ML20133C876
| ML20133C876 | |
| Person / Time | |
|---|---|
| Issue date: | 09/26/1985 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Bernero R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8510070527 | |
| Download: ML20133C876 (20) | |
Text
k SEP 2 61995 MEMORANDUM FOR:
Robert M. Bernero, Director Division of Systems Integration FROM:
Harold R. Denton, Director Office of Nuclear Reactor Regulation l
SUBJECT:
SCHEDULE FOR RESOLVING AND COMPLETING GENERIC ISSUE NO. 87 - FAILURE OF HPCI STEAM LINE WITHOUT ISOLATION The technical resolution for Generic Issue No. 87, " Failure of HPCI Steam Line Without Isolation," is assigned a "HIGH" priority ranking based on the evaluation prcvided in Enclosure 1.
This memorandum directs you to take the actions necessary to resolve this issue.
This issue is being addressed in near term operating license cases. The resolution of th?se cases depends on the specific plant design features.
For example, in the Shoreham plant the HPCI steam line containment isolation valve will be closed during normal operation, while in the Limerick plant this valve is qualified to close under steam line break conditions.
Therefore, the resolution of this issue should be coordinated with EQB and RRAB, who are resolving this issue on operating license applications.
In accordance with NRR Office Letter No. 40, " Management of Proposed Generic Issues," the resolution of this issue will be monitored by the Generic Issue Management Control System (GIMCS). The information needed for this system is indicated on the enclosed GIMCS information sheet. Your schedule for resolving and completing this generic issue should be commensurate with the priority nature of the work and consistent with the NRR Operating Plan.
Normally, as stated in the Office letter, the information needed should be provided within six weeks.
The enclosed prioritization evaluation will be incorporated into NUREG-0933, i
"A Prioritization of Generic Safety Issues," and is being sent to the regions /u)[,
c and other offices, the ACRS, and the PDR for comments on the technical b
accuracy and completeness of the prioritization evaluation. Any changes
[ L< d jgl as a result of consnents will be coordinated with you. However, the schedule Il i
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SEP 2 6115 for the resolution of this issue should not be delayed to wait for these coments.
The information requested should be sent to the Safety Program Evaluation Branch, DST. Should you have any questions pertaining to the contents of this memorandum, please contact Louis Riani (24563).
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, E Leats Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
1.
Prioritization Evaluation 2.
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l ISSUE 87:
FAILURE OF HPCI STEAM LINE WITHOUT ISOLATION l
DESCRIPTION Historical Background This issue concerns a postulated break in the High Pressure Coolant Injection (HPCI) steam supply line and the uncertainty regarding the operability of the HPCI steam supply line isolation valves under those conditions (Ref. A). A similar situation can occur in the Reactor Water Cleanup (RWCU) system.
The HPCI steam supply line has two containment isolation valves in series, one inside and one on the outside of the containment.
Both are normally open in most plants. Two plants operate with the HPCI outboard isolation valve normally closed. A HPCI supply valve, located adjacent to the turbine, and the turbine stop valve are normally closed.
The RWCU also has two normally open containment isolation valres, which must remain open if the system is to function.
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The operation of the valves is tested periodically without steam flow. Also, due to flow limitations at the valve manufacturers facilities only the opening characteristics are tested under operating conditions.
Therefore, the capability of the valves to close wtan exposed to the forces created by the flow resulting from a break downstream has not been demonstrated.
i However, there are reasons why the valves may operate under these accident conditions. The containment isoIation valves are specified to open or close l
within 15 to 20 seconds.
Calculations performed by Bechtel (Ref. F) indicate that the mass flow through the Hl:CI steam line isolation valves reduces from 1470 lbm/sec. at the time of the break to 328 lbm/sec. after 0.135 sec.
and remains constant until the valve closes.
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-2 The valve type is not under the GE scope of control but is sekcted by the plant AE (Ref. B).
This results in a diversity of valves and valve types from plant to plant and increases the difficulty of demonstrating valve operating capability. Some plants have "Y" type globe valves and others have gate valves.
One plant using globe valves for HPCI steam supply isolation has the valve inside containment positioned such that the steam flowing through the valve exerts a force on the valve skirt in the close direction.
This force is expected to reduce the closing torque requirement of the valve motor operator and increase the probability that the valve will close when a large amount of steam is flowing through the valve. Also, some valve experts believe that the force required to open gate valves under pressure is greater than the force required to close the valve under flow.
Safety Significance In Mark I containments the HPCI steam line exits the drywell and enters the torus compartment where it typically traverses approximately a 75* arc before exiting to the HPCI pump room.
In the four corners of the reactor building along side the torus compartment are four triangular shaped rooms which house the RHR/LPCI system, the RCIC system and the core spray systems.
In some reactor buildings there is a ventilation opening or a do.or, usually open, between the rooms housing the emergency core cooling pumps and the torus compartment. Given an unisolated break of the HPCI steam supply line in the torus compartment, the environment in the emergency core cooling pump rooms may exceed design limits. This places in jeopardy the other systems required to cool the core.
In Mark II containments the emergency core cooling components are typically housed in individual rooms which are contained in the large annular shaped area about the suppression pool. The HPCI steam supply line exits the containment and is then routed down through two floors to the room 'containing the HPCI turbine and pump. Again, given an unisolated break of the HPCI steam
)
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. supply line, other systems which may be required to cool the core may be placed injeopardy.
Possible Solutions A proposed solution to the HPCI problem is to require that the outboard HPCI isolation valve be normally closed.
However, a small bypass line, on those plants not having this feature, would be required to prevent thermal shock and water hammer, and to provide assurance that leaks in the line would be detected before they become breaks.
If the HPCI supply valve were kept normally open--currently it is kept normally closed--the probability of not getting steam to the HPCI turbine when needed might not be significantly changed.
Another solution that would apply to valves in any system might be a demonstration by test or the verification of use in other service applications that certifies the operability of the valve under line rupture flow conditions.
If the normal HPCI steam flow rate approximates that estimated for a break in the steam line, the valves might be tested by individually closing them when the HPCI turbine is in operation.
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PRIORITY DETERMINATION Frequency Estimate In the Browns Ferry IREP study (Ref. D) the frequency for intermediate size steam line breaks, in which size category the HPCI steam supply line is included,
~4 is stated as being 2 x 10 per reactor year.
It is also assumed that a break is equally probable at any point in the steam lines of this size category.
The HPCI steam supply lines were estimated to constitute 23% of the steam lines in the intermediate size category. Hence, the frequency of a HPCI steam supply
-5 l
line break will be assumed to be 5 x 10 per reactor year.
. The probability that both steam supply line isolation valves will fail to close is difficult to determine on a probabilistic basis.
First, we are not dealing with random failures but rather with a lack of engineering knowledge.
- Second, if one valve fails to perform its intended function because of conditions which exceed its design capability, it would be most probable that the second valve would also fail to function. As an upper bound calculation, we can assume that the valve failure rate will be unity given a line break, and that the dependency between valves is also unity. The lower bound can be calculated by assuming that the valve design is adequate and that there is no failure dependencies between the valves, so that the frequency of both valves failing is the product of both valves independent failure frequency.
The major contribution to the accident scenario being considered is the depen-dency between the unisolated steam line breaks and the low pressure injection systems.
For both the upper and lower bound calculations, it will be assumed thatthedependenceisunity,thatis,thatthelowpressureinjectionsystems will fail given an unisolated line break.
If, during the accident condition described, the core is maintained covered by the feedwater system, the steam mass flow generated by decay heat should lower
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to a point that would permit the closing of an isclation valve. One means available would be electrically closing the isolation valve inside the contain-ment; the other means available is manually closing one of the isolation valves.
If the steam flow forces prevent the initial closure of the isolation valve, the mctor control breakers will likely trip from the overcurrent condition before motor damage can occur.
Further, the isolation valve inside containment will not have been exposed to the steam environment from the broken line.
Resetting the motor control breaker would then permit energizing the valve motor and closing the isolation valve from the control room.
The second method available is to close one of the isolation valves by manual actuation of the hand crank. This would require suiting the operator in l~
l
. special garments and possibly using an airpack. Due to the expected high temperature in the torus compartment, the isolation valve inside the containment would be the valve most likely closed.
NED0-24708A (Ref. G) analyzes an unisolatable 0.5 square foot steam line break inside containment, which resembles a break of the 10 inch (0.55 square foot)
HPCI steam line from the time of the break up to the time that the low pressure systems would begin injection (225 seconds).
The analysis also includes the water injected by the RCIC, but this should be minimal.
The 0.5 square foot line break model predicts that the system pressure will fall below 300 psia at approximately 210 seconds after the break occurs.
The water level will still be above the core and the condensate and the condensate booster pumps can be used, for those systems having a turbine driven feed pump, to supply feedwater to the reactor.
For those feedwater systems having motor driven feed pumps, the feedwater system can supply feedwater continuously following the reactor trip.
With the feedwater system providing cooling water, the fuel will remain covered until a HPCI isolation valve is closed and the RHR system is restored to operation.
It is calculated that 12,500 gallons / hour of water at 94 F will be converted to steam at 212*F in absorbing the decay heat from the fuel. At this rate of consumption, a 500,000 gallon condensate tank could be emptied in 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
In order to maintain adequate coolant for the extended time period, the vacuum must be restored in the condenser and the decay heat dissipated using the condenser. This will also necessitate using the auxiliary boiler to provide steam for the gland seals.
Having the condensers available will reduce the i
steam pressure in the reactor reducing the amount of steam that will be discharged through the broken HPCI steam supply line, thus decreasing the
. consumption of water from the condensate storage tank.
This action will also lower the amount of heat and humidity being dumped into the torus compartment.
The probability of the loss of the feedwater during a 168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> interval, the time assumed necessary to restore the RHR system following a HPCI steam supply line break, is calculated to be 0.03.
This is based on the Browns Ferry IREP (Ref. D) frequency of transients that result in loss of feedwater (s1.4 per reactor year). This equates to a mean cime between failure of 5,570 hours0.0066 days <br />0.158 hours <br />9.424603e-4 weeks <br />2.16885e-4 months <br />.
Assuming an exponential distribution, a failure rate of
~4 1.8 x 10 per hour results.
Of concern is the operator actions needed to maintain the operation of the main feedwater system.
Although this is an activity with which the operator should be very familiar, detecting that the HPCI is not providing make-up inventory may not be immediate.
Further, the inventory in the hotwell must be maintained by flow from the condensate storage tank.
To obtain an adequate flow, it may be necessary to reestablish the vacuum in the condensers.
As reported in NUREG/CR-3933 (Ref. H), most recent PRAs assign a probability of 0.1 for failure to recover the power conversion system in a short interval.
In this accident, the time needed to make the necessary operating adjustments will not be as short as required for transients or small breaks in liquid coolant lines.
In addition, approximately one-fourth or one-half of the make-up water require-ments will be provided by one or two pump operation of the control rod drive hydraulic system. Thus, a human error probability of 0.05 will be assigned.
The total probability of failing to maintain coolant inventory with the feedwater systems for 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> is 0.05 + 0.03 = 0.08.
Thus upper and lower bound estimates are
-6 (5 x 10-5) x (1) x (1) x (0.08) = 4 x 10 core-melt / reactor year.
(5 x 10-5) x (1 x 10-3) x (1 x 10-3) x (1) x (0.8) =
-12 4 x 10 core-melt / reactor year.
Closing the outboard isolation valve and opening the supply valve, is assumed to result in no net change in the unavailability of the HPCI and, therefore, the frequency of other accident sequences is unchanged.
Closing the outboard isolation valve until the HPCI is commanded does not reduce the accident rate from breaks that occur when the HPCI is energized or go undetected prior to the HPCI being energized. With the inclusion of a bypass line to prevent thermal shock, this contribution is believed to be much smaller than the long-term exposure with the line pressurized.
Hence, the remaining contribution was not considered to be significant.
The Brookhaven estimate (Ref. J) of the frequency of a core melt accident due to an unisolated break outside the containment in a six inch RWCU line was 1.4E-05/RY. The study also conservatively assumes that the conditional probability for the isolation valves failing to close given a line break is 1.0.
Consequence Estimate A break in the HPCI steam supply line would b'e a LOCA outside containment. This would be most closely equivalent to the PWR Event V sequence identified in WASH-1400. The consequences are obtained using the CRAC (Calculation of Reactor Accident Consequences) Code.
Assuming an average population of 340 persons per square mile (which is the average for U.S. domestic sites) from an exclusion area out to one-half mile about the reactor on out to a 50 mile radius about the reactor.
Typical Midwest site meteorology is assumed.
Based upon these assumptions, a 6
release produces an exposure of 5.0 x 10 man-rem. With an upper and lower bound
-6
-12 frequency of 4 x 10 and 4 x 10 core-melt per reactor year, the upper and
_5 lower values of exposure risk is then 20 man-rem per reactor year ahd 2 x 10 man-rem per reactor year.
Based upon an average life of 24 years for each BWR l
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. E having a HPCI system with open isolation valves which involves 24 reactors, the risk posed by this issue has an upper bound of 11,500 man-rem and a lower bound
-2 The consequences of the RWCU line break sequence would of 1.1 x 10 man-rem.
be 70 man-rem per reactor year and 40,000 man-rem total.
Cost Estimate Industry cost estimates for implementation of the proposed change to leave the outboard isolation valve closed is estimated to be 2.5 person years. This includes an engineering review of the logic for HPCI initiation to assure that the valve will be commanded open and will properly isolate if required; preparation of chaN es to procedures (normal and emergency); revision to and operator training covering the change; revision to the FSAR; license amendment; and hardware changes.
No added maintenance costs are anticipated.
No hardware costs were assessed to add a bypass line because it is believed that most reactors already have this feature.
At an average cost of $100K per person-year, the total industry cost is $6.75M.
The NRC cost is estimated to be 1 person-month per reactor, or $210K. However, there is at least one reported instance in which the isolation valve could not be opened under pressure (Ref. E).
If these valves would have to be modified to open under pressure, the costs would be much greater.
Performing qualification tests on a selected sample of RWCU isolation valves and actuators and demonstrating by analyses that the other valves and actuator combinations will perform satisfactorily is estimated to cost $1,000,000.
If actuators would have to be replaced, this would add to the costs.
_g_
Value Impact Assessment At the upper bound values of 11,500 and 40,000 man-rem the value impact score S, is 51,500 man-rem 3=
($6.75 + $0.21 + $1.0) million s
6500 man-rem /$ million
-3 At the lower bound value of 1.4 x 10 man-rem the value impact score S, is
-3 1.4 x 10 man-rem 3=
($6.75 + (3.21) million
-4
= 2 x 10 man-rem /$ million OTHER CONSIDERATIONS Other considerations, which in individual cases may reduce the risk associated with this issue, include the absence of ventilation openings or open doors between the torus compartment and the pump rooms. The absence of these openings reduces the common cause failure potential of the RHR/LPCI, RCIC and core spray system with the HPCI steam supply line break.
Consideration should be given for reducing the risk if the isolation valves were selected given the requirement to close under line break steam mass flow conditions. This concern could be eliminated if it could be shown by test or from actual application that valve operation has been verified under loads equivalent to line break conditions.
A similar situation exists for the RCIC system.
Since the RCIC steam line is smaller than the HPCI, the risk may not be as great but would still' add sub-stantially to the values estimated previously.
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CONCLUSION At the upper bound with both the RWCU and HPCI event sequences and the Event V consequences, this issue would be a HIGH priority item.
Also its occurrence results in the loss of one defense layer (containment).
In addition, the costs to determine the capability of the valves to provide containment isolation are relatively small.
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. REFERENCES A.
Memorandum for Themis P. Speis from Roger J. Mattson, " Request for Prioritization of Generic Safety Issue-Failure of HCI Steam Line Without Isolation," October 18, 1983.
B.
Telephone conversation between Dick Hickman of GE, San Jose and J. W. Pittman of NRR/ DST /SPEB on January 6, 1984.
C.
WASH-1400 (NUREG-75/014) " Reactor Safety Study, An Assessment of Accident Risks in Commercial Nuclear Power Plants," U.S. Nuclear Regulatory Commission, October 1975.
D.
NUREG/CR-2802, " Interim Reliability Evaluation Program: Analysis of the Browns Ferry, Unit 1, Nuclear Plant," U.S. Nuclear Regulatory Commission, July 1982.
E.
LER 84.018 Browns Ferry Power Station Unit 1, April 3, 1984.
F.
Letter to A. Schwencer, Chief Licensing Branch No. 2, DL, NRR, from
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John S. Kemper, V.P., Philadelphia Electric Co., " Limerick Generating Station, Units 1 & 2 Request for Additional Information from NRC Equipment Qualification Branch (EQB)," February, 1984.
G.
NED0-24708A, " Additional Information Require:1 for NRC Staff Generic Report on Boiling Water Reactors," General Electric Company, December 1980.
H.
NUREG/CR-3933, " Risk Related Reliability ReqJirements for BWR Safety Important Systems with Emphasis on the Residual Heat Removal System,"
U.S. Nuclear Regulatory Commission, August 1984.
12 -
J.
BNL TECHNICAL REPORT A-3740, "An Evaluation of Unisolated LOCA Outside the Drywell in the Shoreham Nuclear Power Station," Brookhaven National Laboratory, June 1985.
K.
Memorandum for W. Minners from A. Thadani, " Comments on Generic Issue No. 87 - Failure of HPCI Steam Line Without Isolation," June 28, 1985.
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ENCLOSURE 2 l
GENERIC ISSUE MANAGEMENT CONTROL SYSTEM The Generic Issues Management Control System (GIMCS) provides appropriate information necessary to manage safety related and environmental generic issues through technical resolution and completion.
For the purpose of this management control system technically resolved is defined as the point where the staff's technical resolution has bsen issued.
Generally, speaking, this occurs when the technical resolution has been incorporated into one or more of the following:
(a)
Commission policy statement / orders (b)
NRC Regulations (c)
Standard Review Plan lGenericLetter Regulatory Guide i
GIMCS is gart of a.n integraded system of reports and procedures that would manage generic safety issues, TMI-related issues, and proposed new generic issues through the stages of prioritization, technical resolution, development of new criteria, review and approval, public comments, and incorporation into the Standard-Review Plan (SRP), as appropriate.
NUREG-0933 provides an a
Qvaluation for a recommended priority listing based on the potential safety significance and cost of implementation ~for each issue; NRR Office Letter Number 40 provided procedures and briteria for adding new generic issues to the system; and.GIMCS provides proposed scheduling for resolving and completing issues on the prioritized listing.
GIMCS will provide information to manage and control issues that are ranked High-priority generic issues, Medium-priority generic issues, issues for wh'ich possible resolution has been identified for evaluation, issues for which a technical resolution is available (as documented
.. -by memorandum, analysis, NUREG, etc.), and issues designated by the Director of NRR as issues for which resources have been made available for resolution and completion.
Issues ranked as either " Low" or " Dropped" are not allocated resources. Therefore, there is no resolution to be tracked by GIMCS.
Some new generic issues prioritized and processed in accordance with NRR Off' ice Letter No. 40 may not have resources allocated for resolution and completion.
These issues will be listed in GIMCS as inactive issues.
These will generally l
be Medium priority issues that have no safety deficiency demanding high-priority i
attention, but there is a potential for safety improvements or reduction in uncertainty of analysis that may be substantial and worthwhile.
Efforts for j
resolution of these issues will be planned,.over the next several years, but on a basis that will not interfere with the resolution of High-priority 4
generic issue work or other high priority work.
Thus, some (Medium) generic issues will be inactive until such time as resources become available to resolve the various issues.
As resource allocations are direc'.ed at issue resolution, they will become active.
The detailed schedule for resolvi'ng and completing the generic issue will be developed and monitored by the management control system.
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Management and control indicators used in GIMCS are. defined as follows:
Generic Issue Number.
1.
Item No.
Safety, Environmental or Regulatory Impact i
2.
Issue Type -
High, Note 1 or Note 2 (From NUREG-0933),
- Medium, p
3.
Action Level -
Degree of management attention needed to pr6 cess generic issues in accordance with established schedules L1 - No management action is neces ary L2 - Division Director action is necessary L3 - Director NRR action is necessary 1st listed has lead responsibility for re-4.
-Office /Div/Br -
solving issue, others listed have input to resolution.
5.
Task Manaaer' -
Name of assigned individual responsible for schedule updating.
Each issue should be assigned a TAC f.
6.
Tac Number -
Generic Issue Title.
7.
Title,,
8.
Work Authorization -
Who or what authorized work to be done on generic issue.
.. 9.
Contract Title -
Provide Contract Title (if contract issued).
- 10. Contractor Name/
Identify Contractor Name and FIN Number (as FIN No. -
appropriate).
If contract is not yet issued, indicate whether the contract is included in the FIN plan.
Describes briefly the work necessary to tech-
- 11. Work Scope -
nically resolve and complete the generic issue.
Identifies. documents that the technical resolution
- 12. Affected Documents will be incorporated into to identify new criteria.
Describes current status of work.
- 13. Status -
- 14. Problem / Resolution -
Identifies potential problem areas and describes what actions,are necessary to resolve them.
- 15. Technical Resolution -
Identifies detailed schedule of milestone dates that are required for completing the issue through the issuance of the SRP revision or other' change that documents requirements.
Milestones -
Selected significant milestones.
The " original" schedule remains unchanged.
Changes in schedule are listed under " Current".
Actual completion
.are listed under " Actual".
TYPICAL MILESTONES Other Division Involvement Oriainal Current Actual o
Date information requested from Division o
Date received from Division Contractor Information o
Proposal Solicited o
Proposal Evaluated and Ac,cepted Contract Sghedule, if applicable o
o Testing Schedule, if applicable o
Draft HUREG/CR report from contractor / consultant Staff. review of draft NUREG/CR report Value Impact Statement prepared (coordinated with SPEB and RRAB as applicable)
Final report prepared by Division (include SPEB preliminary comments and SRP revision)
2 wks Final report forwarded to DST for processing
...---------- 2 wks CRGR Package to NRR Director for Review
1 mo OMB Clearance obtained concurrently if applicable 1
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-2 Review Package to CRGR
1 mo CRGR review and EDO approval completed
1 mo Federal Register Notice of Issuance of SRP for Public Comment
3 mo Division review of public comments completed
_________ - 2 wks Comments incorporated and transmitted to DST for
. processing
2 wks Final CRGR package to NRR Director for review
1 mo Review Package to CRGR
1 mo CRGR review and EDO approval completed
.----------- 1 mo Federal Register Notice of Issuance of SRP
GENERIC ISSUE MANAGEMENT CONTROL SYSTEM Issue Issue
. Action Task Number Type Level Office /Div/Br Manaoer Tac No Active-L1 NRR/.
TBP TBP Title ----------------
Work Authorization ---
Memorandum to from H. R. Denton dated l
Contract Title -------
To Be Provided.
Contractor Name/
FIN No. ---,--------
To Be Provided.
Wo r k Scop e -----------
To Be Provided.
Affected Documents ---
To Be Provided.
{
Stat ~us ---------------
To Be Provided.
i Problem / Resolution ---
To Be Provided.
iechnicalResolution-To Be Provided.
Milestones.
Original Current Actual New Issues - Schedule To Be Developed l
1 1
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l f
As of First Quarter FY-84
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CFFICE $
mawr) ent)
-