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NUCLEAR REGULATORY COMMISSION wasmwarow. o.c.2ases e
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DOCKET NO. STN 50-561 VEND 0R:
BABC0CK & WILCOX COMPANY (B&W)
SUBJECT:
SUMMARY
OF MEETING TO DISCUSS SYSTEM OPERATING SEQUENCE DIAGRAMS AND FAILURE MODES AND EFFECTS ANALYSES, BSAR-205 On March 24, 1977 representatives of B&W met with the NRC staff to discuss concerns identified in a report to the NRC by EG&G Idaho, Inc.
An Attendance List is enclosed.
EG&G was engaged by the NRC (Reactor Systems Branch) to review the subject portions of the BSAR-205 standard plant Safety Analysis Report.
EG&G completed their task with the transmittal'6f a ~l6tter ~~
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report to the NRC on February 22, 1977. This letter report was made available to B&W on March 17, 1977, and was the. basis for this 3
meeting. The meeting was intended to clarify both the staff and B&W i
understanding of the coments made by EG&G, prior to the staff's taking any position on any of the matters identified in the report.
v R. E. Lyon of EG&G attended the meeting to facilitate this understanding.
The 21 comments listed in the Discussion section of the report (Enclosure 2) were discussed in order. Following are sunmary remarks on each item. Actions to be taken as a result of the discussion are
, described.
(1) B&W stated that their design philosophy, to consider pressure relief and check valves as pasrive devices which are not subject to single " active" failure assumptions, is consistent. with the ASME Code and with the General Design Criteria. G.Ma;etis(NRC) stated that the staff held similar views regarding spring loaded or other non-powered check valves, but considered power-operated devices (such as an electrically operated pressurizer relief valve) to be subject to single active failure. The staff will give further consideration to B&W's analyses with respect to the identification of single active failsres.
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(2) B&W stated that BSAR-205 system design is such that their inclusion of the " worst case" single failure in a failure mode and effects analysis effectively bounds the expected effects from all initiating events while meeting the Comissions' criteria '.e., to assume single rather than multiple active failures. They used as an example the fact that equipment in containment is designed to withstand the effects of the LOCA (initiating event) while acting to mitigate the consequences of the event: The staff will initiate further dialog with EG&G on the specific nature of EG&G's general comment. Additional single failure analyses may be required in the BSAR-205.
(3) The staff has been aware of, and in agreement with, the B&W fonnat for the System Operating Sequence Diagrams. The diagrams were not intended to document the details of the single failure analyses but rather to show in an abbreviated fonn what systems are required to be single failure proof.
(4) B&W stated that their cesign is predicted on the Regulatory Guide 1.47 " Bypassed and Inoperable Status Indication for Nuclear Power Plant Systems". They feel that the recommendations of this guide are applicable to manual valves within a system intended to provide a safety action. This guide essentially
.1 recomends the provision of control room indication of safety system unavailability when any required element of the system is bypassed or otherwise set in a way that would cause system
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inoperability. The responsibility for implementation of the recommendations of the regulatory guide will lie with the applicant referencing the BSAR-205 design.
G. Mazetis (NRC) stated that curr.ent staff policy in this area OL and its application to BSAR-205 will be reviewed prior to a detennination on this comment.
(5) B&W referred the staff to Table 6.3.7 and their response to request number 212.112.
G. Mazetis (NRC) recommended that the response to 212.112 be incorporated into the table, which will put this type of information in one location. B&W agreed to do this.
(6) B&W stated that where makeup is an essential part of the single failure analysis, it is included. The EG&G report does not identify gf specific deficiencies in this regard; a re-check for specific OA deficiencies will be made by EG&G and the staff.
4 (7) B&W stated that turbine trip was sometimes assumed even though not necessary to the safety sequence. Where turbine trip is a requirement, it is shown on the diagrams and is actuated by the ESFAS.
R. Lyon, (EG&G).. stated that there was no further concern in this area.
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(8) B&W stated that the alternate source of demineralized water is to be used only during relatively rapid reactor power changes during power operation. They will add such a description to the BSAR-205 Chapter 15 analyses and discuss the potential for inadvertant operation. A description of the alternate source design will be included in Chapter 9. The alterr. ate source.
will be accounted for in the systems operating sequence diagram on tFc Chemical 4.nd Volme Control System in Appendix 15 C.
(9) This comment by EG&G is related to item 8 above. B&W will indicate, in the new descriptive material to be added to BSAR-205, that deboration is not terminated automatically.
(10) This item concerns a drawing error which has been corrected.
(11) B&W stated that their analysis for the large LOCA does not take credit for the reactor trip or startup of the auxiliary feedwater system, therefore those actions are not included on the system operating sequence diagrams 15 C.13-1.
(12) B&W stated that the operator has 20 minutes to accomplish the necessary manual action. They will add a detailed description of the actions required, with justification that the time available to complete the action is sufficient.
13 & 14)B&W stated that the diagrams do not contain a sequence for main
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steam system isolation because that action is not required, even
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though assumed in analyses. B&W will clarify their assumption in the text. B&W agrees with the coment on the LPI sequence, and will revise the diagram, g.
(15) The NRC staff wished to consider the overpressurization potential ye '
of the BSAR-205 design further. This subject will be the topic of separate discussions with B&W. Q ~ e_rr dah 4* d O UT.Ls A w a@3 (16) B&W stated that this topic was outside the BSAR-205 design scope.
They stated that NSSS safety was not dependent on turbine trip since secondary side cooling would be accomplished through atmospheric dump and relief valves if necessary. NRC staff will deliberate further on the EG&G concern.
(17 & 20)
B&W stated that the EG&G coments are important if shutdown to unpressurized conditions (" cold shutdown) is required. B&ii reiterated that their design philosphy is that safety grade systems are required only to take plant to a safe " hot" shutdown state.
G. Mazetis noted that the staff is continuing to move toward the requirement to go to " cold" shutdown with only safety grade equipment.
(18) B&W stated agreement with this coment by EG&G and comitted to a correction of a figure reference to maintain consistency.
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B&W acknowledged that they do not describe a make-up line break accident in Chapter 15. They will revise system operating sequence diagran 15 C.38 to incorporate the response to request number 212.227, and will include a reference to a sequence of events discussion elsewhere in the BSAR-205 text.
(21)
B&W stated that the system operating sequence diagram 15 C.40 is based on their position as expressed in responses to request numbers 212.222 and 212.229, which responses also reference the BSAR-205 text at sections 15.1.24 and 7.6.1.1 G.Mazetis(NRC) felt that the staff may already have the information necessary from B&W in order to arrive at a staff position in this matter.
The staff will evaluate B&W's responses and may ask for more discussion with B&W.
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Thomas H. Cox, Project Manager Light Water Reactors Branch No. 3 Divisten of-Project Management
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Enclosure:
1.
Attendance List 2.
Part II, Discussion, from EG&G Idaho, Inc. Report To NRC dated February 22, 1977 t
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ENCLOSURE 1 ATTENDANCE LIST MEETING - B&W & STAFF - 03/24/77 Name Organization T. Cox NRC R. Lyon EG&G G. Brazill B&W J. Happell B&W R. Schomaker B&W L. Cartin B&W R. Brockman B&W S. Newberry NRC G. Mazetis NRC J. Hamilton B&W i
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II. DISCUSSION A review of the System Operating Sequence Dianrams revealed several l
Thr se areas of potential concern regarding the adecuacy of the diaarams.
items are discussed below. Althouah not specifically included in the scope of f
this task, the review of supportinn analyses and other docunentation in conjunction with the diagrams, generated several additional coments related to the supporting information. These items are also listed in the followinc discussion. Some of the items discussed below are ceneral in nature ard are so identified. Others are applicable to a specific transient and are identified with the figure number of the associated System Operating Sequence Diagram.
(1)' General
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BAW considers pressure relief valves and check valves as beira passive devices and thus not considered durino the active failure The Reactor Safety Study, WASH-1400(2), classes them as analysis.
active components with failure rates comparable to those of pumps, valves, etc. If a failure of this type is considered, it could have an effect on system availability, in particular pressurizer and secondary safety relief valves and the core flood and low pressure injection systems.
(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 initiatina i
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event on the availability of the system. Many of the results I
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v listed under the coments headino of the Table will vary widely dependino on the initiating event.
It would seem that in orr3r to
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be meaningful, that the differant initiatino events should be considered in the FMEA's and that the Section 15 analyses she aid show that these configurations, or at least the limitino conflouratic.
have been analyzed and found a.cceptable.
(3) General The System Operating Sequence Diagrams snow the s." stems which must be single failure proof, but in general, it is unclear from r)st of the analyses what single failures have been consirlered/ what iffect they have on the systen, and if they are the worst case sinali failure.
(4) General As a sinnie failure, perhaps it might be appropriate to consiier the possibility that manual valves might be left in the wrono position, undetected, until the accident occurs (c.p. the system test valves in the I.PIS). The Reactor Safety Stuty shows that this i
event has a high probability of occurrence.
(5) General In order to meet the single active failure requirnnent, it is necessary for the breakers on several valves to be racked out to minimite the possibility of inadvertent actuation. At the present tine, it appears that all the necessary valves have been cevered, but the requirements are contained in several locations, i.e., the FF".A, Section 6, Section 7, etc. These requirements will eventualli be included in the Technical Specifications, but it would great 1 s facilitate review of this and later documents if they were co lected in a single location at this time.
(6) General j
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Many of the diagrams show actuation of the pressurizer and/or I
secondary safety valves.
Is it necessary, or do t 3e analyser assinne,
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that makeup is provided? If so, this should be included in tte diagrams, with the appropriate sinale failure designations.
(7) General Many diagrams contain the note that it is assumed in the anal isis tr'-
l the turbine generator is tripped via the Control *.od Drive Coitrol Is this a wo-st System (CRDS) after a reactor trip is actuated.
case condition or is it a necessary condition to achieve the esolv If it is a necessary condition *. hen the approprian of the analysis?
component blocks and single failure desionations should be arHed to the diaorams.
Chemical and Volume C(introl Syste.n (CVCS) Malfunction (15C.4)
(P.)
A recent revision to Fiaure 9.1-1 has added an ale.ernate sour:e cf
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demineralized water which bypasses the makeup tank.
It is pr)hable that this line is used in a shutdown or refuelino mode, with the Engineered Safeguards Actuation Signal (ESFAS) bypassed; thus, the extent of the transient is no longer limited 'y the capacity of In this case the operator must e relied on to termnata the makeup tank.
deboratinn and prevent criticality from occurring.
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Chemical and Volume Control System Malfunction (15C.4)
(9)
The analysis during power operation assumes that the reactor trio closes l
the makeup tank outlet block valves and teminates the deborttior.
The diagram should reflect this action with appreoriate ESFAS system entries and single failure designations.
(la) Loss of Coolant Accident (LOCA) (15C.13)
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Figure 9.3-1 shows and ESFAS-A input to valves Va3A, V43B anr The input to V43D is probably a drawina error, but if not, a failure i
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Paae 7 in the ESFAS-A syster could prevent openino the three lines ollowino l
I a break in the fourth line, resulting in a complete loss of f PI flow.
(11) Large LOCA (15C.13-1)
The diagram does not contain a sequence for reactor trip or 'or startup of the auxiliary feedwater system.
(12) Small LOCA (15C.13-3)
The diagram assums that the operator manually isolates Hioh Pressure Injection (HDI) supply lines which are affected if the break is in an HPI line. Does the operator have sufficient time ard can he be relied on tn accomplish this? Several items will 'end to hinder completion of this action:
- g-(a) The break location will not be apparent untti HPI is l'E 7;w initiated and flow is established in the supoly lines.
(b) Because of the cross-connect lines inside containment 1' will be necest sry to isolate two supply lines.
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single failure is loss of a vital power source, one of these lines must be isolated by closina the valve with the hardwheel
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located at the valve. The discussion in Chapter 6 and the analysis in Chapter 15 do not ad<iress this particular accident.
(13) Small LOCA (15C.13-3) l The diagram does not contain a sequence for main steam syster l
i isolation. In addition, since piggyback operation is assume (, the Low Pressure Injection (LPI) system is not used for initial core cooling as shown on the diagram. The LPI sequence would mort
- 3 appropriately enter the diagram at the point at which the ope rator realigns the system for piggyback operation.
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(14) Core Flood Tank Line Break (15C.13-4)
The diagrvn does not contain a sequence for main steam system isolation.
Overpressurization of Decay Heat Removal System (15C.24)
(15)
As noted in a previous report (3), the system desion may be surh that it is not immune to this event as stated on the diaoram.
(16) Loss of Condenser Yacuum The analysis assumes that the turbine trips on hiah condenser pressure.
If the turbine Is this inherent.or is some control action required?
If les!
did not trip would this be more or less conservative?
conservative, then any control actions required to trip the turbine should be single failure proof.~ In this i se it woulf be appropriate to have the systems necessary to cause the turbint trip, with the applicable single failure designations, shown rn 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 Asnotedinapreviousreport(3), failure of the Steam System (NSS).
instrument air may prevent nomal cooldown of the reactor coo' ant system.
Inadvertent Closure of Main Steam Isolation Yalve (15C.35)
(18)
There is a disagreement between the diagram and the Chapter 14 The diagram analysis as to the limiting event for this transient.
refers to Figure 15C.7, the tureine trip, while the analysis t
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refers to Section 15,1,8, the loss of nomal feedwater system.
l (19) Makeup Line Break (15C.38) v There is no corresponding description of the accic'ent in Section 15.1, i
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Attachment to 1tia-62-77 Pace 9 212.147 that but it would appear fr)m the B&W resoonse to NRC cuestion The operator relies on certain als rms and the diagram is incomplete.
These should be 'hown indications to inform him of the condition.
on the diagram along with the appropriate single failure desienations.
1 (This may be inferred in the balance of plant safety related control and instrumentation (BqP SRCI) controls boxes, but this is no : clear).
This sequen :e Also the reactor trips on low pressurizer level.
should als; be shown on the diagram.
(20) Cold Shutdown Systems (15C.39)
As noted in a previous report (3), some of the systems may not be sinple failure proof.
(21) Overpressure Transient There is no corresponding description of the accident conside ed in Section 15.1, but it would appear that the diapram may not be adequate, especially in the mode where the decey heat removal system is not in operation.
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N'IN:
Mr. James H. Taylor l
4 Manager of Licensing Nuclear Power Generation Division i
P., O. Box 1260 i
Lynchburg, vi.Jinia 24505 i
i cc: Washington Public Power Supply System l
ATIN: Mr. N. O. Strand Managing Director (Acting)
P. O. Box 968 f
3000 George Washington Way Richland, Washington 99352 Mr. Robert Borsum Bethesda Representative Babcock & Wilcox Nuclear Power Generation Div1Gion Suite 5515, 7735 Old Georgetown Road Bethesda, Maryland 20014 i
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B. G. Shultz Project Engineer Stone & Webster Engineering Corporation P. O. Box 2325.
(j Boston, Massachusetts 02107.._ _ _,
I Mr. W. E. Kessler Comonwealth Associates, Inc.
j 209 East Washington Jackson, Michigan 49201 I
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Robert J. Kafin, Esq.
i 115 naple Street t
Glen Falls, New York 12801 l
l Mr. B. M. Miller Ohio Edison Company i
76 South Main Street Akron, Ohio 44308
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[D MEETING
SUMMARY
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Docket File H. Denton tiRC PDR D. M ller y
Local PDR Ject Manager / m (fy#4 LWR-3 File Attorney, OELD
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7 TIC M. Rushbrook NRR Reading J. Knight B. Rusche D. Ross E. Case T. Novak R. Boyd R. Tedesco R. DeYoung R. Bosnak D. Vassallo S. Pawlicki D. Skovholt R. Bosnak J. Stolz I. Sihweil K. Kniel P. Check O. Parr Z. Rosztomczy S. Varga IE ( )
R. Denise G. Lainas R. Clark V. Benaroya T. Speis.
V. Moore P. Collins T. Ippolito C. Heltemes R. Vollmer R. Houston M. Ernst L. Crocker F. Rosa J. Miller W. Gammill
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R. Heineman ACRS(16)
NRCParticipants A Na+W L A n r a__>
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DOCKET NO. STN 50-561 VENDOR:
BABC0CK & WILCOX COMPANY (B&W)
SUBJECT:
SUMMARY
OF MEETING TO DISCUSS SYSTEM OPERATING SEQUENCE DIAGRAMS AND FAILURE MODES AND EFFECTS ANALYSES, BSAR-205 On March 24, 1977 representatives of B&W met with the NRC staff to discoss concerns identified in a report to the NRC by EG&G Idaho, Inc.
An Attendance List is enclosed.
EG&G was engaged by the NRC (Reactor Systems Branch) to review the subject portions of the BSAR-205 standard plant Safety Analysis Report (SAR). EGtG completed their task with the transmittal of a letter report tc the NRC on February 22, 1977. This letter report was
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made available to B&W on March 17, 1977, and was the basis for this meeting. The meeting was intended to clarify bot the, staff and B&W understanding of the coments made by EG&G, prior to the staff's taking any position on any of the matters identified in the report.
R. E. Lyon of EG&G attended the meeting to facilitate this understanding.. -
The 21 comments listed in the Discussion rection of the report (Enclosure 2) were discussed in order. Following are stmnary reinarks on each item. Actions to be taken as a result of the discussion are described.
(1) B&W stated that their design philosophy, to consider pressure relief and check valves as passive devices which.are not subject to single " active" fa110re assumptions,' is consistent with the ASME Code and with the General Design Criteria.
G. Mazetis (NRC) stated that the staff held similar views regarding spring loaded or other non-powered check valves, but considered power-operated devices (such as an electrically operated pressurizer' relief valve) to be subject to single active failure. The staff will give further consideration to B&W's analyses with respect to the identification of single active failures.
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i (2) B&W stated that BSAR-205 system design is such that their inclusion of the " worst case" ' ingle failure in a failure s
mode and effects analysis effectively bounds the expected effects from all initiating events while meeting the Commissiors' i
criteria i.e., to assume single rather than multiple active
' failures. They used.as an example the fact that equipment s'ft.W containment is designed to withstand the effects of the LOCA (ini'tiating event) while acting to mitigate the consequences of the event: The staff will initiate further dialog with EG&G on the specific nature of EG&G's general com. ment. Additional single failure analyses may be required in the BSAR-205.
(3) The Staff has been abre of, and in agreement with, the B&W format for the System Operating Sequence Diagrams. The diagrams were not intended to document the details of the single failure analyses but rather to show in an abbreviated fom what systems are required to be single failure proof.
(4) B'&W sbted that their design is predicted on the Regulatory
. Guide 1,47 " Bypassed and Inoperable Status Indication for
- duclear Power Plant Systems". They feel that the recommendations of this guide are applicable to manual valves within a system intended to provide a safety action. This guide essentially recommends the provision of control room indication cf safety system unavailability when any required element of the system is bypassed or otherwise set in a way that would cause system inoperability. The responsibility for implementation of the recommendations of the regulatory guide will lie with the applicant referencing the BSAR-205 design.
G. Mazetis (HRC) stated that current staff policy in this area and its application to BSAR-205 will be reviewed prior to a detemination on this comment.
6 (5) B&W referred the staff to Table 6.3.7 and their response to request number 212.112.
G. M tis (NRC) recommended that the response to 212.112 be incorporated into the table, which will put this type of information in one location. B&W agreed to do this.
(6) B&W stated that where makeup is an essential part of the single failure analysis, it is included. The EG&G report does not identify specific deficiencies'in this regard; a re-check for specific deficiencies will be made by EG&G and the staff.
(7)
B&W stated that turbi e trip was sometimes assumed even though not ll necessary to the safdFy sequence. Where turbine trip is a requirement, it is shown on the diagrams and is actuated by the ESFAS.
R. Lyon, (EG&G), daMdet! that there was no further concern in this area.
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(8) B&W stated that the alternate source of demineralized water is (S
to be used only during relatively rapid reactor power changes during power operation. They will add such a description to the BSAR-205 Chapter 15 analyses and discuss the potential for inadvertent operation. A description of the alternate so ce design will be included in Chapter 9. The alternate sourc ill will be accounted for in the systems operating sequence d agram on the Chemical and Volume Control System in Appendix 15 C.
(9) This comment by EG&G is related to item 8 above. B&W will indicate, in the Jew descriptive material to be added to BSAR-205, the*. deporation is not terminated automatically.
(10) This item concerns a drawing error which has been corrected.
(11) B&W stated that their analysis for the large LOCA does not take credit for the reactor trip or startup of the auxiliary feedwater system, therefore those actions are not included on the system operating sequence diagrams 15 C.13-1.
(12) B&W stated that the operator has 20 minutes to accomplish the m
necessary manual action. They will add pfg' detailed description of the actions regyied, with justification that the time available to complete the action is sufficient.
(13 & 14)B&W stated that the diagrams do not contain a sequence for main steam system isolation because that action is not required, even though assumed in analyses. B&W will clarify their assumption in the text. B&W agrees with the coment on the LPI sequence, and will revise the diagram.
(15) The NRC staff wished to consider the overpressurization potential on the BSAR-205 design further. This subject will be the topicg of separate discussions with B&W.
(16) B&W stated that this topic was outside the BSAR-205 design scope.
They stated that NSSS safety was not dependent or, turbine trip since secondary side cooling would be accomplished through atmospheric dump and relief valves if necessary. NRC staff will deliberate i
further on the EG&G concern.
(17 & 20)
B&W stated that the EG&G coments are important if shutdown to unpressurized conditions (" cold shutdown) is required. B&W reiterated that their design philosphy is that safety grade systems are rgquired only to take plant to a safe " hot" shutdown state.
G. M4ttis noted that the staff is continuing to move toward the requirement to go to " cold" shutdown with only Safety grade equipment.
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(18) B&W stated agreement with this coment Eby EG&G and comitted to a correction of a figure reference to maintain consistency.
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B&W acknowledged that they do not describe a make-up line break accident in Chapter 15. They will revise system operating sequence diagrams 15 C.38 to incorporate the response to request number 212.227, and will include a reference to a sequence of events discus ~sion elsewhere in the BSAR-205 text.
(21)
B&W stated that the system operating. sequence diagram 15 C.40 is based on their position as expressed in responses to request numbers 212.222 and 212.229, which responses also reference the BSAR-205 text at sections 15.1.24 and 7.6.1.1. G. Mazetis (NRC) felt that the staff may already have the information necessary from B&W in order to arrive at a staff position in this matter.
The staff will evaluate B&W's responses and may ask for more discussion with B&W.
Tom H. Cox, Project Manager Light Water Reactors Branch No. 3 Division of Project Management
Enclosure:
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Attendance List
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Part II, Discussion, from x
EG&G Idaho, Inc. Report To NRC dated February 22, 1977 t
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A_TTENDANCE LIST MEETING - B&W & STAFF- 03/24/77 Name Organization T. H. Cox NRC-DPM R. E. Lyon EG&G G. Brazill B&W-Plant Integration (BSAR)
J. J. Happell B&W-Licensing (BSAR)
R. J. Schemaker B&W-Safety Analysis L. R. Cartin B&W-Plant Integration R. J. Brockman B&W-BSAR Project Management S. Newberry NRC J. Mazetis NRC J. Hamilton B&W
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