ML20040A340
| ML20040A340 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 02/05/1977 |
| From: | Lyon R, Oswald A, Ralkovetz F EG&G, INC. |
| To: | |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201200789 | |
| Download: ML20040A340 (20) | |
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t REVIEW OF BABCOCK AND WILCOX p
SYSTEM OPERATING SEQUENCE DIAGRPiS AND FAILURE MODES AND EFFECTS ANALYSES i
AS DOCUMENTED IN THE STANDARD PLANT SAR 1
B-SAR-205 1
by
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R. E. Lyon A. J. Oswald February 1977 Systems Research Division Probabilistic and Risk Evaluation Branch EG8G Idaho, Inc.
l DATE: [/68 APPROVED BY:
e F' // Bal covert, M PrMabilistic an Evaluation Branch 4
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> tib DATE:
F/I 77 L.
L". Wheat, Vanager Regulation Support Branch 8201200789 810403 PDR FOIA MADDEN 80-515 PDR
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ABSTRACT l
The Sys. tem Operatina Sequence Diagrams and Failure Modes and Effects Analyses contained in B-SAR-205 have been reviewed to assess completeness and consistency with system ds~i9n, peR5inance criteria
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and safety analyses. Several areas of concern were identified and are discussed.
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INTRODl!CTION k
Appendix 15C of Babcock and Wilcox (B&W) Standard Safety Analysis Report (SAR) B-SAR-205(I) contains a series of flow diagrams called System Operating Sequ:nce Diagrams. These diagrams are prepared for each of the transients analyzed in Chapter 15, Accident Analyses, of the SAR and represent the actions r: quired to bring the plant to a stable condition following initiation of the transient. The diagrams identify the systems which are necessarf to achieve the stable condition as well as the approximate sequence of actions.
Also given are special requirements on the systems, such as the requirement that selected systems must be imune to a single active failure.
Section 6.3 of the SAR contains Failure Modes and Effects Analyses (FMEA's) which have been included to demonstrate that sufficient redundancy exists in the Emergercy Core Coolino System (ECCS), assuming that a !.0CA has occurred coincidentally with the failure i
i active component within the ECCS. Active components such as electrical circuits, pump motors, and valve operators have been considered in the analyses.
The scope of this study was to review the diagrams and FMEA's to detemine if the necessary systems have been included, that appropriate conditions have been placed on the systems and that the diagrams, descriptions and the correspondinc l
analyses are consistent. A basic assumption in the review was that a single active failure occurs in one of the systems following the initiation of the l
svent considered. The Chapter 15 analyses also assumed that only safety grade systems are operating following the initiation of the transient, and no credit is taken for any actions as a result of nomal operation of non safety arade l
equipteent or control systems.
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DISCUSSION (J
A review of the System Operating Sequence Diaorams revealed severel areas of potential concern regarding the adeauacy of the diagrams. These items are discussed below. Althouah not specifically included in the scope of this task, the review of supportinn analyses and other documentation in conjunction with the diagrams, generated several additional coments related to the supporting information. These items are also listed in the following discussion. Some of the items discussed below are general in nature and are so identified.
Others are applicable to a specific transient and are identified with the figure number of the associated System Operating Scquence Diagram.
(1) reneral 1
B&W considers pressure relief valves and check valves as beina passive devices and thus not considered during the active failure analysis. The Reactor Safety Study, WASH-1400(2), classes them as active components with failure rates comparable to those of pumps, valves, etc.
If a failure of this type is considered, it could have i
an effect on system availability, in particular pressurizer and secondary safety relief valves and the core flood and low pressure injection systems.
l (2) General The failure modes and effects analyses (FMEA's) presented by B&W in Table 6.3-7 to establish the effect of sinole active failures appear not to have considered the effect of the initiating j
event on the availability of the system. Many of the results l
l listed under the coments heading of the Table will vary wide ly l
l dependinn on the initiating event.
It would seem that in creir to l
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be meaningful, that the different initiating events should bc
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considered in the FMEA's and that the Section 15 analyses sho11d show that these configurations, or at least the limiting conflauration, have been analyzed and found acceptable.
(3) General The System Operating Sequence Diagrams show the s.vsteihs which must be single failure proof, but in general, it is unclear from mst of the analyses what single failures have been considered
- what affect they have on the system, and if they are the worst case single failure.
(4) General As a single failure, perhaps it might be appropriate to consiier fthe possibility that manual valves might be left in the wrong position, undetected, until the accident occurs (e.g. the system I
(test valves in the LPIS). The Reactor Safety Sturiy shows that this
-s ievent has a high probability of occurrence.
s (5) General In order to meet the single active failure requirement, it is necessary for the breakers on several valves to be racked out to minimite the possibility of inadvertent actuation. At the present time, it appears that all the necessary valves have been ccvered, but the requirements are contained in several locations, i.e., the FPIA, Section 6. Section 7, etc. These requirements will eventualle be ti.cluded in the Technical Specifications, but it would greatl e facilitate review of this and later documents if they were co'lected in a single location at this time.
(6) General Many of the diagrams show actuation of the pressurizer and/or l
l secondary safety valves.
Is it necessary, or do t le analyser assume, l
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diagrams, with the appropriate sinale failure designations.
(7) General Many diagrams contain the note that it is assumed in the anal esis that the turbine generator is tripped via the Control o d Drive Co1 trol o
System (CRDS) after a reactor trip is actuated.
Is th'is a wo-st case condition or is it a necessary condition to achieve the esults of the analysis? If it is a necessary condition *. hen the appropriate i
component blocks and single failure designations.hould be aMed to the diaorams.
j (P.) Chemical and Volume Control System (CVCS) Malfunction (15C.4)
A recent revision to rinure 9.3-1 has added an alternate sour:e cf demineralized water which bypasses the makeup tan'.
It is pr)haMe that this line is used in a shutdown or refuelina mode, with the Engineered Safeguards Actuation Signal (ESFAS) bypassed; thus, the extent of the transient is no longer limited 'y the capacity of the makeup tank.
In this case the operator must he relied or to terminst<
deboration and prevent criticality from occurrino.
(9) Chemical and Volume Centrol System Malfunction (15C.4)
The analysis during power operation assumes that the reactor trip closet the makeup tank outlet block valves and tenninate-the deborttion.
The diaoram should reflect this action with approoriate ESFA! system entries and single failure designations.
(10) Loss of Coolant Accident (LOCA) (150.13)
Figure 9.3-1 shows and ESFAS-A input to valves Va3A, V43B ant V43D.
The input to V43D is probably a draw.ina error, but if not, a failure l
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in the ESFAS-A system could prevent opening the three lines <ollowing "
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a break in the fourth line, resulting in a complete loss of I PI Cow.
(11) Large LOCA (15C.13-1)
The diagram does not contain a sequence for reactor trip or 'or f
startup of the auxiliary feedwater system.
'Q (12) Small LOCA (15C.13-3)
The diagram assumes that the operator manually isolates Hioh
,6 Pressure Injection (HPI) Supply lines which are affected if 1he break is in an HPI line.
Does the operator have sufficient time er'd can he be relied on to accomplish this? Several items will 1end to
[j hinder completion of this action:
(a) The break location will not be apparent until HPI is initiated and flow is established in the supply lines.
(b) Because of the cross-connect lines inside containment it will r
be necessary to isolate two supply lines.
If the subseruent single failure is loss of a vital power source, one of these lines must be isolated by closina the valve with the hardwheel located at the valve. The discussion in Chapter 6 and the analysis in Chapter 15 do not address this particular atcident.
J (13),Small LOCA (15C.13-3)
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! The diagram does not contain a sequence for main steam syster-s.,
isolation.
In addition, since piggyback operation is assumer', the Low Pressure Injection (LPI) system is not used for initial core cooling as shown on the diagram. The LPI sequence would more s
appropriately enter the diagram at the point at which the optrator realigns the system for piggyback operation.
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(14) Core Flood Tank Line Break (15C.13-4) ' "-
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.i The diagram does not contain a sequence;for,, main steam system
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isolation.
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s-e (15) bye [ pressurization of Decay Heat RembdfSystem (15C.24) f N'
i As noted in a previous Esport(3), the system design may be such I
u that it is not kqne,^o this event as stated on the diagram.
i s-(16) Loss of Cordens'er Vahuum s.
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- he turbine trips on hiah condenser pressure.
Is this inhe' rent,or is some control action required?
If the turbine i
did not trip would this be more or less conservative? If les!
conservative, then any control actions required to trip the i
turbine should be single failure proof.
In this case it woul/ be appropriate to have the systems necessary to cause the turbint trip, with the applicable single failure designations, shown r n the diagram, and then reference Figure 15C.7 for the remainder of the systems.
(17) Loss of Instrument Air (15C.31)
The diagram states that no actions are required to support the Nuclear Stum system (NSS). ' As noted in a preyious report (3) failure of the instrumkM air may pr' event nonnal cooldown of the reactor coo' ant system.
(18) Inadvertent Closure 'cif Main $ team Isolation Valve (15C.35)
There is a dfsagreement between the diagram and the Chapter l';
analysis as togthe limiting event for this transient. The diagram t
i refers to Fig'ure 15C.7, the turbine trip, while the analysis
._ refers to Section 15M.D. the loss of normal feedwater system.
e (19) Makeup Line Break (15C.3f1) l There is no correspondits descriptioh of the accident in SectIon 15.1, y
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but it would appear from the B&W response to NRC ouestion 212.147 that f
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the diagram is incomplete. The operator relies on certain alkms and l
indications to inform him of the condition. These should be rhown l
l on the diagram along with the appropriate single failure desir nations.
l (This may be inferred in the balance of plant safety related contml and instrumentation (B0p SRCI) controls boxes, but this is not clear).
Also the reactor trips on low pressurizer level. This sequence should also be shown on the diagram.
(20) Cold Shutdown Systems (15C.39)
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As noted in a previous report
, some of the systems may not be
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(21) Overpressure Transient There is no corresponding description of the accident conside vd in Section 15.1, but it would appear that the diagram may not be adequate.
l especially in the mode where the decay heat removal system is not in operation.
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REFERENCES l
Standard Safety Analysis Report, B-5AR-205, Babcock and Wilcox, Docket No.
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i STN-50-561, knendment 7 December 15, 1976.
2.
Reactor Safety Study An Assessment of Accident Risks in U.S. Comercial j
i Nuclear Power Plants, WASH-1400, U.S. Nuclear Regulatory Comissier,
October 1975.
Review of Babcock and Wilcox Systems, Phase I, Reactor Shutdown Coclina, 3.
B-SAR-205, Informal Letter Report Stig-106-76, December 7,1976.
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i,,O As there is no ava'itable operating experience with t System 80 F
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,ctor ecolant pumps, we will require as par test program a
-)g. verify tne performanca of the pump motor thrust 7
test te verify the analysis.in Appendix A of ENPD-201 The test will bearing during a loss Q of component cooling water to the cooling coils in the sump oil reser-g:.b, Ue will require that this test on a prototype pump be conducted nir.
4 fre-a nominal initial temperature until the temperature of a lube oil gt. reaches a value which permits the conservatism in the analysis to be assessed. For this purpose, sump temperatures in the range of 200*F h
wou'd be tested. During this pericd, component cooling water to the
@ pu:p shaf t seal asse-bly shculd also be terminated. To the greatest T g extent possible, the test should be conducted under reactor operating
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RECULATORY POSITION tt i
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Our acceptance of the Combustion Engineering topical report i
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CD.'?D-201, " Performance of C-E System 80 Reactor Coolant Pump With less of Component Cooling Water," is subject to the following conditions.
As a part of the test program of the System 80 Reactor Coolant Punps, n(
Cambustion Engineering will demonstrate that with loss of component i<
f cooling water to the pump seal assembly and pump-motor thrust bearings,
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the design limit of the seal assembly, or damage to the thrust bearings i
y that could af fect the pump coastdewn characteristics.
Satis facto ry y/
M1ution of the pump tcat as defined above would provide assurance l.
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- omponent quality and permit the cooling lines to the reactor coolant l,
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pumps to be constructed to Quality Group D (Safety Class 4) and designed
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In the event the System 80 reactor coolant pumps are unable to f
perform as predicted, we will require thh: the component cooling lines
- o the pump seal assembly and the pump-motor thrust bearings be constructed to Quality Group. C (Safety Class 3) in conformance with Regulatory Guide 1.26, and designed to seismic Category I requirements in conformance with Regulatory Guide 1.29.
i In addition, a utility applicant that utilizes a System SO nuclear steam supply system should provide administrative controls that are j
I defined in the plant technical specifications and which will require that:
(1) the af f ected ruactor coolant pumps be. shutdown 30 minutes l
after loss of component cooling unter, (2) the affected reactor ceolant l
I pumps will not be restarted until component cooling water.is restored i
and pump thermal conditions are normal, and (3) prior to shutdown of the last reactor coolant pump, a sufficient amount of boron will be introduced into the reactor coolant system to facilitate cooldown to ter,idual heat r.anoval system cperating conditions.
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?ll.f In Tables 3.2-4 and 5.1-2, and in Figure 5.5-8 your Quality Group D 3.2) classification and non-seismic Category 1 classification of the reactor coolant pump auxiliaries that are an integral part of each reactor coolant pump assembly is not in conformance with Regulatory Guides 1.26 and 1.29 and is unacgeptable.
In Table 3.2-4 and in Section 5.5.1 of the SSAR provide a detailed list and description of those components that are generally identified in Table 3.2-4 as reactor coolant pump auxiliaries.
It is our position that those components that are an integral part v
of the reactor coolant pump, such as the oil lubrication system, oil coolers, seal coolers and cooling water lines to the it:terface point with the component cooling water system should be classified Quality Group C, constructed to ASME Section III, Class 3 and designed to seismic Category 1 requirements.
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It is also our position that the following criteria 4 rc: 5 1
est lli U foi that portion of the component cooling water system which interfaces with the reactor coolant pumps to supply cooling water to pump seals and bearings during normal operation, anticipated transients, and following accidents:
(1) A moderate energy leakage crack or a ingle failure in the component. cooling water system shall not result in fuel damage
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caused by an extended loss of cooling to the reactor coolant j
pumps. Single failures include operator error, spurious actuation of motor operated valves, and loss of component cooling water pumps.
Moderate leakage cracks should be determine'd in accordance with the g'uidelines of Branch Technical Position APCSB 3-1.
(2)
An accident that is initiated'
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from a failure in the component cooling water system piping shall not result in excessive fuel damage or a brekch of the reactor X
coolant system pressure boundary when an extended loss of cooling to the reactor coolant pumps occur.
A single active failure shall be considered when evaluating the consequences of this accident.
Moderate leakage cracks should be determined in accordance with Branch Technical Position APCSB 3-1.
In order to meet the criteriu established above, a B SAR-205 interface X
requirement should be imposed on the balance of plant component cooling water system that provides cooling water to the reactor coolant pump seals and bearings so that the system will meet the following conditions:
(1) A period of 20 minutes is considered acceptable within which an operator can trip the reactor coolant pumps and initiate a safe plant shutdown.
Therefore, in nrder to provide an adequate
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-ut coolant pumps are designed so that they can operate i!O. k;; :f 1i cooling water for a minimum period of 30 minutes without loss of function or the need for reactor operator action, h the com-5 ponent cooling water system should be designed to meet the i
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following.
(a) Seismic Classification: Seismic Category 1 (b) Quality Classification: Quality Group C Component Code: ASME Section III, Class 3 (c) Single Failure: Should be capable of withstanding a single active failure and should be capable of withstanding a A
moderate energy leakage crack in accordance with Branch Tech-nical Position APCS 3 3-1, with respect to cooling reactor coolant pumps or else, item (1)(d) must be implemented.
A single failure includes malfunctioning of any valve or pump in the component cooling water lines to the reactor coolant 1
pumps.
(d)
Instrumentation and Controls:
Safety grade instrumentation is required to detect loss of comptnent cooling water and initiate automatic protection of the plant.
(e) Containment Isolation of Systems:
Only when reactor coolant pumps are not functioning.
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For a reactor coolant pump design that can operat p 4t' '^r M cooling water for longer than 30 minutes without loss of function -
or the need for reactor operator action, Jdf{p the component cooling V
water system should be designed to meet the following:
i (a) Seismic Classification:
Non-seismic Category I, except for that portion of the component cooling water system that forms an extension of the containment boundary.
i (b) Quality Classification:
Quality Group D, except for that portion of the component cooling water system that foir.s an extension of the containment boundary.
F (c) Single Failure: The system should be capable of withstanding C
a single active failure and must be capable of withstanding a moderate energy leakage crack in accordance with Branch Technical Position APCSB 3-1, with respect to cooling' reactor coolant pumps, or else, item (2)(d) must be implemented.
A single failure includes malfunctioning of kvalves or pumps in the component cooling water lines to X
the reactor coolant pumps.
(d)
Instrumentation and Controls:
Safety grade instrumentation is required to detect the loss of component cooling water and provide an alarm in the control room to satisfy item (2)(c).
(e)
Containment Isolation of Systems:
Only when reactor coolant pumps are net functioning.
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t The reactor coolant pumps are within the B-SAR-205 scope of' supply; i^
l therefore, in order to demonstrate that a reactor coolant pump design can operate with loss of component coolii;g water for longer than 30 i
minutes without loss of function or the need for reactor operator action, provide the following:
(1) A detailed description of the events following the loss-of
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component cooling water to the reactor coolant pumps and am.
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demonstrat at there are no consequences important to safety which may result from this event.
Include'a discussion of the effect that the loss of coolia water to the seal coolers has on the reactor coolant pump seals.
Show that the loss of cooling water does not result in a loss-of-coolant accident
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due to seal failure.
(2) A detailed analysis to show that loss o,f cooling water to the reactor coolant pumps and motors will not cause a loss of the flow coastdown characteristic or cause seizure of the pumps)
K assuming no administrative action is taken. The response should include a detailed description of the calculation procedure including:
(a) The equations used_.
(b) Tne parameters used in the equations, such as the design para-meters fe,r the motor bearings, motor, pump and any other equip-ment entering into the calculation, and material property values for the oil and metal parts.
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(c) A discussion of the effects of possible variations in part dimensions and material properties, such as, bearing clearance f.,
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tolerances and misalignment.
l (d) A description of the cooling and lubricating systems (with
- i appropriate figures) associated with the reactor coolant pump and motor and the design criteria and standards therefore.
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(e) 'Information to verify the applicability of the equations and material properties chosen for the analysis (i.e., references should be listed, and if empirical relations are used, provide -
a comparison of their range of application to the range used in the analysis).
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Should an analysis be provided to demonstrate that loss of component cooling water to the reactor coolant pumps and motor assembly is x-
. acceptable, we will require certain modifications to the plant technical specifications and a reactor coolant pump test conducted under operating conditions and with component cooling water terminated for a specified i
period of time to verify the analysis.
l
!11.E In Table 3.2-4, the fuel transfer tube gate valve is incorrectly l
~ 3.2) identified as Safety Class III-2.
This component should be classified Safety Class III-4.
Revise Table 3.2-4 accordingly.
.11 7 In Table 3.2-4, the deborating demineralizer of the chemical addition
- 3.2) and boron recovery system is incorrectly identified as Quality Group D.
This component should be classified Quality Group C.
Revise Table 3.2-4 accordingly.
l 11..
In Table 5.2-1, delete the last sentence of the footnote and add a l3.2.W3) new sentence as follows:
"The code edition and addenda identified i
l above for each component are minimum requirements and will be
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superfededasapplicablebasedonautilityapplicant'sdocketdate i
of his PSAR or the actual purchase order date for a specific component -
in compliance with the rules of 10 CFR Part 50, Section 50.55a."
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.In Table 5.2-2, the use of Code Case 1448-2, "Use of Case Interpretations (5.2.1.4) of ANSI B31 Code for Pressure Piping,Section III, (8/14/72)", is acceptable, except for Piping Code Case 80, referenced therein, which I
has not been approved for use by NRC.
Clarify the use of Code Case 1448-2 in Table 5.2-2.
7 111. 5 Code Case 1572, " Fracture Toughness, Class 1 Components,Section III",
l5.2).4) l was annulled 3/8/74 and the provisions of the code case were published
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in the Winter 1973 Addenda to Section III.
Therefore, Code Case 1572 t
should be deleted from Table 5.2-2 unless it is your intent to i
utilize this code case in conjunction with the 1971 Edition of Section III.
T 11.1 The following code cases have been superseded by a later code case 5.2.1.4) revision than those identified in Table 5.2-2.
Review these code cases and revise Table 5.2-2 where applicable.
CODE CASE IDENTIFIED CODE CASE - LATEST IN TABLE 5.2-2 REVISION 1337-9 1337-10 1484-1 1484-2 1644
~1644-2 1618 (Supp. 1) 1618-1 1335-8 1335-9 1553 1553-1 m
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11.4 Code Case 1603, " Toughness Tests When Cross-Section Limits; l
5.2.1.4) f Orientation and Location of Specimens,Section III" was annulled 5/10/74 and the provisions of the code case were published in the
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I Summer 1974 Addenda to Section III.
Therefore, Code Case 1603 should be deleted from Table 5.2-2.
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Ill.f In. Table 5.2-1, the 1971 Edition of Section III of the ASME Boiler l5.2.1.3) and Pressure Vessel Code that you have identified as applicable to:
(1) Reactor Vessel, (2) Steam Generators, and (3) Pressurizer, is not in conformance with the Codes and Standards Rule, Section 50.55a of 10 CFR Part 50.
,_(s Based on a docket date of 3/1/76 for this application, these components as a minimum should be constructed to the 1974 Edition of Section III.
Revise Table 5.2-1 accordingly.
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>:11. It In Table 9.3-2, the component code for the Makeup Tank of the Makeup
- 9.3) and Purification System is incorrectly identified as ASME Section III, Class 3.
This component should be constructed to ASME Section III, Class 2.
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?ll.M In Tables 9.2-3 and 9.3-5 add the seismic classification of the
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components identified therein.
13 "11. M In Table 9.3-5, the Quality Group and component code for the Deborating 9.3)
Demineralizer of the Chemical Addition and Boron Recovery System is J
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t 2// - S A incorrectly identified as Quality Group D and the component code as s
ASME Section VIII. This component should be classified Quality Group C and constructed to AStiE Section III, Class 3.
If ill. M In Tsbles 3.2-4 and 9.3-5, the Quality Group, component code and Seismic 9.3)
Classification for the Reactor Coolant Degasifier Package of the Chemical Addition and Baron Recovery System is incorrectly identified as Quality Group D, the component code as ASME Section VIII and non-seismic Category 1.
This component should be classified Quality Group C, constructed to ASME Section III, Class 3, and designed to seismic Category 1 requirements.
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Add the following components of the Chemical Addition and Boron Recovery 3.2)
System that are shown in Figure 9.2-2, Sheet 2 to Table 3.2-4.
These components are:
(1) R.C. Bleed Holdup Tank, and (2) R.C. Distillate Storage Tank.
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- ll. V In Figure 10.1-1, Sheet 2, Secondary Plant System, identify the changes
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in Quality Group classifications at the appropriate valves in the main steam, feedwater and auxiliary feedwater lines.
'7 111.1&
In Section 10.5.2, item 6, the reference to isolation valve V20 in 10.5.2)
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Figure 10.0-1, is incorrect.
The correct reference is Figure 10.1-1.
In addition, revise the last sentence of item 6 as follows: "All components and piping in the condensate storage facilities for the auxiliary feedwater system shall be Seismic Category I and designed to comply with ASME Section III, Class 3 requirements.
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