ML20059H315
| ML20059H315 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 10/29/1993 |
| From: | Piet P COMMONWEALTH EDISON CO. |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20059H318 | List: |
| References | |
| IEB-93-003, IEB-93-3, NUDOCS 9311100046 | |
| Download: ML20059H315 (14) | |
Text
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Camm::nwallth Edison
-1D; 1400 Opus Place r
Downers Grove, Ilhnois 60515 October 29,1993 Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation i
U. S. Nuclear Regulatory Commission f
Washington, D. C. 20555 l
l Attn:
Document Control Desk l
Subject:
LaSalle County Nuclear Power Station Units 1 and 2 i
Request for EXIGENT TECHNICAL SPECIFICATION AMENDMENT, j
to Appendix A, Technical Specification 3.6.3, Table 3.6.3-1, " Primary Containment isolation Valves" of Facility Operating Licenses NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374
References:
(a)
P. Piet letter to T. Murley, dated October 8,1993.
(b)
W. Morgan letter to T. Murley, dated October 18,1993.
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(c)
Meeting between Commonwealth Edison and the NRC Staff, dated October 21,1993.
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Dear Dr. Murley:
Pursuant to 10 CFR 50.91(a)(6), Commonwealth Edison Company (CECO)
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proposes to amend the Technical Specifications of Facility Operating Licenses NPF-l 11 and NPF-18 and requests that the Nuclear Regulatory Commission (NRC) grant an Exigent amendment to Technical Specification 3.6.3, Table 3.6.3-1, " Primary i
Containment isolation Valves." This exigent Technical Specification Amendment i
request supersedes the Reference (a) and Reference (b) submittals in their entirety.
f The amendment is needed by November 24,1993 at 11:59 PM.
The proposed Technical Specification amendment adds primary containment l
isolation valves to the list of Primary Containment isolation Valves in Table 3.6.3-1 i
of the LaSalle County Unit 1 and Unit 2 Technical Specifications. The addition of
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these valves will assure the valves are maintained and controlled as primary l
containment isolation valves. This ensures that the design and operation of LaSalle will be maintained in accordance with the existing Safety Analysis.
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The attached safety analysis shows that the Technical Specification Amendment is administrative in nature and will have minimal impact on safety. An i
exigent change is needed and could not be avoided. This exigent change is needed I
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Dr. Murley October 29,1993 to allow completion and operation of the reactor vessel level indication reference leg backfill modification by the end of the current LaSalle Unit 2 refueling outage.
This backfill modification will enhance the safety performance of the reactor vessel level instrument system in accordance with IEB 93-03. This exigent change could not be avoided due to the expedited schedule and testing of this backfill modification. The detailed design / design review process included the decision to connect the backfill instrument line inboard of the containment isolation boundary i
that eliminated the consequences of the inadvertent closure of the root valve. The design chosen by LaSalle is different than the standard industry design. This decision required that the backfill check valves be considered primary containment isolation valves and thus, be included within Technical Specification Table 3.6.3-1.
The time frame specified in IEB 93-03 and current industry information available at l
the time precluded the possibility of LaSalle's decision for choosing their design in a time frame that allowed the technical resolution to be pursued in any other manner than under exigent circumstances. The technical resolution resulting in the decision, however, exceeded a time frame that would allow normal Technical Specification processing by the NRC Staff and still allow startup from the current l
refuel outage (scheduled for November 24,1993).
Therefore, this condition was not created by the failure to make a timely application for a Technical Specification Amendment. The need for this exigent Technical Specification Amendment was previously discussed with members of the NRC Staff during the Reference (c) meeting.
In support of this request, the following information is provided:
Attachment A provides an Executive Summary outlining LaSalle's overall response to IEB 93-03.
Attachment B provides a description and safety analysis of the proposed change.
Attachment C provides the marked-up Technical Specification pages for the proposed change.
Attachment D describes CECO's evaluation performed in accordance with 10 CFR 50.92(c), which confirms that no significant hazards consideration is involved.
Attachment E provides an Environmental Assessment for the proposed change.
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. Dr. Murley October 29,1993 i
1 Attachment F provides a summary of the estimated consequences of the inadvertent closure of the Reference Leg root valve (if connected outboard) for the RVLIS piping that provides trip actuation, j
l Attachment G provicias a summary of the estimated consequences of I
the inadvertent closure of the Reference Leg root valve for the RVLIS 2
I piping that is indication only.
This request for an Exigent Technical Specification Amendment has been i
reviewed and approved bv CECO Station Managenient as well as On-Site Review l
and Off-Site Review in a.:cordance with CECO procedures.
3 i
To the best of ruy knowledge and belief, the statements contained above are true and correct. In r,ome respect the:;e statements are not based on my personal knowledge, but obtrined information furnished by other Commonwealth Edison i
i employees, contractor employees, and consultants. Such information has been reviewed in accordance with company practice, and I believe it to be reliable.
j Pursuant to 10 CFR 50.91(b)(1), a copy of this request has been forwarded to the Illinois Department of Nuclear Safety.
I county of SN truly yours, r., e
.WlgW" f
e f, 84.-
ter L. Piet Netary PobHe //y,
I Nuclear Licensing Administrator i
Attachments:
A-Executive Summary B-Description and Safety Analysis of the Proposed Change C-Marked-Up Technical Specification Pages j
D-Evaluation of Significant Hazards Consideration E-Environmental Assessment F-Consequences of Root Valve Closure
- Trip Actuation Reference Leg G-Consequences of Root Valve Closure i
- Indication Only Reference Leg cc:
J. B. Martin, Regional Administrator - Rlli 4
5 J. L. Kennedy, Project Manager - NRR D. Hills, LaSalle Senior Resident inspector - NRC R. Hague - Branch Chief - Rlli SEAL " f
" OFF!CI AL SANDRA C.LARA j
NOTAM RBUC. STATE OF ILLWO'S c f/Y C0WLSSiON E9AES {Q ?
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i ATTACHMENT A i
EXECUTIVE
SUMMARY
I LaSalle County Station is proposing to modify the plant Reactor Vessel Water Level Instrumentation (RVLIS) in response to NRC Bulletin (IEB) 93-03, " Resolution of i
issues Related to Reactor Water Level Instrumentation in BWRs." The installation of these modifications will enhance plant safety by assuring that the degassing l
phenomenon described in IEB 93-03 will not be encountered.
i The proposed modifications eliminate the phenomenon described in IEB 93-03 by providing degassified Control Rod Drive (CRD) water to the RVLIS reference leg piping. The proposed design ensures that the RVLIS reference leg piping has a i
solid, continuous column of water, free of non-condensible gases. Dissolved gases 4
in the RVLIS piping could produce uncertainties in the levelinstrumentation during i
RPV depressurization.
LaSalle's modifications consist of two different piping designs: one configuration for RVLIS piping designed to provide indication only information (industry design);
another configuration for RVLIS piping designed to provide trip actuation functions.
The different designs were chosen to physically eliminate the consequences of an inadvertent closure of the root valve associated with the RVLIS configuration that provides the trip actuation function.
l Both proposed configurations connect the non-safety-related CRD system piping to 3
each safety-related division of RPV instrumentation. The failure of the CRD piping l
may result in instrument line leakage. However, the proposed modification includes redundant check valves to isolate the safety related piping from non-safety-related piping and loss of reactor coolant in the event of a postulated failure in the non-safety-related CRD system.
i l
The proposed modifications meet the requirements of General Design Criterion (GDC) 55. GDC 55 requires lines that penetrate primary reactor containment and are part of the reactor coolant boundary have a specific valving configuration -
"unless it can be demonstrated that the design is acceptable on some other defined j
basis." The basis for the acceptability of LaSalle's proposed RVLIS configuration is described herein.
A Technical Specification Amendment is required to add the appropriate RVLIS reference leg modification check valves to the list of Primary Containment isolation Valves. These valves are added to the listing in Table 3.6.3-1, " Primary Containment Isolation Valves," and will be maintained and controlled as primary containment isolation valves.
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ATTACHMENT B Descriotion and Safety Analysis of the Pronosed Change
Background
Modifications are being made to add a backfill system for the Reactor Vessel Water Level Instrumentation (RVLIS) reference legs in response to NRC Bulletin (IEB) 93-03, " Resolution of issues Related to Reactor Water Level instrumentation in BWRs." These modifications will be completed prior to the LaSalle Unit 2 startup from its current refueling outage scheduled for November 30,1993. Similar modiiications will be installed in LaSalle Unit 1 during the next Unit 1 refuel outage (scheduled for March,1994), or during the first Cold Shutdown outage after l
December 1,1993, whichever comes first. The design evaluation for these modifications determined that NRC approval would be required for Technical Specification Amendments for both Units 1 and 2.
The modifications will connect backfill piping to the Reactor Pressure Vessel (RPV) reference leg instrument lines. The purpose of the backfill piping is to provide a continuous flow of degassified Control Rod Drive (CRD) system water to the RVLIS reference leg piping. This will ensure that the reference leg piping has a solid column of water, free of any non-condensible gases. This modification will eliminate the potential that dissolved gases in the reference leg piping could produce a notching effect in the RPV level instrumentation during RPV depressurization. Figure 1 provides a simplified diagram of the backfill modification.
Descriotion of the Backfill Line Reference Leg Modifications Six (6) RVLIS reference legs are affected by the proposed modifications. Four (4) reference legs provide input to instruments that have active automatic trip or actuation functions. The other two (2) reference legs provide input to instruments with no automatic trip or actuation functions.
The design shown on the bottom of Figure 1 is utilized on those reference legs that have automatic trip functions. This design injects water from the CRD system to the instrument reference leg lines between the containment penetration and the existing containment isolation valves.
The design shown on the top of Figure 1 is utilized on those reference legs with instruments that have no automatic trip or actuation functions. This design injects water from the CRD system to the instrument reference legs on the outboard side of the existing containment isolation valves. The design is similar to that installed at other BWRs including Dresden and Quad Cities Stations.
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ATTACHMENT B Basis of the Backfill Line Reference Leg Modifications The modification design chosen for the four (4) reference legs with automatic trip functions physically eliminates the consequences of an inadvertent RVLIS reference leg root valve manipulation error. If the backfill flow is injected on the instrument rack side (outboard) of the reference leg isolation valves, closure of the instrument root valve with the backfill system in service will pressurize the connected instruments to approximately 1300 psig. For the worst case, this would result in multiple safety system actuations up to and including the simultaneous opening of all eighteen (18) Safety Relief Valves (SRVs). A summary of the plant response to i
the closure of one of the instrument root valves is detailed in Attachment F. For LaSalle, this unique design configuration was preferred over the alternative of administrative control of the RVLIS reference leg root valves. The root valves are readily accessible to plant personnel with the exception of one root valve located in the Reactor Water Cleanup heat exchanger room (a high radiation area).
The design selected for the indication only reference legs does not eliminate the probability of the inadvertent closure of the instrument root valves. Inadvertent I
closure of the manual root valve would cause overpressurization of the affected instruments. However, the inadvertent closure of the root valve on the indication only legs does not cause a significant plant transient. The effect of closing a root valve on the indication only legs is detailed in Attachment G.
i This potential event does not initiate or defeat any automatic equipment actuations or protective functions, and is not a significant challenge to the operator's ability to j
monitor Reactor water level because:
The event will be immediately detected by the alarm.
No plant transient is induced, so evaluation of the plant status is not complicated.
l Numerous unaffected water level indications are still available in the Main Control Room, including j
2 other wide range signals (1 in each of 2 ECCS divisions) 3 narrow range signals 1 fuel zone signal SPDS (fault tolerant, auctioneered level selection) 1 upset range signal 8 backpanel wide and narrow range indications i
The backfill injection point for these two (2) racks was not redesigned as for the trip legs, because the addition of 2 more containment line penetrations, and added redesign costs including longer piping runs with seismic supports were not warranted, given the minor consequences of the potential valving error.
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ATTACHMENT B A further consideration related to the differences in injection point for the (4) trip legs versus the (2) non-trip legs is the potential for a plant worker error to be induced by the differences between the two applications. As described above, the i
design avoids unacceptable plant response to the worst case valve operation.
During normal evolutions, valving operations related to the backfill stations is done at the backfill flow control station. All six of these stations are identical, and are j
operated identically. Operation of the containment penetration root valves is an abnormal evolution, and is not normally required for more than one instrument leg at a time in order to maintain Technical Specification requirements for protection j
system operability.
l Simple check valves in series were selected for use as the containment / system
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isolation valves. These check valves have a very low opening pressure and a very i
l soft valve seat. Opening and closing of the valve, therefore, will not cause l
pressure spikes in downstream instruments that can occur with power operated i
valves and cause undesirable actuations, trips and alarms. Furthermore, the proposed RVLIS modifications will result in the non-safety-related CRD system being connected to each safety-related division of RPV instrumentation. The failure of the CRD piping integrity or a CRD pump trip may resu!t in reference leg leakage 3
from multiple divisional instrumentation line penetrations. This condition is i
mitigated by the isolation action of the backfillline check valves, which could not automatically occur for other valves such as Motor, Solenoid or Air Operated valve designs.
Comoliance to General Design Criterion (GDC) 55 For the active trip function legs, the backfill piping connects into the reference leg piping inboard of the existing containment isolation valves. These new lines are part of the reactor coolant pressure boundary. Therefore, the requirements of GDC 55 are applicable to these four (4) RVLdS reference leg backfilllines.
GDC 55 requires that each line that is part of the reactor coolant pressure boundary and penetrates primary reactor containment be provided with containment isolation valves meeting specific criteria. GDC 55 allows deviation from these specific criteria if it can be demonstrated that the containment isolation provisions for a specific class of lines are acceptable on some other defined basis.
The proposed new RVLIS backfill reference leg backfill system piping will be a class of lines accepted on some other defined basis. The valves associated with the.
backfill lines will be open during normal operation and are not considered automatic
. i isolation valves. However, the following review describes a suitable basis which i
may be used to implement the provisions of GDC 55 by demonstrating the acceptability of a particular group of lines, namely, the reference leg backfill lines.
The acceptability of the backfilllines is demonstrated as follows:
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i ATTACHMENT B i
1)
The backfill piping is not a part of the protection system.
i 2)
A 1/4 -inch orifice exists on each reference leg. The orifice is located l
inside primary containment and is unaffected by the proposed modification. The orifice limits the rear, tor coolant leakage rate to a i
value which has previously been found acceptable for the existing t
sensing lines penetrating the primary containment.
3)
The backfill piping is designed to the same quality requirements (ASME piping) of the current instrument lines. For it.1 four (4) trip function lines which inject between the containment and Excess Flow Check Valves (EFCV) the piping and supports are safety related, seismically qualified and ASME Class 2 piping.
1 4)
The backfill local flow instrument racks for all six reference legs are located as close as practical to the containment penetration with -
allowance for minimizing radiation dose to operating and maintenance personnel. The (4) piping runs of 8',13',18', and 54' achieve this objective. The longest run (54') was necessary because the associated containment penetration is inside of a High Radiation heat exchanger room (Reactor Water Cleanup System), which has normal i
dose levels of several hundred mR/ hour. Routine entry for monitoring (required shiftly) and maintenance would involve unwarranted worker i
exposure, and increased challenges to radiation area controls.
j 5)
The backfill piping is routed to minimize the potential for being i
damaged accidentally and is protected or separated to prevent failure i
of one line from inducing failure of any other line. This design t
preserved the divisional separation by building quadrants. The closest approach for two of the stations / piping is approximately 30 feet in l
distance, and 60 degrees azimuth.
6)
The same provisions are made for visualinspection of the backfill l
piping as for the original instrument lines up to an including the 1
containment isolation check valves.
7)
The backfill line connection made to the reference legs is such that the i
response time of the connected instrumentation is not affected.
8)
The backfill lines will not close accidentally during normal reactor l
operation because CRD drive water flow will keep the check valves l
open. The CRD drive water flow is checked daily on operator rounds.
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ATTACHMENT B 9)
The backfill line will be isolated by the use of simple check valves if the backfill instrument line integrity outside containment is lost during normal reactor operation or under accident conditions.
10)
The backfilllines will re-open under conditions that necessitate re-opening because the CRD water pressure is greater than reactor pressure under all circumstances. If a CRD pump is not operating, the backfill check valves will remain closed and cannot be re-opened until a CRD pump is operating with adequate water pressure and flow re-established.
11)
The backfill modifications do not provide remote position indication for the backfill isolation valves. The primary purpose for remote indication of the instrument line check valves is to inform the operator of impaired instrument capability to sense process conditions. The concern does not affect the backfill lines because no reactor processes (reactor level / pressure) are measured from these lines. Control room levelindications and/or alarms will indicate abnormally high reactor water level if a reference leg without backfill flow is leaking. A significant error due to loss of backfill flow would take several days to accumulate. Therefore, a daily surveillance of flow precludes the possibility of this event.
12)
The isolation capability of the check valves will be periodically verified by testing to leak rate criteria that is significantly more restrictive than Appendix J leak rate testing requirements.
13)
The offsite exposures from a single failure associated with the backfill lines during normal operations are below 10 CFR 100 limits.
Acceotabilitv of the Reference Leg Backfill Lines The valving provided for each line penetrating the primary containment must reflect the importance of two safety functions: 1) the function the line performs; and 2) the need to maintain containnient leaktight integrity.
Functional oerformance The backfill lines are connected in such a manner that they do not have an adverse effect on the capability of the connected instruments to perform their function. The backfill lines have no effect on the response time and an insignificant impact on instrument eccuracy. The design of the backfill system satisfies the redundancy, independence and testability requirementr -f the reactor protection system. The backfill lines are designed to the evel of knnla'Jasalleirvhs999.wpfi9
ATTACHMENT B quality as the existing instrument lines. The check valves will not close inadvertently during normal operation but will close if the backfill instrument line integrity is challenged during normal or accident conditions. The backfill line check valves will re-open under conditions that necessitate re-opening because the CRD water pressure is greater than reactor pressure under all circumstances.
Containment leaktight integrity For the indication only reference legs, the containment boundary is unaffected by the proposed backfill modifications. For the lines that provide active trip functions, the criteria fo acceptable containment leaktight integrity is: 1) maintenance of the integrity and functional performance of the secondary containment system; 2) maintenance of the rate and extent of coolant loss within makeup capability; and 3) ensuring that the calculated offsite exposures from a single failure during normal operations are substantially below 10CFR100 limits.
integritv/ functional oerformance of secondarv containment A 1/4 - inch orifice, located within the primary containment, is currently installed on each reference leg. The orifice is unaffected by these modifications and limits reactor coolant leakage to a value which has previously been found acceptable for the existing lines for line breaks inside or outside the containment. Outside of containment, isolation is provided by two simple check valves located as close as practical to the containment with allowance for minimizing radiation dose to operating and maintenance personnel. Remote position indication is not required for the reference leg backfill lines because there are no direct reactor processes (reactor pressure / level) directly measured from these lines. The same provisions are made for visual inspection of the backfill piping as for the original instrument lines up to and including the containment isolation check valves. The check valves will be leak tested by use of procedural methods which are adequate to accurately verify leakage below the chosen criteria. This leak testing method will account for measurement accuracy effects at the low flow rates needed to preserve reference leg inventory. The injection piping is routed to minimize the potential for being accidentally damaged and is protected or separated to prevent common mode or propagating failures where one failed line causes additional failures.
Maintenance of the rate and extent of coolant loss within makeuo caoability The check valves themselves provide 0: cater leak tightness than other valves used on similarly sized lincs. They have low opening pressure and a soft k:\\nla\\lasalleirvhs999.wpf\\10
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ATTACHMENT B seat which will not cause pressure spikes in the downstream instruments.
l The isolation capability of the backfill line check valves.will be periodically verified by testing to leak rate criteria that is significantly more restrictive than Appendix J leak rate testing requirements.
Offsite exoosures below 10 CFR 100 limits The proposed plant modifications for the reference leg backfill check valves do not increase the radiological con equences of any previously evaluated accident. The radiologicalimpact from a reference leg backfillinstrument line break is bounded by LaSalle's Instrument Line Break analysis (UFSAR Section 15.6.2). Therefore, the offsite exposures from a single failure associated with the backfilllines during normal operations are substantially below 10 CFR 100 iimits.
For the indication only reference legs, the backfill piping taps bto the reference _
legs outboard of the containment isolation valves. The primr. antainment barrier will still exist at the Excess Flow Check Valve (EFCV) but - a g instrument lines will now include CRD system flow. The CRD system flow oow not affect the ability of the EFCVs to fulfill a primary containment isolation function.
j Based upon the previous discussion, the proposed modifications do not adversely j
affect the function that the reference leg performs and maintains the containment leaktight integrity of LaSalle County Station. The function of the line is not l
impaired because the backfill piping is designed to the same quality as the existing instrument lines and do not have a significant impact on existing instrument accuracy. The design of the backfill system satisfies the redundancy, l
independence and testability requirements of the reactor protection system. The containment leaktight integrity is maintained because the 1/4 - inch orifice located within the existing instrument line limits reactor coolant leakage to an acceptable j
value and is unaffected by the proposed modifications. The design of the check valves provides greater leak tightness than other valves used on similarly sized lines. The isolation capability of the backfill check lines will be periodically verified to stringent requirements that ensure the integrity of the lines are maintained. The radiological impact of the proposed modification is insignificant as existing line break analyses bound the consequences of a loss of backfill line integrity.
Therefore, the proposed modifications to the reference leg backfill lines meet the requirements of GDC 55.
Reference Leg Integrity with CRD Svstem Flow The non-safety-related CRD system piping will be connected to each of the safety-related divisions of RPV instrumentation. The connection of the non-safety-related backfill piping to the safety-related vessel instrumentation line requires that an k:\\nlailasalle\\rvhs999.wpfu l
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i ATTACHMENT B l
l isolation boundary be established. The isolation boundary will ensure that the vessel reference leg piping remains filled in the event of challenges to the piping integrity or depressurization of the CRD system piping. This boundary is provided i
by two (2) safety-related check valves in series. The failure of the CRD piping integrity could result in challenges to the RPV instrumentation due to reference leg leakage. The isolation action of the two (2) backfill instrument line check valves mitigates this condition. These check valves are designed for use in an instrument i
I application and have soft seats which provide for very tight backseating and low leakage rates. The check valves allow flow to the vesselinstrumentation reference leg piping and prevent flow out of the reference leg piping.
l Leakage criteria must be established to provide assurance that vessel level instrumentation integrity is adequately maintained in the event of CRD system depressurization. The basis for the check valve leakage shall be that leakage which j
ensures that the loss of water inventory of the reference leg piping over an acceptable time period is limited to the water volume contained in the reference leg condensing pot. A loss of this inventory to the reference leg would result in indication errors that are within acceptable uncertainties for the affected j
instrumentation. This ensures that adequate vessel level indication is provided to the Operator for assessing plant operating conditions.
l A Critical Performance Leak Rate Limit of 45 cc/ hour was calculated. This ensures that the volume of water in the condensing pot is retained for at least an eight hour period. A manufacturer's standard for check valve leakage of 3.8 cc/ hour was selected as the periodic (refuelinterval) leakage test acceptance criteria. This provides significant feak rate margin for assurance that instrumentation accuracy will be maintained.
Effects of Possible Line Breaks This discussion summarizes the flow rates that would result from failures in various parts of the backfill system. To aid in this discussion, refer to Figure 1 for the location of the following break points:
Break Point Letdown Flow rate. mass flom Point A:
No Excess Flow Check Valve throttling Initial (water) clearing consisting of 4 lbm of liquid:
42 gpm = 20,700 lbm/hr (analyzed value for instrument line break)
Equilibrium (steam) blowdown flow: 1549 lbm/hr k:\\nla\\tasalle\\rvlis999.wpf\\12
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I Point B:
Excess Flow Check Valve Limited l
Initial (water) clearing: 0.5 gpm = 247 lbm/hr j
Point C:
Injection system orificing limited (assumes both Isolation Check valves fail to backseat)Y(300)] = 7.3 lbm/hr 4 lbm/hr * (Y(1000)/
j b
Point C:
1 Isolation Check valve failure, 2nd Isolation Check valve at leak l
1 rate test criteria: 3.8 cc/hr =.009 lbm/hr i
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The line break at Point A is identical to the existing LaSalle instrument line i
break analysis. In the analyzed event, the 42 gpm flow is sustained because I
an instrumentation variable leg is broken, instead of a reference leg as in this i
case. The variable leg is below the reactor water level and therefore has a j
continuous supply of liquid, as opposed to the reference leg which taps into the RPV in the steam space and clears its (minor) inventory of liquid quickly.
l The steam flow rate given is a bounding value for choked flow, assuming no line losses. Actual flow would be somewhat lower.
l l
The line break at Point B is limited by the 0.5 gpm flow through an Excess i
Flow Check Valve.
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For line breaks at Point C, the first case assumes that the letdown flow from l
2 the reactor is only limited by the orificing that establishes the injection flow l
rate. The backfillinjection flow has a driving differential pressure (dP) of l
~300 psid (1300 psig CRD pressure versus 1000 psig Reactor pressure). For l
4 the specified break, the driving dP would be 1000 psid (Reactor pressure
!l versus atmospheric). This break assumes that both Safety Related Check Valves fail to backseat, and that the reverse flow checking function of the installed flow meter does not restrict flow.
I 1
l The realistic result from a break at Point C is given as the last value, which j
assumes a single check valve failure, and the second check valve leaking at j
its test criteria limit.
i The breaks at Points A and B are applicable to the original plant design, and are not affected by the installation of the backfill system. A break at Point C would result in effects that are bounded by the original plant design, with considerable. margin.
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i ATTACHMENT B t
Current Technical Soecification Reauirements Technical Specification 3/4.6.3 includes a list of the specific components, Primary Containment isolation Valves, as Table 3.6.3-1. Table 3.6.3-1 includes Automatic isolation Valves, Manual isolation Valves, Excess Flow Check Valves, and Other Isolation Valves. Table 3.6.3-1 is used as a basis for defining Primary Containment Integrity and is referenced throughout the LaSalle County Technical Specifications.
Basis for Amending the Technical Soecifications For the active trip function legs, the backfill piping taps into the reference legs inboard of the containment isolation valves. This design categorizes these lines as part of the reactor coolant pressure boundary. Therefore, the valves associated with the active trip backfill instrument lines are considered primary containment isolation valves and need to be added to Technical Specification Table 3.6.3-1. For the indication only reference legs, the backfill piping taps into the reference legs outboard of the containment isolation valves. Therefore, the valves associated with the indication only backfill instrument lines are not considered primary containment isolation valves and do not need to be added to Technical Specification Table 3.6.3-1.
Conclusions The valving provided for each line penetrating the primary containment must reflect the importance of two safety functions: 1) the function the line performs; and 2) the need to maintain containment leaktight integrity. The proposed backfill designs comply with the requirements of GDC 55, have no adverse effect on the capability of the associated instrumentation to measure RPV water level, and j
maintain the leaktight integrity of the primary containment system boundary.
Schedule The instrument reference leg backfill modifications are being instclied at LaSalle Unit 2 during the current refuel outage (L2R05). Startup from this outage is scheduled for November 24,1993. The Unit 2 Technical Specification Amendment is required prior to startup from the refueling outage (L2R05).
The Unit 1 Technical Specification amendment is necessary to be made effective upon completion of the reference leg backfill modifications. TF m difications will be installed for LaSalle Unit 1 during the next Unit 1 refuel outage (scheduled for March,1994), or during the first Cold Shutdown outage after December 1,1993, whichever comes first.
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