ML20054K029
| ML20054K029 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/24/1982 |
| From: | James Shea Office of Nuclear Reactor Regulation |
| To: | Counsil W NORTHEAST NUCLEAR ENERGY CO. |
| References | |
| TASK-03-05.A, TASK-3-5.A, TASK-RR LSO5-82-06-095, LSO5-82-6-95, NUDOCS 8206300315 | |
| Download: ML20054K029 (17) | |
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June 24, 1982 Docket No. 50-245 LS05-82 095 tir. W. G. Counsil, Vice President Nuclear Engineering and Operations Hortheast Nuclear Energy Conpany Post Of fice Box 270 Hartford, Connecticut 06101
Dear !!r. Counsil:
SUBJECT:
MILLSTOME 1 - SEP TOPIC III-5. A EFFECTS OF PIPE BREAX OH STRUCTURES, SYSTEtiS AND COMPO!.EllTS INSIDE C0llTAINf1ENT In your letter dated April 13, 1982, you subnitted a safety assessrent report on the above topic. He have completed our evaluation, which is enclosed. We conclude that the plant is adequately protected fron'the dynanic effects of pipe break inside containment subject to resolution of the following in the Integrated Plant Safety Assessnent:
A.
Clarification of jet inpingerent criteria, B.
Further justification of adequacy of pipe whip interaction with the dryuell liner and containnent wall, C.
Basis for acceptability of consequences of cascading breaks, and D.
Any significant changes in total pipe stresses as a result of SEP Topic III-6, Seisnic Design Considerations.
The need to actually inplerent changes as a result of these itens will be deternined during the integrated safety assessment. This safety evaluation nay be revised in the future if your facility design is changed or if HRC QQ cp,g 'g criteria relating to this topic are nodified before the integrated assess-c nent is conpleted.
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Docket No. 50-245 LS05-82 ffr. W. G. Counsil. Vice President fluclear Engineering and Operations Northeast fluclear Energy Company Post Office Box 270 Hartford, Connecticut 06101
Dear Mr. Counsil:
SUBJECT:
MILLSTONE 1 - SEP TOPIC III-5.A. EFFECTS OF PIPE BREAK Off STRUCTURES, SYSTEMS AND COMPONENTS INSIDE C0fiTAINfiENT In your letter dated April 13, 1982, you submitted a safety assessment report on the above topic. We have completed our evaluation, which is enclosed. We conclude that the plant is adequately protected from the dynamic effects of pipe break inside containment subject to resolution of the following in the Integrated Plant Safety Assessment:
A.
Clarification of jet impingement criteria, B.
Further justification of adequacy of pipe whip interaction with the drywell liner and containment wall, C.
Basis for acceptability of consequences of cascading breaks, and D.
Any significant deviations identified in future resolution of SEP Topic III-6 concerning seismic analysis be recon-ciled to this assessment as addressed in Section B.2 of this SER.
The need to actually implement changes as a result of these items will be deternined during the integrated safety assessment. This safety evaluation may be revised in the future if your facility design is changed or if flRC criteria relating to this topic are modified before the integrated assess-ment is completed.
Sincerely, SEPB:DL ORB #5:PM ORB #5:BC AD:SA:DL WRussell JShea DCrutchfieldames Shea, Project Manager Glainas 6/ /82 6/ /82 6/ /82 Operating Reactors Branch flo. 5 6/ /82 Division of Licensing J
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Mr. W. G. Counsil cc William H. Cuddy, Esquire State of Connecticut Day, Berry & Howard Office of Policy & Management
' Counselors at Law ATTN:
Under Secretary Energy One Constitution Plaza Division Hartford, Connecticut 06103 80 Washington' Street' Hartford, Connecticut 06115 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission Region I Office 631 Park Avenue King' of Prussia, Pennsylvania.19406 Northeast Nuclear Energy dompany
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ATTN:
Superintendent Millstone Plant P. O. Box 128 Waterford, Connecticut 06385 Mr. Richard T. Laudenat Manager, Generation Facilities Licensing Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 Resident Inspector c/o U. S. NRC P. O. Box Drawer KK Niantic, Connecticut 06357 Ffrst-Selectman of the Town of Waterford Hall.of Records 200 Boston Post Road Waterford, Connecticut 063B,5 John F. Opeka Systems Superintendent Northeast Utilities Service Company P. O. Box 270 Hartford, Connecticut 06101 U. S. Environmental Protection Agency Region I Office ATTN:
Regional Radiation Representative JFK Federal Building Boston, Massachusetts 02203 O
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SEP EVALUATION 0F i
EFFECTS OF PIPE BREAK ON STRUCTURES,.
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SYSTEMS AND COMPONENTS INSIDE CONTAINMENT t
TOPIC III-5.A MILLSTONE NUCLEAR POWER STATION, UNIT N0. 1 e
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TABLE OF CONTENTS I.
INTRODUCTION II.
REVIEW CRITERIA III.
RELATED SAFETY TOPICS AND INTERFACES IV.
REVIEW GUIDELINES V.
EVALUATION A.
BACKGROUND B.
REVIEW CRITERIA USED ON THE MILLSTONE.
NUCLEAR STATION UNIT NO. 1
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. 1.
HIGH ENERGY SYSTEMS 2.
PIPE BREAK LOCATION AND TYPE 3.
PIPE WHIP AND JET IMPINGMENT C.
PIPE BREAK EFFECTS ON STRUCTURES D.
EFFECTS ON SYSTEMS AND COMP 0NENTS 1.
SYSTEMS UTILIZED FOR ACCIDENT MITIGATION AND SAFE SHUTDOWN 2.
INTERACTION ANALYSIS VI.
CONCLUSIONS e.
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I.
INTRODUCTION The safety objective of Systematic Evaluation Program (SEP) Topic III-5.A, " Effects of Pipe Break on Structures, Systems and Components Inside Containment," is to assure that pipe breaks would not cause the loss of required function of " safety-related" systems, structures and components and to assure that the plant can be safely shut down in the event of such breaks.
The required functions of " safety-related" systems are those functions required to mitigate the effects of the pipe break and safely shut down the d
plant.
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II.
REVIEW CRITERIA General Design Criteria 4 (Appendix A to 10 CFR Part 50) requires-in part that structures, systems and comp'onents important to safety be appropriately protected against dynamic effects, such as pipe whip and discharging fluids, that may result from equipment failures.
The current criteria for review of pipe breaks inside containment are contained in Standard Review Plan 3.6.2, " Determination of
- Break Locations and Dynamic Effects Associated with the Postulated Repture of Piping," including its attached Branch Technical Position, Mechanical Engineering Branch 3-1 (BTP MEB 3-1).
III.
RELATED SAFETY TOPICS AND INTEftFACES 1.
This review complements that of SEP Topic VII-3, " Systems Required for Safe Shutdown."
2.
The en'vironmental effects of pressure, temperature, humidity and flooding due to postulated pipe breaks are evaluated under SEP Topic III-12, " Environmental Qualification of Safety-Related Equipment."
3.
The effects of potential missiles generated'by fluid system ruptures and rotating machinery are evaluated under'SEP Topic III-4.c,
" Internally Generated Missiles."
4.
The effects of containment pressurization are evaluated under SEP Topic VI-2.D, " Mass and Energy Release for Possible Pipe Break
'Inside Containment."
5.
The original plant design criteria in the areas of seismic input, analysis, and design ctiteria are evaluated under SEP Topic III-6,
" Seismic Design Consideration."
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IV.
REVIEW CUIDELINES The licensee's. break location criteria and methods of analysis for evaluating postulated breaks in high energy piping systems inside
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containment have been compared with the currently accepted review criteria as described in Se: tion II above. The_ review relied upod information submitted by the Licensee, Northeast Nuclear Energy Company (NNECO)u in Reference 1.
The scope of review under this topic was limited to avoid duplication of effort since some a'spe' cts of the topic vers previously reviewe'd by
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the staff or are.. included under other.SEP topic _(see III above).
When deviations from the review criteria are identified, engineering
' judgement is utilized to evaluate the consequence of postulated pipe break and to assure that pipe break would not cause the loss of required function of " safety-related" systems, structures and com-ponents and to assure that the plant can be safely shutdown in the event of such break.
V.
EVALUATION I
A.
BACKGROUND On July 20, 1978, the SEP Branch sent a letter (Reference 2) to KMC, Inc. requesting an analysis of the effects of postulated pipe breaks on structure, systems and components inside containment.
In that letter, the staff included a position that stated three approaches were appropriate for postulating breaks in high energy piping. systems 2
2 either P 275 asig or T 200*F.
The approaches are:
1.
Mechanistic 2.
Simplified Mechanistic
- 3.. Effects Oriented The staff further stated that combinations of the three approaches could be utilized if justified.
s In response to our letter, the licensee submitted Reference 1 con-cerning postulated high energy pipe rupture inside containment. The y.
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3 licensee'sstudyincludest$efollowing.
1.
Definition of the criteria and assumptions used in'the study.
2.
Identification of the high energy piping syst. ems inside the drywell.
3.
A discussion of the effects of postulated ruptures in each of t.he high energy systems.
4.
An evaluation of plant capcbility to mitigate the con-sequences of pipe breaks and to plac'e' the reactor 'in a safe shutdown condition.
B.
REVIEW CRITERIA USED ON THE MILLSTONE NUCLEAR STATION UNIT NO. 1 1.
HIGH ENERGY SYSTEMS The licensee has classified high energy fluid systems as those that are maintained under conditions where either or both the maximum operating-temperature and pressure exceed 2000F and 275 2
psig respectively during normal operation. This is consistent with current MEB criteria, Using the above criteria, the systems inside containment which were considered, are as follows:
a.
Isolation Condenser b.
Core Spray c.
Main Steam d.
CleanuT) Water c.
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Recirculation h.
Containment Cooling i.
Reactor Vent j.
Reactor Heat Cooling k.
Control Rod Driv.e 2.
PIPE BREAK LOCATION AND TYPE The licensee's study is based on the Simplified Mechanistic Approach which basically assumes a pipe rupture at the terminal.,
ends and at each intermediate weld in the system.
The 11censee's study also utilizes the Mechanistic Approach in certain cases.
where detailed stress analysis informatioh is available.
In Appendix G of Reference 1, the licensee has identified some postulated breaks of main steam lines that could result in unacceptable consequences based on the Simplified Mechanistic Approach.
For these piping systems, a mechanistic evaluation was then performed.
Based on its analysis, the licensee has concluded that these pipe break locations with unacceptable.
consequences are in fact, not the high stress points.
Therefore, the pipe rupture will not have to be postulated at these loca-tions on a mechanistic approach basis. This is acceptable in accordance, with Reference 2.
However, since the seismic pipe stress-has not been completely resolved, any significant changes
in total pipe stress as a result of Topic III-6 should be reviewed to demonstrate conformance with the licensee's conclusions in Appendi)c G of Reference 1.
Break types are postulated as follows, a.
Circumferential breaks are postulated to occur in runs of greater-than one inch nominal pipe size at each location in accordance with the Simplified Mechanistic Approach.
b.
Longitudinal breaks are postulated to occur in runs of greater
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than four inch nominal pipe size at each location in accordance with the Simplified Mechanis :1c Approach.
The above criteria are acceptable in accordance with Reference 2.
PIPE WHIP AND JET UhINGEMENT
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The following assu=ptions are used for the licensee's high energy pipe rupture interaction analysis, Pipe whip is assumed to occur as a result of a circumferential
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a.
rupture in a high energy system provided there is a s_ignificant reservoir of energy.
For circumferential breaks, the free end of a moving pipe is b.
assumed to move in only one direction parallel to its reactor force. This type of pipe break event does not cause dynamic instability (large amplitude oscillations) since the critical length required for this phenomenon is substantially greater than any major pipes in the drywell of BWR plants.
' Impacted active equipment (e.g, valves and instruments) is c.
considered unable to perform its intended function.
Pipe anchors and whip restraints are considered to be capable d.
of continuing to perform their intended functions.
Valves which are normally closed and are not signaled to open, e.
are not assumed to fail open. Valves normally open shall re-main open during and after loss of power or impact..
f.
Plastic hinge for=ation due to pipe rupture is assumed to occur at system anchors or at other intermediate locationg as dictated The by the compicxity of the particular system configur9 tion.
hinges can form in either bending or torsion mode depending on the configuration.
Longitudinal breaks are assumed to cause a jet in the form of a g.
cone with a 20' angle of divergence.
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A whipping pipe is considered to have cufficient energy to cause damage to:
. Pipes of smaller nominal size and lighter wall thickness.
. Electric notor operators.
. Electric condult and cable trays...
- i., A steam jet is considered to have sufficient energy to cause damage to:
. Electric cable trays.
. Electric motor operators.
- Based on a review of the above information, we have determined-thah the licensee's pipe whip and jet impingement analysis are, in general, consistent with the currently accepted standards except are follows:
In considering the pipe whip damage, it is acceptable to use the conservative assumption that the particular safety system becomes inoperative, irrespective of the actual energy which would be in-volved in the collision or the strength of the impacted piping system. Nevertheless, the staff's position is that it is necessary to consider all the possible sequences of cascading failures, i.e.,
the damaged target may damage another piping system and so forth, and to evaluate the overall effects of cascading f ailures such as,.
the eff ect on multiple blowdown transients and to provide a shutdown /
cooldown scenario for the cascading failures.
This subject is further discussed below in Section D.2.
With respect to jet impingement analysis, the licensee has utilized assumptions g and 1.
Assumption g specifically ref ers to "longitudhml breaks" when considering the j et expansion model.
It is the sta'f f ' p'osition that j et imping ement eff ects should be considered as a result of both circumf erential breaks and longitudinal breaks.
Fur t hermore, in the case of circumf erential breaks, j ets in conjunction with pipe whip should be considered to sweep the arc traveled during the whip.
The licensee should expand its evaluation to address the criteria used for j et impingement from circumf erential breaks.
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Assumption i only refers to " steam jet" when considering the jet impingement effects on electric cable trays and electric motor operators.
It is the staff position that jet impingement effects from all high energy lines, not just steam lines, are to be con-sidered. The licensee is requested to confi,rm,that such was the,
case in its analyses.
In addition, based on the 'information
- submiyted in Reference 1, it is not clear how.the licensee has assessed the jet impingement effects on the impinged target piping system including the. methodology of.calculat'ing the jet impingement loads.
,In accordance with staff positions transmitted on January 4,1980 (Reference 3), the effects of jet impingement should be considered and evaluated regirdless of the ratio'6f impinged and postulated b'roken pipe sizes. The licensee is requested to assess its evaluation given the staff position.
C.
PIPE BREAK EFFECTS ON STRUCTURES in Appendix E of Reference 1, the licensee submitted Chicago Bridge and Iron Company (CB&I) Test Report, " LOADS ON SPHERICAL SHELLS",
when considering pipe whip damage on containment integrity. The CB&I test indicates that when the_ containment wall is loaded over a
' large enough area, i.e. equivalent to a 14 inch diameter or larger circle, deformation of over 3 inches can occur without failure of the containment wall.
Based on this test result, the licensee concludes that for breaks occurring in piping of 14 inch diameter or greater, if contact occurred with the containment wall, the wall could deform without faifure until deformation was limited by the concrete shield wall. Accordingly, there would be no postulated break which results in containment wall penetration.
t Based the,information submitted, we.have determined that the licenset has not provided a valid rationalization to justify the use of CB&I test results in their case.
It is noted that this test was c
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It is not clear that the test result is also valid for the dynamic loading which would be experienced as a result of the postulated pipe whip for Millstone 1.
Specifically, the particular test applies a concentrated load of 235 tons over an area, equivalent to a 14 inch diameter or larger circle.
This assumption-may not always be valid because the impact area of a 14 inch diameter or larger pipe may be smaller than the assumed area.
Thus, our concern is that :Ln the case of applying a concentrated dynamic load over a small area the steel plate may be perforated before the deformation could b~e backed up by the ~ concrete' shield wall.
It is also noted that the CB&I test was performed on a spherical
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steel plate section for a 70-foot diameter sphere with a plate thickness of 0.75 inch.
However, the thickness of the Millstone 1 drywell liner is only five-eighth of an inch.
The licensee.
should select a worst case configuration to demonstrate that the impact load or energy produced as a result of postulated pipe break for piping of 14 inch diameter or greater does not exceed the load or energy required to penetrate the containment liner and wall.
In performing this evaluation with static analysis or static test, the dynamic load factor has to be considered.
The licensee can take into account the following considerationst a.
Actual liner thickness with respect to the impact location, b.
The combined crack propagation time and break opening time
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of the pipe may be long enough to depressurize the system such that the whipping pipe could not produce sufficient energy to penetrate the containment wall.
D.
EFFECTS ON SYSTEMS AND COMPONENTS 1.
Systems Utilized for Accident Mitigation and Safe Shutdown The systems that would be used to mitigate the effects of a high energy line break inside containment and reach staff shutdown are the following:
a.
Feedwater Coolant Injection (FWCI)
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Low Pressure Coolant Injection Containment Cooling-(LPCI)
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Core Spray (CS) d.
Isolation Condenser e.
Automatic Depressurization System (ADS).
f.
Emergency Power (gas turbine, diesel generator) g.
Control Rod Drive Pumps h.
CRD Mechanisms 1.
Vital Instrumentation J
This set of systems is also capable of performing a shutdown in the absence of.a break as described in the NRC safe shutdown.
systems report (Reference 4).
Under the review assumptions of loss of offsite power and a s, ingle active failure (gas turbine) the FWCI system would not be available. An assumed single failure of the normally closed isolation condenser isolation 4
7 valve would prevent its operation.
This valve is, however, capable of being manually opened from the reactor building.
Many of the components in the above systems are located outside containment and thus are not affected by an inside containment break.
Essential instrumentation inside the drywell consists of two redundant systems for reactor pressure and level located on opposite sides of the drywell. Within each system there exists redundancy so that a single active failure concurrent with a break affecting one system would not result in total loss of indication.
The only equipment inside containment that is required to operate to perform its function is associated with the ADS valves. For small breaks this system is used to depressurize the reactor so that low pr. essure ECCS systems can provide flow (if no high pressure makeup is available).
For large breaks the reactor coolant system depressurizes through the break.
For smaller line breaks, therefore, interactions with ADS valve controls were evaluated as described in the interaction matrix.
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10 2.
INTERACTION ANALYSIS The licensee has performed an interaction matrix to study the consequences of postulated pipe breaks on safety-related systems, structures and components.
The matrices are prepared on a system basis showing the potential interactions between the source, for each postulated break paint, anB the selected target.
Interactions are defined as follows:
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1.
Acceptable:
Interacticn causcs no damage, 2.
No Interaction:
Interaction physically not possible, 3.
Damage Possible:
Further evaluation required.
Each interaction following within the last category is further
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evaluated in Appendix C of Reference 1 to assure that the remaining equipment is capable of performing a safe shutdown under the review assumptions of loss of offsite power and single failure.
For several break points, multiple targets were identified as possibly damaged. We could not determine whether the source line damaged all of the identified targets directly or if cascading ef f ects as discussed in B.3 above were involved.
It also could not be determined for target piping whether the aff ected area is also high energy and, thus subj ect to possible cascading ef f ects or if damage to the target results only in e.
unavailability of that piping section.
If cascading breaks occur the reactor coolant pressure boundary may be breeched in more than one location and the resulting blowdown eff ects may not be bounded by the ref erence analyses.
For example, f or break point 10 in the "A" main steam line, interactions with possible damage were identified with the cleanup ret' urn line, part of the "A" f eedwater line, one of the
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recirculation risers, the "A" LPCI piping and one of the. main steam relief valve discharge lines. All of these lines are smaller than the 20 inch diameter steam line and thus are as-sumed to be damaged.
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In this case, f ailure of the gas turbine is the most limiting f ailure since it causes unavailability of the Th'CI,and of the "B" train of ECCS (LPCI and CS) since loss of of f site power has.
been assuped. The "A" train of core spray, which is delivered into the vessel through dedicated spray nozzles, would be available for core cool.1,ng.
The licensee should clarify how he has considered'c'ascading
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breaks including the basis for concluding that available mitiga-ting systems, in the example, one core spray system, are suffi-cient to handle the break consequences and maintain core cooling.
VI.
CONCLUSIONS Based on the information submitted by the licensee, we have reviewed the criteria pertaining to the locations, types and effect of postulated pipe breaks in high energy piping systems inside contain-ment.
We have concluded that the criteria used to define the break i
locations and types are in accordance with currently accepted standards.
We have also determined that it is acceptable under current SEP criteria to use the interaction study to evaluate the effects of postulated pipe breaks and to determine the accept-ability of plant response to pipe breaks.
However, we have found that the subject of the evaluation of jet impingement effect, the containment integrity evaluation and the consequences of cascading breaks as identified in Sections B.3, C, and D.2, respectively, have not been addressed adequately in l
the licensee's evaluation.
In addition, we reconcend any significant changes in total pipe stress as a result of SEP Topic III-6, Seismic Design Considerations, should be reviewed to demonstrate that higher seismic stresses do not result in new potential break locations.
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REFERENCES 1.
Report, " Millstone Nuclear Power Station, Unit No. 1 SEP TOPIC III-5. A, HJgh Energy Pipe Break Inside Containment," Northeast Nuclear Energy Company, dated April 13, 1982.
2.
Letter, D. K. Davis (NRC,'SEPB) to KMC, "AsYessment-6f Postulated Pipe Breaks Inside Containment for SEP Plants," dated July 20,..
1978.
3.
Letter, D. Ziemann to W.' C. Counsil, " Evaluation of Pipe Whip Impact and Jet Impingement Effects of Postulated Pipe Breaks for SEP Topic III-5.A and III-5.B,"
dated January 4, 1980.
4.
Letter, D. Crutchfield to W. Counsil. "SEP Topics V-10.B, V-11.B. VII-3, Safe Shutdown Systems Report", dated May 11,1981.
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