ML20062B787
| ML20062B787 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 10/15/1990 |
| From: | Wallace E TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097 GL-82-33, NUDOCS 9010260136 | |
| Download: ML20062B787 (11) | |
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t TENNESSEE VALLEY AUTHORITY CH ATTANOOG A. TENNESSEE 374ot 5N 157B Lookout Place OCT 151990 U.S. Nuclear Regulatory Commission ATTN Document Control Desk Washington, D.C.
20555 Gentlement In the Matter of
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Docket Nos. 50-259 Tennessee Valley Authority
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50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - RESPONSE TO NRC SUPPLEMENTAL SAFETY EVALUATION REPORT (SER) ON REGULATORY GUIDE (RG) 1.97 COMPLIANCE DATED FEBRUARY 8, 1990 This letter provides BFN's response to NRC's Supplemental SER dated February 8, 1990.
In addition, this letter requests NRC review of new deviations and of previously submitted deviations which have not been addressed.
On December 17, 1982 NRC issued Generic Letter (GL) 82-33, Supplement 1 to NUREG-0737.
In response to GL 82-33, TVA submitted a letter dated April 30, 1984, which provided a detailed evaluation of RG 1.97 requirements and the implementation plans for BFN.
TVA's positions and commitments were revised in submittals dated May 7, 1985 November 20, 1985 and August 23, 1988 in response to the NRC letters dated January 23, 1985 and June 23, 1988.
The NRC evaluation and response to these submittals is documented in an SER and a supplemental SER which were transmitted to TVA on January 18,1989 and February 8, 1990 respectively.
TVA has reviewed the NRC's SER and the supplemental SER with regard to TVA positions and commitments. The results of this review are documented in the two enclosures. Enclosure 1 identifies previously submitted deviations to RG i
1.97 for which NRC review and approval is required and addresses discrepancies I
between the TVA letters and the NRC SER's.. Enclosure 2 identifies new deviations to RG 1.97 which NRC review and approval is required.
In addition, this letter supercedes the letter from TVA to NRC on September 27, 1990 regarding the remote position indication on the traversing in-core probe isolation valve.
Very truly yours, i
TENNESSEE VALLEY AUTHORITY l'
c-c.
E. G. Wallace, Manager 0n0O1G
"" * * " ' ' * * " ' * " 8 "" d Regulatory Affairs q
Enclosures cc See page 2 9010260136 901015 gDR ApoCK0500g'g9 j;{
An Equal Opportunity Employir
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U$S. ' Nuclear Regulatory Commission OCT 151990 cc (Enclosures):
Ms. S. C. Black, Deputy Director Protect Directorate II-4 U.S. Nuclear Regulatory Commission.
One k.'ite Flint, North 11555 hockville Pike, Rockv1113, Maryland 20852 NRC Resident-Inspector.
Browns Ferry Nuclear Plant Route 12, Box 637.
Athens, Alabama 35609-2000 i
Thierry M. Ross, Project Manager U. S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852 i
Mr. B. A. Wilson, Project Chief U.S. Nuclear Regulatory Commission i
Region II 101 Marietta Street, NW, Suite 2900 4
Atlanta, Georgia 30323 l
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ENOLOSRE 1 i
PREVIOUSLY SUBMITTED REGULATORY CUIDE (RG) 1.97.
DEVIATIONS REQUIRING' APPROVAL FOR BROWNS FERRY NUCLEARLPLANT UNITS 1. 2 AND 3 1
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It EN0LOSUREl-PREVIOUSLYSUMITTEDDEVIATIONSTO REGUIATORY GUIDE 1.97 REQUIRING APPR0VALL I.
NEUTRON FLUX.
Category 3 instrumentation has been provided for this variable in'11eu of the category 1 instrumentation specified by RG 1.97.
TVA letter dated September 14, 1990, endorsed the BWR-Owners' Group.(BWROG) appeal of the NRC staff position that directed the installation of upgraded neutron monitoring instrumentation. The NRC's~ September 18, 1990 letter to the BWROG acknowledged this appeal and allowed plant _ specific action to be deferred until the appeal is addressed.
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REACTOR' COOLANT SYSTEM-(RCS) PRESSURE.
The SERs do not address a deviation to the:RCS= pressure range or acknowledge receipt of TVA's letter from J. W. Hufham to D. Vassallo-dated November 20, 1985 which transmitted this' exception.7 That letter transmitted the following justification'.to address a revision to the RCS pressure range from 0-1500 psig to 0-1200 psig in order to-increase the accuracy of the instruments.
RG 1.97 states that the range of the RCS pressure indication will be I
15 psia to 1500 psig. Contrary to that, _the range of'the instruments provided-will be 0 to 1200 psig. The 0_to'1200 psig. range provided in-the category 1 RCS pressure indication covers the full range of j
pressures for which operator actions are' initiated during accident-conditions. The only postulated need for monitoring pressure greater than 1200 psig would.be to document peak pressure during'a reactor-4 translent. This-would, require recording capability which has been.
l provided by non-safety related-instruments since it_is'not essential'for-the operator's direct and immediate trend or transisnt:information.
Therefore, the O to 1200 psig category 1Linstruments provided to monitor j
RCS pressure are adequate for postaccident monitoring.
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EN0LOSURE1:
1 PREVIOUSLYSUBMITTEDDEVIATIONS70 j
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REGUMTORY GUIDE 1.97 REQUIRING APPROVAL III. RHR HEAT EXCHANGER 0UTLET TEMPERATURE The Supplemental SER-dated' February 8, 1990, indicates the RHR heat exchanger outlet temperature instrumentation will meet category 2.
j requirements except for environmental qualification (EQ).- TVA has i
identified this variable as category 3 based on the following
-l justification which was previously submitted by letter to the NRC dated August 23, 1988.
i The RHR heat exchanger outlet temperature instrumentation is for. the.
'd purpose of monitoring RHR heat exchanger performance..The variable can be used to determine if an individual heat exchanger is removing heat.
However, the suppression pool water temperature trend can-also be used
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to monitor.the performance of.the combination of heat exchangers being-used. The suppression pool water temperature, suppression' pool. water level, drywell atmosphere temperature and pressure, and' reactor coolant level and pressure are primary indicators to the operator'of the-heat 1
energy remaining within the primary containment. The Emergency Operating Instructions (E0I) rely'on.these variables and each of these.
j variables is included in the EQ program. Thus,La. qualified alternate set of parameters exists which is more useful to the operator.
Therefore, category 3 instrumentation is provided for ~ this variable in t
lieu of category 2 instrumentation as specified by RG 1.97..
i IV.
PRIMARY CONTAINMENT ISOLATION VALVE (PCIV) POSITION INDICATION A deviation to the category 1 position indication requirements for l
primary containment isolation valves which fail closed was previously l'
identified in TVA's letter dated May 7,i1985. This deviation is not l
addressed by either the SER or Supplemental-SER.
There are eighteen primary containment iso 1ation valves which receive l
power from the non-safety grade reactor protnetion! system (RPS) power; supply system. These valves fail closed on loss of power. ;When the RPS power supply is functional. position indication for' these = valves-will be available in the control-room.
Failure of the RPS power' supply.that
- r results in the loss of position. indication will;also cause these valves-to fail. closed..There are other parameters (temperatures,' pressure, i
radiation) specifically addressed in:the E0Is which are available to the g
operator to confirm maintenance of primary containment integrity.
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Therefore, the. position indication.for'these valveshis not required to i
ensure the containment isolation' function for.these. valves.is accomplished in the event the RPS. power supply fails. Therefore. use of the non-safety grade RPS power supply'for the position' indication 1 function for these valves meets the intent of'R0 1.97.
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.1 EN0LOSURE2 i
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r NEW REGULATORY GUIDE (RG) 1.97 DEVIATIONS' REQUIRING APPROVAL' FOR BROWNS FERRY' NUCLEAR PLANT. UNITS 1, 2 AND.3
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I EN0LOSURE2 NEWDEVIATIONSTOREGULATORYGUIDE1.97 REQUIRING'A i
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PCIV POSITION INDICATION FOR-THE RHR SHUTDOWN COOLING VALVES The RHR shu'.down cooling isolation valves (FCV-74-47, FCV-74-48) have been identified as new deviations to the RG 1.97 requirements for primary containment isolation valve (PCIV)' position indication.
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FCV-74-47 and FCV-74-48 are the outboard and inboard containment isolation valves ror the RER shutdown cooling supply line to the RHR pumps.
Except during shutdown cooling operation, FCV-74-47 is normally.
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maintained in the closed position with power removed for Appendix RE compliance. 'This. valve is not required to operate for accident mitigation. With power removed,.the position indicating.' lights do not-function. However, since this valve is administratively maintained in the closed position with power removed, position indicating lights are.
not required to ensure the' containment isolation function has been accomplished.
FCV-74-48 is normally closed and maintained closed by administrative f
controls. This valve is not required to operate.'for accident I
mitigation. The valve control circuits have been environmentally' qualified to ensure the valve will remain closed under post-accident conditions. However, the power supply cables for this. valve have not~
been environnentally qualified for post-accident conditions: since valve operation is not required.
Failure of the power supply circuits-under post-accident conditions would result in-the loss of position indication in the control room.
Since.this valve is'normally closed and remains closed in the event of a loss of power,-position'. indicating lights'are not required to ensure the containment isolation function has been accomplished. However, since the control circuits.for this valve are environmentally qualified, the position indication:will remain i
functional unless a power supply failure occurs.-
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I EN01,0SURE 2 1
NEW DEVIATIONS TO REGULATORY GUIDE 1R REQUIRING APPROVAI, NEDC-22253 concludes' that the existing TIP. system ' design is adequate for :
containment isolation provided.that three provisions are met..
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The first provision statest. "The TIP guide tubes and purge lines from i
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l their containment penetrations out to and including their isolation valves are verifled on a plant-unique basis to be safety grade."
Further review of NEDC-22253 shows that safety-grade refers)to the-structural and seismic design of this part of the TIP system. The design of this part of the TIP system at BFN meets this provision.-
The second provision states: "The logic.for the TIP' ball valves'is modified to prevent automatic reopening of the' valves upon res_et of the=
l isolation signal." The logic for the~TIP ball valves at BFN has been modified to meet this provision. The' post modification test-for acceptance of this modification.will be' performed prior to restart.'
t The third provls n states:- "The ball, valves are verified on a plant-unique basis to be normally. closed." The TIP ball valves at BFN are normally closed valves and the-logic is such that they will'open only when specified conditions are met for the operation _of the-TIP system.
q The analysis in NEDC-22253 determines that,the:non-safety,related controls and power supply for the-TIP is'olation valves:are acceptable-based on the extremely low probability of a fission < product release via the TIP guide tubes. The low probability is primarily due-to the ball valves remaining open only a small percentage of;the time. The analysis-also evaluated the consequences of a: post-LOCA release and concluded; that the resulting offsite dose would be significantly below 10 CFR'100 limits. The above discussion shows BFN.to be in compliance-with the conclusions stated in NEDC-22253.
In addition, these valves.~are-recognized in the BFN FSAR Section 5.2.3.5 as exceptions to the primary containment isolation valve criteria.
Regulatory Guide 1.97 Analysis As discussed below TVA concludes the TIP system design achieves the
-i four underlying purposes described in RG 1.97.
First, Paragraph B, Item 1. sates:
" Indications of plant variables are required by,the control room 1
operating personnel during a'cident situations to provide 3 information c
required to permit the operator to take' preplanned manual actions'to.-
accomplish safe plant shutdotm."
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NEWDEVIATIONSTOREGULATORYGUIDE1.97REQUIRINGAPPROVAL l
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Safe plant shutdown includes placing and maintaining the core in a subcritical condition and providing for long term; core cooling.- The TIP
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system is not required to operate.to accomplish safe' plant shutdown.
l Therefore, the TIP system position indication ~does not_ require qualification to RG 1.97, category 1 criteria to meet this objective.
Second, Paragraph B Item 2 statest
" Indications of plant variables are required by the control _ room
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operating personnel during accident situations'to' determine whether the reactor trip, engineered-safety-feature systems, and; manually-initiated safety systems and other systems important to safety'are I
performing their intended function."
The TIP system position indication.is not required by BFN contro1' room 3
operations ~ personnel during accident conditions..Therefore,-
qualification to RG 1.97, categoryil criteria is not' required to meet
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this objective.
Third, Paragraph B, Item 3 discusses indications to i
"... provide information.to-the operators that will enable them to determine the potential for causing a' gross breach of-the barriers to radioactive release and to determine if.a gross breach of a barrier has occurred."
TheanalysisprovidedinNEDd-22253showsthattheworsticasebreachEof 1
the TIP guide tubes would result in offsite-doses-significantly below.
c 10 CFR 100 limits and therefore would not'be. considered a gross breach.
Based on this discussion, qualification of theTIP remote position indication to RG 1.97, category 1 criteria is'not required'to meet'this objective.
Finally, the guidance provided in RG 1.97 - Paragraph B, requires data to be furnished regarding the operation of plant systems,and-the release of' radioactive materials.
This wouldl allow the operator to make~
appropriate decisions as to the.use of systems to estimate the: threat to the public, and to provide corrective actionsuns' required' The TIP system is an operationally non-safety.related system which'is assumed to.
j fai1~as a result'of any postulated accident.1 The TIP' system is not required to. operate to maintain safe plant shutdown or to mitigate the'-
T consequences of any-postulated event. The worst case. event involving a' l
breach of the TIP system. conservatively 1 assumes that;all five TIP guide L
tube lines. suffer. guillotine breaks just outside the primary ~ containment Li L
boundary. 'In.such an event, position indication'would serve no usefu11 purpose since no operator action could be -taken.l : The results of this analysis showed that the offsite dose would be less than the 10 CFR 100 limits. -Therefore, the-~ system design iseacceptable and satisfies the underlying purpose of'RG 1.97 Paragraph B~.
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NEW DEVIATIONS TO REGUIATORY GUIDE 1.97 REQUIRING APPROVAL II.
PCIV POSITION INDICATION FOR THE TIP BALL VALVES The Traversing Incore Probe (TIP) ball valves have been identified las' primary containment isolation valves with non-safety grade power and.
control circuits. Therefore, TVA considers it necessary to document the position that the TIP system design satisfies the underlying _ purpose of RG 1.97 without qualification of the position indication to category 1 requirements.
System Description
'i The TIP system is designed _to; measure and record the axial thermal neutron flux-profile at 43 radial-core locations and to calibrate the fixed incore detectors (local power range monitors). The TIP probe is-directed through the primary containment wall and into the-indexing-j mechanism by one of five TIP guide tubes. The indexing mechanism-l directs the TIP probe to.the selected'incore guide tube. The incore guide tubes extend nearly to the top of the active; core and are sealed at the upper end._ The tubes pass through nozzles-and seals beneath the reactor vessel and connect to the indexing mechanism located inside the primary containment.. Isolation valves are provided on each of the five
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TIP guide tubes inanediately outside the primary containment. Dual' barrier protection is,provided by a solenoid: operated ball valve _and an explosive actuated cable shearing valve.
This arrangement is discussed in NEDC-22253 '.'BWR:0wners' Group j
Evaluation of Containment Isolation Concerns," October 1982, and< in.the BFN Final Safety Analysis Report-(FSAR), Section 5.2.3.5.
NEDC-22253-was developed to address BWR owners group concerns related to' primary containment isolation. One of these: concerns was related to the isolation of the TIP system.
i System Analysis NEDC-22253 states the following conclusion with respect to the TIP,
3 isolation system design.
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I "This design has been previously reviewed and accepted by.the NRC on-l numerous dockets.
Isolation, provisions for the TIP purge line i
consists of a check _ valve which is adequate for containmentiisolation of.this line. The TIP guide' tubes are normally closed. When the guide tubes are open during'TIP system operation, automatic isolation-j occurs upon receipt of a'LOCA signal and retraction of the probe.
Because the_TIP system electrical: circuits-are not safety grade and l
not. separated, failure to' isolate TIP guide tubes could'be.
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postulated.
For such scenarios a probabilistic analysis shows the..
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. probability of a fission product release through the TIP guide tubes for the current isolation design to be 4.6.x 10 (-13) per reactor year. Even if such a. release was to. occur, the offsite doses for all.
i plants would be below 10 CFR 100 11mits'."
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NEWDEVIATIONSTOREGULATORYLGUIDE1.97REQUIRINGAPPROVAL j i
Conclusion In summary, the NEDC-22253 analysis was based on the complete breach of all five TIP guide tubes at.the primary containment' boundary.- This-
-breach'would render the isolation valves ineffective and the associated position indication meaningless.- Based cm this analysis, the-of f site.
L doses would remain below the 10 CFR 100 limits. :As.noted above each of.
l the four purposes of RG 1.97 are achieved without the need for j
category 1 qualification of the TIP ball-valve position; indication.
.Therefore, TVA requests n' deviation for the valve position indication on
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the'TIP ball valves. This deviation would permit TVA to qualify the.
position indication switches to category 3 criteria rather than: category -
1 as described in RG 1.97.
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