specific. The revised wording will be:
"Two RBM channels are provided and only noe of these may be out of service for testing and/or maintenance provided this condition does not last longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any thirty day period". This is not a change to the requirements in the.
Technical Specifications but simply a change in wording of the requirement and its location in the Technical Specifications.
We conclude that the proposed action is an improvement in phraseology and is acceptable.
3.5 Secondary Containment Testing Section 4.7.C.b of the Technical Specifications required additional tests of secondary containment during the first operating cycle of each of the three Crowns Ferry units to supplecent the other specified tests which are conducted throughout the life of the plants. All three Browns ferry units have completed their first operating cycle and the additional tests specified in Section 4.7.C.b.
TVA, therefore, t
proposes to delete this requirement, since it is no longer applicable.
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We conclude that the proposed deletion is acceptable.
3.6 Milk Sample Locations As part of the environmental radiological monitoring program at the Browns Ferry Nuclear Plant, TVA collects and analyzes a number of samples.
The Browns Ferry Nuclear Plant' Environmental Technical Specifications state that " milk shall be collected....from at least four farms in the vicinity of the plant..." and that"...any location from which milk can no longer be obtained may be droprad from the surveillance program. The r
NRC shall be notified in writing that milk-producing animals are no
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longer present at that location. An additional milk sampling location will then be added to the program..." (Section 4.2.3.b).
As of May 15, 1978, milk is no longer available from the dairy farm located approximately four miles north of Browns Ferry Nuclear Plant.
The milk producing animals have been sold and removed from the farm. A dairy farm located approximately five miles north of the plant has-been ad::c: to tha : :ter' ; n egrar..
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_9-We have reviewed the meteorological data and deposition factors for the Browns Ferry plant and conclude that the new sample location is
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accepta bl e.
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 this determination, we have further concluded that these amendments -involve an action which is insignificant from the standpoint of environmental s
impact, and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
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5.0 Conclusion r
l We have concluded that:
(1) because the amendments do not involve a l
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 not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations 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.
I Dated: November 16, 1978 J
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