ML19320D473
| ML19320D473 | |
| Person / Time | |
|---|---|
| Site: | Atlantic Nuclear Power Plant  |
| Issue date: | 07/31/1980 |
| From: | OFFSHORE POWER SYSTEMS (SUBS. OF WESTINGHOUSE ELECTRI |
| To: | |
| Shared Package | |
| ML19320D463 | List: |
| References | |
| RTR-NUREG-0660, RTR-NUREG-660 36A93, 36A93-R1, NUDOCS 8007210404 | |
| Download: ML19320D473 (122) | |
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OFFSHORE POWER SYSTEMS RESPONSES TO POST-TMI NRC REQUIREMENTS l
TOPICAL REPORT NO. 36A93 (REVISION 1)
JULY 1980 80 07210'/of
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Topical Report No. 36A93 (Revision 1)
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INTRODUCTION
'Ibpical Report 36A93 was subnitted originally -in December 1979 and at that time provided responses to the recomendations of the Bulletins and Orders Task Force and the Lessons Learned Task Force.
Since that time, the NRC staff has compiled into one document, all of the recomendaticas fra the various groups which have investigated the 'IMI accident. %is document, NUREG-0660, was issued in May 1980 and constitutes the emprehensive action plan required by the Comission to resune reactor licensing. Revision 1 to Offshore Power Systems Topical Report 36A93, which is a emplete rewrite of the original report, responds to NUREG-0660.
Table 1 stmnarizes the action plan items of NUREG-0660 noting which items cb and do not apply to the manufacturirg license application. %e applica-bility categorization was developed jointly by offshore Power Systems and the NRC staff III I2)
Irans which are not applicable to the Manufacturing License application are not addressed in this report beyond their listing in Table 1.
v In nest cases there does not exist a concise statement which accurately states the intent of an action plan item with respect to a manufacturing license (or near-term construction permit) plant. For this reason, opical Report 36A93 (Revisica 1) contains only the response to each applicable item, and the reader is referred to NUREX3-0660 and associated docunents for a' statement for the NRC requirements.
(1)NRC Meeting Sumary by Mr. R. A. Birkel dated May 28, 1980.
(2) OPS letter, P. B. Haga to R. L. Baer, FNP-PAL-088, dated May 6,1980.
1 Revision 1
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L TABLE 1 i
SUMMARY
OF NUREG-0660 l
ACTION PLAN ITEMS l
M.L. APPLICABILITY l
NUREG-0660 ITEM NO.
SUBJECT YES - PALE NO. OF REMARKS NO
RESPONSE
1 I.A Operational Safety X
I.B.1 Management for Operations X
Subitem I.B.1-1 applies (See below) l I.B.1-1 Organization and Management 10 Imorovements I.B.2 Inspection of Cperating X
Reactors Y
I.C Operating Procedures X
Subitems I.C.1 and acolv (See belowl I.C.1 Short-Term Accident Analysis 12 nnd Procedure Rev.
4 I.C.5 Feedback of Operating Experience 15 I.D.1 Control Room Design Review 16 I.D.2 Safety Parameter Console 17 i
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i.D.3 System Status Monitoring 18 4
oo I.D.4 Control Room Design Standard 26 w
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-TABLE 1
SUMMARY
OF NUREC-0660 ACTION PLAN ITEMS M.L. APPLICABILITY NUREG-0660 ITEM'NO.
' SUBJECT -
YES - PAGE NO. OF RDIARKS NO
RESPONSE
I.D,5 Control Room Research X
I.D.6 Control Room Technology Transfer X
I.E '
' Analysis & Dissemination of
.X Subitem I.E.4 applies'
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Operating Experience (See below) i I.E.4 Coordination of Programs 49 ta l
I.F.1 Expanded QA List 50
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i I.F.2 Detailed QA Criteria 53 I.G Training During Low Power Testing X
3 II.A Siting X
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II.B.1-RCS Vents 56 g.
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II.B.2 Plant Shielding 57 o"
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II.B.3 Post-Accident Sampling 61 i
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TABLE 1 i'
SUMMARY
OF NUREG-0660 ACTION PLAN ITEMS M.L. APPLICABILITY NUREG-0660 ITEM NO.
SUBJECT YES - PAGE NO. OF REMARKS NO
RESPONSE
II.B.4 Training X
II.B.5 Research X
II.B.6 High Population Density Sites X
II.B.7 Containment Inerting X
II.B.8 Rulemaking in Degraded Core 62 Accidents II.C Reliability Engineering & Risk X
Subitem II.C.4 applies Assessment (See below)
II.C.4 Reliability Engineering 63 II.D.1 RCS Safety Valve Test Requirements 64 i
II.D.2 Demonstrate Valve Research 65 Q
Applies to FNP II.D.3 Safety Valve Position Indication 66 s
II.E.1-1 Auxiliary Feedwater Evaluation 67
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TABLE 1
SUMMARY
OF NUREG-0660 ACTION PLAN ITEMS M.L. APPLICABILITY NUREG-0660 ITEM NO.
SUBJECT YES - PACE NO. OF REMARKS NO RE.c?ONSE II.E.1-2 Auxiliary Feedwater Auto-Start &
71 Flow Ind.
r II.E.1 SRP and Reg. Guide on Aux.
X Feedwater II.E.2-1 Reliance on ECCS 72 i
II.E.2-2 ECCF Research X
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II.E.2-3 Uncertainties in LOCA Predictions 73 II.E.3 Deca;- Heat Removal X
Subitem II.E.3-1 spplies (See below)
II.E.3-1 Decay Heat Removal - Natural 74 Circulation II.E.4-1 Dedicated Ctat. Penetrations 76 II.E.4-2 Containment Isol. Dependability 77
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II.E.4-3 Containment Integrity Check X
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. II. E.4-4 Containment Purging 81 i
i TABLE 1 i
SUMMARY
OF NUREG-0660 ACTION PLAN ITEMS
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M.L. APPLICABILITY NUREG-0660 ITEM NO.-
SUBJECT YES - PACE NO. OF REMARKS NO
RESPONSE
II.E.5~
Sensitivity of B&W Reactors X
j II.E.6 In-Situ Valve Testing X
II.F.1 Additional Accident Monitoring-82 Instr.
1 II.F.2 Detection of Inadequate Core 86 Cooling e
II.F.3 Regulatory Guide 1.97 90 II. F. 4 Control & Protection Actions X
II.G Power for PORV, Isol. Vlv., and 96 PZR Level II.H.
TMI Cleanup X
Subiter: II.H.3 applies (See below) t II.H.3 Evaluation and Feedback 97 E
p II.J.1 Vendor Inspection Program X
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Construction Insp. Program X
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TABL 1
SUMMARY
OF NUREC-0660 ACTION PLAN ITEMS M.L. APPLICABILITY l
NUREG-0660 ITEM NO.
SUBJECT YES - PAGE NO. OF REMARKS NO
RESPONSE
II.J.3 Management for Design & Construc-98 tion II.K.1 I&E Bulletins 99 II.K.2 Applies to B&W Only X
II.K.3 Bulletins & Orders T. F.
100 Recommendations b'
III.A.1-1 Upgrade Emergency Preparedness X
III.A.1-2 Upgrade Support Facilities 104 III.A.1-3 Thyroid Blocking Agent X
III.A.2 Improving Licensee Emergency X
Preparedness III.A.3 Improve NRC Emergency Preparedness X
Subitem III.A.3-4 applies i5' (See below)
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III.A.3-4 Nuclear Data Link 106 8
III.B Emergency Preparedness State X
and Local Governments
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TABLE 1
SUMMARY
OF NUREG-0660 ACTION PLAN ITEMS M.L. APPLICABILITY NUREG-0660 ITEM NO.
SUBJECT YES - PAGE NO. OF REMARKS NO
RESPONSE
III.C Public Information X
III.D.1-1 Radiation Source Control Outside 107 Containment III.D.1-2 Radioactive Gas Management 110 III.D.1-3 Ventilation System and 111 Filter Criteria
,1 02' III.D.1-4 Radwaste Design Features X
III.D.2 Public Radiation Protection X
Subitem III.D.2-3 applies Improvement (See below)
III.D.2-3 Liquid Pathways Dose Analysis 114 III.D.3-1 Worker Radiation Protection 115 III.D.3-2 Health Physic Improvements X
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- 1 fl III.D.3-3 In-Plant Radiation Monitoring 117 0
III.D.3-4 Control Room Habitability 120 e
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ITEM I.B.1-1:
OIKIANIZATIQi AND MANAGDENT IMPRO'! DENTS
,a Tasks I.B.1-1 and II.J.3 have as their collective objective to im-prove the plant operator's safety performance and ability to respond to accidents. Task I.B.1-1 is mainly concerned with the size, qualification and organization of the operating utility staffs, both onsite and offsite. Such considerations are beyond the scope of the manufacturing license application. Task II.J. 3 (and to a lesser extent Task I.B.1-1) seek to increase the familiarity of the opera-ting-utility personnel with the design of the plant through greater involvement in OA activities related to design and construction. Some of these activities require cooperation on the part of the plant designer and constructor.
We Floating Nuclear Plant will be of standardized design, with changes between successive units held to only the essential minimum.
It is therefore not possible for each customer to participate in the design review process to the same extent as in a " custom" plant.
Were are, however, numerous ways in which Offshore Power Systems can work with the plant owner to assure intimate familiarity with the plant prior to power operation. Offshore Power Systems will be prepared to offer these services, sme of which may fall outside the basic ENP purchase contract. Examples of such services include:
1.
Customer review of plant specifications and other design docments but with limited approval rights.
2.
Custmer hands-on participation in plant pre-operational testing activities.
3.
Customer participation in OA activity related to plant manufac-ture.
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4.
Classroan training by OPS personnel covering Floating Nuclear
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Plant Structures and Systems. General training in nuclear technology and in specific NSSS design topics is offered by Westinghouse, and is not duplicated by Offshore Power Systems.
After the first FNP becomes operational, subsequent purchasing utilities will have the unique opportunity to obtain field familiari-zation in a plant virtually identical to their own. Altinu3h OPS cannot make comitments for its customers, it is likely that a limited number of personnel from a prospective FNP owTier w;uld be welcome to observe and assist in the operation of another Floating Nuclear Plant.
Offshore Power Systems will be unique among nuclear plant designers and constructors in that its activities will be conducted under a license issued by the Nuclear Regulatory Commission. %is establishes a direct line of responsibility between offshore Power Systems and the NRC for plant design and construction. %is relationship has been recognized since the intiial filing of the manufacturing license application; therefore, the OPS design, CA and construction organi-zations have received intensive NRC review. These organizations are described in Chapters 13 and 17 of the Plant Design Report.
b)
'v 11 Revision 1
ITEM I.C.1:
OPERATING PROCEDURES
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We objectiJe of this recommendation is to improve the performance of reactor operators durirg transient and accident conditions. Offshore Power Systems is maintaining cognizance of the following work being performed by Westinghouse, throtgh the Westinghouse Owners' Group, which 1.s pertinent to the Floating Nuclear Plant.
(1) Short-term accident analysis and procedures revision: This item requires analysis, procedure guidelines, emergency procedures, and operator trainity related to mall break LOCAs.
a.
Analysis - Small break IOCA analyses have been performed, documented in WCAP-9600/9601, " Report on Small Break Accidents for Westinghouse NSSS Systems," and subnitted to the NRC (Bulletins and Orders Task Force) on June 29, 1979.
Eis report gesents a comprehensive study of Westinghouse system response to small break LOCAs for non-ice condenser
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containment plants.
We Bulletins and Orders Task Fbrce issued one se* of questions on WCAP-9600/9601 on August 13, 1979. %ese westions were responded to by Westinghouse in September 1979. Analyses have been performed and subnitted to the NRC (specifically in support of TVA and Duke) for small break LOCAs for UHI/ ice condenser containment plants in W.AP-9639. NCAP-9639 is applicable to the FNP.
b.
Procedure Guidelines - In light of the 'IMI accident, small break IOCA analyses, inadequate core cooling analyses, transient and accident analyses, discussions with NRC (and subsequent NRC reviews), and inputs from utilities, Westinghouse (in support of the owners' Group) has revised the four basic emergency procedure guidelines (E-0, E-1, E-2, and E-3). %e NRC has reviewed and approved E-0 and E-1; IRC review of E-2 and E-3 is continuing.
In addition, f3
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12 Revision 1
the abnormal operating procedure guidelines (A-0 through
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N,) have also been reviewed and revised in light of the
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1MI incident.
c.
Dnergency Procedures
'this requirement is a owner action to incorporate the gmeric Westirghouse emergency procedure guidelines and in plant specific emergency procedures.
d.-
Operator Trainirg
'Ihe plant owner will be required to train operators in the use of the latest plant-specific operatirg procedures.
(2) Inadequate Core Cooling: 'Ihis item requires analysis, procedure guidelines, emergency procedures, and operator trainire related to inadequate core cooling.
a.
Analysis - Westinghouse performed and subnitted WFUGH analyses of inadequate core coolire to the NRC for non-UHI plants. A further analysis of inadequate core cooling (utilizing the NOIRUMP code instead of the WFIASH code) is being performed for non-UHI plants and is schedu3 ed for ccznpletion durirg the sumer of 1980. Westirghouse is performing similar analyses for UHI/ ice condenser contain-ment plants. These analyses, which are scheduled for completion in the summer of 1980, will be applicable to the FNP.
b.
Procedure Guidelines - The response to Item (1) b., above, is generally applicable; however, it is noted that the Westinghouse generic procedure guidelines will undergo additional review and modification if appropriate when the results of the analyses of inadequate core cooling are available.
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c.
Emergency Procedures - Se response to item (1)c., above, is generally applicable; however, it is noted that if the Westirghouse generic procedure guidelines are modified as a result of new analyses, corresponding changes will be necessary to the owner's plant-specific procedures.
d.
Operator Training - The response to item (1)d., above,
. applies.
(3) Transients and Accidents: WCAP-9691, " Transient and Accident Analysis," responds to this item and is applicable to all Westinghouse PWRs (including ice condenser /UHI plants). %e information derived frcan the analysis in WCAP-5591 was used in reviewing and evduating emergency operating guideli as.
We results of this review indicated that present guid.ines are adequate.
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.i ITEM I.C.5:
FEEDBACK OF (PERATING EXPERIENCE t'~\\
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Each plant owner will be responsible to the NRC for the preparation and updating of plant operating procedures, offshore Power Systems will assist plant owners in discharging this responsibility by serving as a clearinghouse for important information derived from Floatirg Nuclear Plant in-service experience, design developnents and experience gained during testing. It is expected that NSSS-related information will be supplied to OPS by Westinghouse for transmittal to ENP owners. Die response to item I.E.4 deals with the question of how operating experience will be fed back into the design and construction processes.
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i ITEM I.D.1:
CCNIROL ROOM IESIGN REVIEW l
(A) i Offshore Power Systems proposes to consolidate this item with item I.D.4.
'Ihis reflects both the intent and flexibility to finalize control roca design in conformance with post-MI standards. Since the PNP control room is in a preliminary design status, there are no design comitments, other than overall size, which would be difficult to reverse 6. ring final design. Considerable flexibility remains in the important areas of type and location of information displays, control board layout and the use of computer data processing. Given this flexibility there is no need for an imediate design review aimed at identifying problem areas prior to the imenent comencement of construction activities. Rather, it is more efficient, yet equally effective, for Offshore Power Systems to await the developnent by both NRC and industry of post-MI control room standards. Then standards are expected to be issued in ample time to play a central role in control rocm final design. In the meantime Offshore Power Systems will remain abreast of developnents in the field of control h
room technology, in part through continued participation in the Westinghouse Owner's Group.
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16 Revision 1
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l ITEM I.D.2:
PIANT SAFEIY PARAMETER DISPIAY CCESOLE f%
i OPS proposes to use the following criteria for future implementation l
of " Plant Safety Parameter" Displays:
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'1.
Input parameters will be selected and the parameter data processed in such a way as to yield a concise, reliable indi-cation that addresses the following safety concerns: core cooling capability, reactor coolant saturation, reactivity 1
excursion, loss of primary coolant inventory, loss of tempera-ture and pressure control, and radioactive releases.
l 2.
These input parameters will be sufficient to monitor all design basis plant transients.
3.
Diversity, redundancy, and error-checking will be utilized to assure reliable safety status indication.
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4.
We display devices will be prominently located on the Unit Control Console described in Section 2.0 of Attachment I.D.4-1.
5.
We design provisions for safety-related display instrmentation described in Section 7.5 of the PDR will be utilized so as to complement the Plant Safety Parameter Display described here.
Durirg final FNP design the Control Rocm and control board designs will be developed in consonance with new requirements which may result from Industry or NRC recomendations.
As explained in the response to item I.D.4, TS does not expect problems in implementing future requirements for Plant Safety Parameter Displays.
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Revision 1
ITEM I.D.3:
SYSTDI STARIS MCNIR) RING
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Since as early as 1974, Offshore Power Systems has been consnitted to total conformance. with Regulatory Guide 1.47 (Plant Design Report Section 7.1.7). The Floating Nuclear Plant presently includes signif-icant provisions for status monitoring; these are outlined in the following paragraphs. Offshore Power Systems will remain abreast of continuing developnents in this area, includirg those by INPO, and particularly those affecting Regulatory Guide 1.47. Should new requirements arise, they can be addressed during the final design phase.
1 l
Assurance of proper operation and/or positioning of safety-related equipnent (including equipnent in ergineered safety features support-ing systems) during all operating activities is provided by:
1)
Main Control Board (MCB) Display Features:
include position /
status indicating lights, position / status disagreement indi-o
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cation, availability indication, and system level bypass indication. These features meet or exceed Regulatory Guide 1.47.
Scrae additional criteria are stated in Section 7.5.1 of the PDR.
These features are as follows:
Position / Status Indicating Lights (Backlit PusDutton)
(PIL)
Backlit red (open) and green (closed) pashbuttons indicate actual valve position frcm limit switches on the valve. The pushbutton is part of the PCB module for that valve.
Backlit red (on) and green (off) pushbuttons indicate breaker or contactor status from appropriate auxiliary b
LJ 18 Revision 1
contacts. Se pushbutton is part of the KB module for that component (ptsnp, fan, etc.)
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2ese position / status signals are also inputs (through isolation devices) to the Plant Computer Systems.
Valve Position Indicating Lights (Lights Only) (PI L*)
This valve position signal is also an inpat to the Plant Computer Systems.
Position / Status Disagreement Light / Alarm (Backlit Push-button) (PDL)
A backlit alarm indication / acknowledgement pushbutton (normally extinguished) flashes in conjunction with an audible alarm if the equipnent fails to achieve the last position or state comanded. In addition, the comanded position / status indicating light flashes. Eis backlit
'v pushbutton is part of the MCB module for that equipnent.
Both of these flashing lights are acknowledged by this pushbutton, chargirg the alarm indication pushbutton from flasrdng to steady, and the comanded PIL from flashing to exting.aished. %e steady alarm indication light is not extinguished until the comanded and the actual equipnent state are in agreement.
- indicates " Lights only", see Table I.D.3-1 n
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19 Revision 1
Availability Light / Alarm (Same Backlit Pushbutton as PDL
/
above) (AVL)
+
If the equipment is removed from service (i.e.,
if motive power is unavailable or locked out) either deliberately or due to failure, the backlit alarm indication / acknowledge-ment pushbutton (the same device actuated by the PDL) flashes in conjunction with an audible alarm.
For equipnent removed fra service, this alarm signal is also an input (through an isolation device) to the Plant Computer System. The Plant Computer System flashes a system level display (BYP) on the K:B indicating that the appro-priate system ESF train is bypassed.
System Level Bypass Indication (BYP)
An engraved backlit window, prominently displayed to the Q
operator, is provided for each division of each major b
Safety Subsystem (e.g., SIS, RHR).
'Ihis window flashes whenever any of the following con-ditions (within the scope of the window) indicates a bypass of a protective action:
a)
Motive p>wer unavailable to an ESF actuation device (for example, an MOV, power unavailable to the
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reversing contactor), due to deliberate bypass or circuit failure. This condition is derived from " AVL" signal. (AVL /BYP) b)
Valve positioned so as to create a bypass of a protective action. This condition is derived from actual valve position. (PIL/BYP)
\\.J 20 Revision 1 i
c)
Window activated manually by operator f rom MCB, responding to information received through adminis-trative control. (ADM/BYP)
If a redundant division of any subsystem were concurrently placed in a bypass mode (due to any of the above inputs),
the second division window would flash and an audible alarm would occur.
Acknowledgement of the first division level bypass causes the first window to change from flashing to steady, until the bypass is cleared. Acknowledgement of the second (concurrent) division level bypass silences the audible alarm, but leaves the second window flashing until one of the bypass conditions is cleared.
The plant comp 2ter systems perform the combination and sequence logic that is required to control the system level O) bypass indication windows. The position / status inputs to tv the computer that are derived f rom Class IE control circuits are isolated in accordance with Regulatory Guide 1.75.
l 7he bypass indication system meets or exceeds the require-ments of Regulatory Guide 1.47. Additional design criteria
)
for the bypass indication system are provided in Section 7.5.1 of the PDR.
System Level Monitor Indication (MCN)
An engraved, backlit window, prcxninently displayed to the operator, -is provided for each division of each major safety subsystem (e.g., SIS, CSS). This window flashes, in conjunction with its corresponding PDL light (s), whenever any equipnent (within the scope of the window) has failed
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to respond to an ESF signal.
21 Revision 1
2),
Control Circuit Design Features: In addition to these display
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's features, circuit design features are provided to assure proper alignment of equipnent. These features include assignment of control priorities to ESF signals and selection of failure modes. These control features are described below.
Control Priority Assignment (CP)
While' the equipnent is in service (i.e., while motive power is available to it), its control priorities are assigned such that ESF signals will always override non-ESF signals 1
(with the exception of electrical and mechanical ci rcuit protection features which must override ESF signals in order to prevent component damage).
i Failure Mode of Actuation Device (FM)
Removal of an air operated or solenoid operated valve from
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service (i. e., removing motive pwer) will cause the valve to move to the safe position.
Administrative Control Input (Manual) to Bypass Indicatig Systen (ADM)
'Ihe system level bypass indication (BYP) can be manually input by the operator through administrative control.
Computer software supplements plant administrative controls by trackirg these manual inputs (together with non-manual inputs), determining the system level effects, and pro-vidirg appropriate displays.
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22 Revision 1
Table I.D.3-1 illustrates the specific application of these' design
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features to the generic types of FNP equipment that could be in-
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correctly operated. 'Ihe table indicates which of the ENP control and display design features provide direct defense against:
a)
The eftects of mispcsitioned circuit breakers or contactors b)
The effects of mispositioned valves, g c)
Undetected mispositioning of equipnent for various con-ditions of plant operation and for various types of equipnent.
Considered in the table are:
a)
The nature of the safety system bypass (deliberate vs.
inadvertant)
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The plant operating mode (periodic test, maintenance, etc.)
% -l c)
The ergineered safety features systems mode (standby vs.
active) d)
The type of safety equipment (circuit breaker, motor operated valve, hand operated valve, etc.)
Table I.D.3-1 doer not address any ENP design features that are not relevant to safety consequences, nor is credit taken for other types of design features (e.g., process alarms) that in some cases would further enhance safety.
Additional information is contained in the response to ite.n I.D.4.
4 n-23 Revision 1
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TABIE I.D.3-1, Sheet 2 (3^
. L.!
NCTTE 1.
Valves are locked in safe position, and are mder adminis-trative control.
NOTE 2.
" Operator error" includes failure to recognize a valve that is.lef t improperly positioned (for power operation) follcwing startup.
NCTTE 3.
Safety-related hand operated process valves that have the capability of significantly degradirg a protecuve action if left mispositioned are subject to the following cri-teria:
a)
If rormally operated more frequently than once per year with the plant at power, shall be locked in the safe position under administrative control. In addi-tion, remote position indication shall be provided.
v b)
If normally operated at startup, shutdown and/or refuelirg, shall have the provisions of paragraph a),
c)
If only operated _ for ron-routine maintenance or repair (e.g., to isolate a ptunp or heat excharger for repair) with the plant at power, shall be locked in the safe position under administrative control.
NOTE 4.
%is table includes only those control and display features that provide direct defense against these conditions, recognizing that others of these features might be provided for a particular component, but would be less relevant.
. NOTE 5.
Where more than one design feature provides defense, the most prominent one is listed first.
G 25 Revision 1
' ITEM I.D.4:
CGf!BOL ROCM DESIGN STANDARD f~T i
i
'~
IEEE-566 and IEEE-567 are currently in draft form, and do not represent an Industry /Nhc consensus. However, OPS has established design bases, criteria, and functional specifications for the FNP Control Room which meet or exceed draft 1C of IEEE-566 and draft 3 of IEEE-567. A sunnary of the relevant design bases, criteria, func-tional specifications / descriptions for the FNP Control Room is presented in Attachment I.D.4-1. Further description of safety system status monitorirg (includirg description of the light and alarm sequences for control board modules) is presented in the response to item I.D.3.
We FNP Control Room design bases, criteria and general functional requirements are presently adequate for meeting all of the concerns i
identified after the 'IMI accident.
OPS does not expect problems in ir.plementirg any future requirements, because:
1.
We status of detailed design is such thet schedule constraints are not expected.
2.
%e design concept is flexible (modular construction and modular display software will be used).
3.
OPS current practice for control room design evolution includes a fonnal, orgoing multi-discipline review process (including tne use of a full-scale mockup). Any significant changes will be implenented with that process.
A'w,)
-26 Revision 1
ATTACHMENT I.D.4-1:
EEERPFS PROM 'lHE PNP CONIROL ROOM fg SYSTEM SPECIFICATIONS
()
1.0
' Control Room Design Bases and Criteri.
1.1 Equipnent Design Basis
1.1.1 Classification
he control room panels, their foundations and supporting structures are classified as follows:
Panels that have Class lE equipnent or circuits, are classified as Class lE, Seismic Category I.
Panels that house no Class 1E equipnent or circuits are classified as Class NIE Seismic Category II.
The classification,1E or NIE, of the individual display and control devices and their associated wirire are as designated in their respective schematic connection diagrams. We FNP Instrtraent List will provide a composite listing of all the devices mounted on the panels and include, amorg other information, the classification.
1.1.2 Environment
O i b he main control panels and all equipnent mounted therein shall be designed to operate under the following atmo-spheric conditions:
Temperature 5 C to 50 C (40-122 F)
Pressure 1 Atra Humidity 10 - 90% RH NOTE: We Control Building Air Conditioning System will maintain the control room at 75+ 2 F and a relative humidity of 40 - 70% under all postulated plant conditions, l
All equipment associated with the main control room panels are located within the control module and will receive less than 2 RAD over the design life of the ENP.
1.1.3 Criteria
The basic criteria for the design, fabrication and testing of the main control panels are derived frcra the application of IEEE 279 and IEEE 384 to the class 1E equipnent and circuits contained therein.
- . p) 5%./
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27 Revision 1
,.. ~.
The main control panels must provide sufficient support and
(,]
physical protection to its Class lE equipent and circuits
(/
to enable them to perform their essential functional i
requirements before, during, and after motion conditions, up to and including design basis motion conditions.
he panels shall be so designed that, at the frequencies and accelerations of the floor resulting fra design basis j
motions, they do not amplify the forces beyond the level at which the equipent contained therein is qualified to j
function properly.
To meet this requirement, the panel shall be designed with sufficient rigidity so that no natural frequencies or resonances can exist at a frequency less than (later) Hz.
Welded stiffeners, diagonal braces
)
and thick plate skin shall be used singly or in any combination that will satisfy this rigidity requirement.
The panel design shall be seismically qualified in accordance with the requirements of IEEE 344.
The panel design shall include provisions for securely j
mountirg the board to its supports.
A dynamically equiva-lent support stiall be used in the seismic testing of the j
panel.
Plug-in or slide mounted equipent shall be provided with mechanical constraints, if needed to maintain positional integrity.
All equipment within, attached to, or adjacent to, the panel shall be mounted such that the structural failure of this equipment cannot damage Class lE equi p ent or cir-cuits,
The internal structural design of the main control panel shall provide for the physical separation of redundant Clc.ss lE circuits and equipment, as required by IEEE 384 so that no single credible event can prevent the proper functioning of any Class lE system. %e required separation shall be achieved by an adequate air space or a fire-retardant barrier between redundant Class lE circuits and equipent. The circuit wiring shall be supported in a manner that will assure maintenance of the air space throughout the design life of the panel. %e required separation shall be maintained from the point of entry of the circuit into the panel to the final termination on the surface mounted devices. Non-Class lE circuits and equipnent shall likewise be separated from all Class 1E equipment and circuits.
Inherent flame-retardant characteristics and properties shall be a major ' consideration in the selection of materials for use in the main control room panels.
he
- p/
structural framework and surfaces of the panel shall be
(,
fabricated of steel or aluninun stock.
Any nonmetallic l
28 Revision 1
cmponents and devices should be manufactured from self-p/
extirguishing material as defined by ASM Std. D635-1972.
s I-a in ts or other. applied surface preparations should contribute only nominally to the total combustible potential of materials or components in or on the panel.
Considerction should be given to the release of toxic or corrosive gases and dense smoke and their effects upon personnel and equipnent.
1.1.4 Electrical Design:
In accordance with IEEE-279, components and modules shall be of a quality that is consistent with minimtzn maintenance requirements ' and low failure rates *.
Qt 21ity levels shall be acnieved throtsh the specification of requirements known to promote high quality.
All control and instrtznent wiring shall have sufficient mechanical strergth, current capacity, thermal rating cad insulation characteristics to meet the circuit and installation requirements established by plant design.
Wire and cable insulation shall be flame retardant with self-extinguishirg nonpropagatino characteristics.
Con-sideration shall be given to the stential release of toxic or corrosive gaces and dense snoke and their possible p
effects upon personnel and equipnent.
All wire and cable V
installed within the main control panels must be capable of meeting the flame test requirements of IEEE 383.
1.2 Operator Interface Criteria
1.2.1 General
The primary criterion for the ENP operator interface is that it shall provide to the operator the information and control facilities that he needs to safely and efficiently operate the plant under normal and upset conditions and present them in such a manner as to enhance his under-standirs of the plant status and reduce the probability of an operational error.
- Mean Time Between Failures:
20,000 hr.
Mean Time To Repair: 30 min.
).
el 29 Revision 1 l
1.2.2 Information and Control Requirements:
-()
he determination of which information and controls are to be provided in the control rom begins with the responsible OPS process system engineer, in close cooperation with the corresponding control system erg ineer. This process is formalized by evolution of the following controlled docments:
1.
Process System Specifications 2.
Instrtment Block Diagrams 3.
Control Logic Diagrams 4.
Schematic Connection Diagrams troughout this design process, each system is analyzed
- frce an operatirg point of view as well as from a design point of view. 'Ihe control systems engineering group is responsible for the systems integration as well as for the application of control and display hardware.
In the process system specifications, the information and coatrol requirements are defined in functional terms. %ese functional requirements are further defined in the Instru-ment Block Diagrams and Control I.ogic Diagrams and then n}
(
converted to hardware requirements (i.e.
indicators, k
lights, switches, etc.) in Schematic Connection Diagrams.
The I&C hardware requirements for all the WP systems are consolidated into a single doctment - the WP Instrument Lis:. This doctment is a computerized list of all the I&C hardware provided on the WP and includes sufficent information to completely describe each item.
Included in the bank of information is the mounting location for each item.
1.2.3 Arrangement Requirements:
he configuration and arrangement of the cont.rol center panels and the placement of indication and control devices on the panels shall be based on the following:
- he WP control center shall be designed to enable a single operator to safely control the plant under all operating conditions.
Provisions shall also be made for accorro-dating additional operating personnel during periods of high activity when it is desirable to relieve the burden of the lead operator.
I' V
30 Revision 1
In multi-operator situations, the following organizations
--/sS will be assumed:
'O he lead operator will be singularly responsible for the safe conduct of control center operations. %e second oper-ator will be assigned a subordinate role and shall take action only at the direction of the lead operator or in accordance with written procedures authorizing specific independent action. Se subordinate operator will make reports to the lead operator (1) prior to the initiation of independent action (2) when difficulty is encountered in the performance of an assigned or independent action (3) periodically on the status of lorg term assignments and (4) whenever an abnormal situation is noticed.
%e lead operator will likewise be responsible for keeping the assistant' operator (s) informed of the plant status.
Functional Areas -
The - control center shall be subdivided into distinct cperatirg areas.
%e functional requirements defined for each area determine the major criteria for the allocation of display and control devices within the control center.
%ese operating areas are designed to provide for a separ-ation of safety-related systs and auxiliary and supporting devices from those required by the operators to monitor and control the plant under normal conditions.
This method of device allocation affords a reduction in nmber of displays that the operator must observe under normal conditions and consequently, a reduction in the probability of misinter-pretation and erroneous action.
We operating areas to be incorporated in the FNP control center design are as follows:
Normal Operations Area The normal operations area is the primary control location for the Floatirg Nuclear Plant under hot non-upset condi-tions. We display and control devices located in this area will be the minimum required for the operator to perform the following:
1.
to assess the status of the plant and its systems at any time 2.
to be alerted to abnormal situations and changes in plant status 3.-
to maintain the plant in a safe hot shutdown condition 4.
to maneuver the plant from a hot shutdown condition to
' p full power operation C/
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31 Revision 1
5.
to manually initiate safety systems (on a system (V '_
level)
)
Safety System Operating Area his area provides the facilities required by the operation to:
1.
quickly assess that safety systems are performing their required safety functions 2.
monitor long term course of the accident 3.
determine when conditions exist that require specific manual actions, to take such action and monitor the results 4.
perform safety system functional testing 5.
determine the availability status of protection and safety systems at any time during panel operation Infrequent Operations Area mis area is allocated to display and control devices needed to perform auxiliary or supportire functions that are required infrequently.
(An example would be the (n).
devices that are used only during heatup, cooldown, cold
'v shutdown and refueling.)
Historical Records Area mis area is allocated to the devices which are required to provide hard copies of computer stored data.
Area Arrangement The arrangement of the operating areas within the control center will be consistent with the followirg criteria:
1.
Se normal operations area will be centrally located and provide the operator with surveillance and access capability to other operating areas.
2.
The safety systems operating and the infrequent operations area shall be directly accessible and visible fra the normal operations area and not be in a separate enclosure.
3.
We historical records area shall be located apart frm the operating areas.
(~3 4..
We supervisor's office shall be located as to give 4j him a visual connand of all control center activities.
32 Revision 1
w 1.2.4 Htenan Engineering Requirements:
~
he following human engineering considerations - shall be factored into the design of the FNP Control Center.
Anth: 7pometric Considerations -
he control panels will be designed to permit 5 to 95 percentile (in height) operators to read or reach all indicators and ' controls frczn a standing. position in front of the auxiliary panels and from a seated or standing position in front of the consoles.
The 5th percentile operator is 5'4" tall. and the 95th percentile operator is 6'4" tall.
Task Analysis -
The assignment of controls and displays to the functional areas ard their placement within the areas will be based on the operators need for the devices in the performance of his assigned tasks.
+
Each control and display device will be analyzed to determine:
1.
the operating modes during which the operator needs
['w])
the device, 2.
how often the operator uses the device when it is
- needed, 3.
in the case of controls, how fast does the plant respond to a control manipulation, and 4.
in the event of a malfunction, how fast must the operator take corrective action.
Other Considerations -
During the last few years a number :of human engineering reviews of existirg nuclear power plant control panels have been conducted by the Electric Power Research Institute, W Research Laboratories and others. 'Ihe.results of these studies shall be used to develop a checklist for the. ENP design to ensure that the typical deficiencies noted in Table I.D.4-1 are avoided.
^ (3, Q/
33 Revision 1
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- 1. 2.5 '
Panel Mock-Up:
i
.A full-scale control center mock-up will be constructed and used to evaluate the hunan ergineering aspects of the FNP design. 1he evaluation will include " walk-throughs" of the FNP operating procedures to ensure that no operational problems are overlooked.
a i
4 4
4 4
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O 34 Revision 1 w
e.
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2.0 FUNCTIONAL DESCRIPTION
,m
)
Figure I.D.4-1 show the preliminary ENP control room layout and the locations of the various functional areas.
We correlation of equipment with these areas is as follows:
(1)
NORMAL OPERATIONS ARFA -
- UNIT CONEOL CONSOLE (2)
SAFETY SYSTD4S OPERATING AREA - SAFETY CENTER (3)
INEREQUENT OPERATICN AREA
- SIIONDARY CONTROL CENTER (4)
HIS'lI)RICAL RECORE AREA
- COMIUTER OPERATORS CONSOLE 2.1 Unit Control Console:
The Unit (bntrol Console (UCC), shown in figure I.D.4-2, is a compact modularized console from which all normal plant operation is conducted.
It includes all the displays and controls necessary to brirg the unit frcrn hot shutdown to rated pwer (and back to hot shutdown) and fo r controlling and monitoring load changing operations. The UCC design permits one-man operation while providing space for two.
%e UCC provides three computer generated visual displays (CRT's).
These displays, together with their associated keyboards, provide the operator with all the information be needs to assess the status of the plant ard its systens at any time. We center CRT contains a process overview in @ich all key parameters are continuously updated
- in a single display.
he left hand CRT contains an alam display
[sv]
showing the status of all pints in alarm.
he right hand CRT is used to display parameters for any selected system in detail.
We remainder of the UCC is arranged in stations that are dedicated to those portions of the followirg systems used during normal operation:
1.
Ibd Control System and Rod Position System 2.
Nuclear Instrunentation System 3.
Chemical and Volune Control System 5.
Feedwater and Condensate Systems and Auxiliary Feedwater 6.
Main Steam System 7.
Wrbine Generator System 8.
. Generator Circuit Breakers and Synchronizing The UCC will also include safety system manual actuation controls and any permissives and blocks required for normal operation.
(3 t
35 Revision 1 l
2.2 Auxiliary Panels (p,)
2.2.1 Safety Center:
he Safety Center, shown in figure I.D.4-3, provides for the monitoring, control and testing of the ENP protection and engineered safety features systems. The panel is arranged in stations that present a logical flow of information to the operator. Se left most station provides the displays and controls associated with the Reactor Protection System (SSPS) and those displays and recorders required for Post-Accident Monitoring. %e stations located immediately to the right contain the component level displays and controls for the ESF systems. These include:
1.
Upper Head Injection 2.
Safety Injection 3.
Essential Service Water 6.
Essential Raw Water Also included in this area are the CRT and keyboard and system level binary status displays required by the
- v)
Protection and Engineered Safety Feature Availability and Test System.
To the right of the ESF stations will be located the indicators and controls associated with lE support systems including:
1.
Air Conditioning Systems 2.
Hydrogen Recombiner Systems 3.
Containment Isolation Valves (those not used during normal control) 4.
Ice Condenser Systems 2.2.2 Secondary Control Center:
he Secondary Control Center, shown in figure I.D.4-3, contains all of the reqaired control and indicators that are not located on the UCC or the Safety Center.
We arrargement of the systems generally follows the order in which they are used in bringing the unit to power opera-tion.
The controls and indicators located here are primarily those that are used only during refueling,
. f7 heatup, cooldown and maintainirg cold shutdown.
tu!
l 36 Revision 1 l
2.2.3 Component Arrangement:
j3V The controls and indicators required for the operation of each individual system will be integrated into a comon work station. Se arrangement of components within the work station will follow the placement of the controlled compon-ents in the actual process system. For complicated systems or those used infrequently, a graphic display will be pro-vided above the work station. A typical example of this arrangment is shown in figure I.D.4-4.
2.2.4 Annunciators
Annunciators will be located along the upper sloping por-tion of the safety center panels.
These will bn conven-tional hardwired alarm points and will be used as backups in the unlikely event that the computer generated alarm display is inoperative.
%is use of annunciators will be restricted to alannirg only those fault conditions that could affect the ability to reach and maintain a safe shut-down condition or those required to meet regulatory requirements.
2.3 Historical Records Area The Historical Records Area, shown in figure I.D.4-1, is centered at the computer operators console.
At this console an operator will be (d
able to obtain a hard copy of CRT displays, computer calculations,
)
test results, etc.
L. !
37 Revision 1
3.0
. EQUIPMENT DESCRIPTION ql
\\
y) 3.1 Unit Control Console he unit control console (UCC). will be a free standing sit-stand console as defined by IEEE-27. Figure I.D.4-2 shows the general size and diape of the console.
%e UCC will be provided with front and rear removable panels for access to internally mounted equipnent and cable terminations Equipment will be mounted within the panel with a view towards maximtsn accessibility for testing and maintenance.
We console design will provide for the entry of cables through bottom access holes centered below vertical wire-ways housing terminal blocks and connectors. Five sets of vertical wire-ways will be consistent with the requirements of IEEE-384 for cable spread rooms.
Horizontal raceways will be provided for supporting cable along the lergth of the console.
Five raceways will be provided, one for each set of vertical wire-ways.
Cable access from the wire-ways to the raceways will be provided only between those of the same division.
We spacing of the raceways will be consistent with the requirement of IEEE-384 for panel internals.
Cable runs from the raceways to the panel mounted equipnent will be
(
by the most direct route consistent with the following division V
separation requirements.
'l.
From the raceways to a distance of one foot from the panel surface, cables of different divisions will be maintained at least six inches _ apart.
2.
Within one foot of the panel surface the division separation may be reduced to one and one-half inch when, due to equipnent proximity, six inches cannot be maintained.
The reduced separation requirement at the panel surface is provided to allow for cases when, due to operational considerations, it is desirable to mount equipnent belorging to different divisions at adjacent locations. We justification for the one and one-half inch separation will be provided throtgh analysis and testing. We analy-sis will show that no single credible event, with the exception of an
' internally generated fire, could prevent the proper functioning of a lE system. %e testing will demonstrate that, with - the low enargy circuits used in the panel, an internally generated fire that could affect redundant divisions, is incredible.
. -3. 2 Back Panels:
The back panels will be free-standing duplex benchboards as defined by IEEE-27. he panels will be provided with rear access door a and p).
removable front panels.
(
38 Revision 1
(
.p Provisions for cable entry and routing are similar to those described
.,v)-
(
in section 3.1.
3.3 Instrtsnent Modules:
%e majority of the discrete display and control functions required on the panels will be accomplished usirg a modular system of instru-mentation.
%e modules are all of the same height and their widths are multiples of a fixed modular dimension.
We basic modules in-clude: an indicator module; and auto / manual module; a pishbutton module; and a recorder module.
A typical indication module, shown in' figure I.D.4-5, has two vertical displays. Each of the displays will provide one percent reading accuracy and will be scaled in engineering units.
The auto /maneal module, shown in figure I.D.4-6, contains four backlighted pushbuttons and an edgewise indication. We auto / manual module together with an indicator module will be used to perform auto-manual control functions.
In this application, one of the displays will be used for the me. sured variable and the other for the set point.
Se pushbutton control module, shown in figure I.D.4-7, is the primary binary control and indicating means.
We module can contain up to six backlighted pushbutton operators and each button can be
/]
split to display two messages.
C A recorder module will-be used whenever a hard copy record of a process variable is required ard the plant computer cannot be used to provide it. We recorder module will be four module widths wide and will be available in 1, 2 or 3 pen configurations.
All of the modules will be removable from the front of the panel and further, can be removed with the circuit active without affecting the state of the controlled component or parameter.
-h g
LJ 39 Revision 1 l
TABLE I.D.4-1 TYPICAL HUMAN DKiINEERING DEFICIENCIES Readirg Indications Recomended viewing distance exceeded Meter design causes glare and improper viewirg argle Control design obscures position setting Reachirg Contrcls Functional reach exceeied 4
Workirg posture leads to accidental activation Activating Controls Inconsistent direction of movement relationships between control and associated display Violation of operator expectation of direction of movement of control Nomenclature that violates operator expectation Use of same nmenclature for different functions Different rx,menclature used for functionally identical controls Interpreting Codirg Use of the same color for more than one function Color - function associations that violate operator expectation Functionally identical controls color-coded differently Inconsistent use of illumination codirg Interpretire Alarms No differenti tion of the severity of alarms Nuisance alarms Ob 40 Revision 1 i
?>
TABLE I.D.4-1 (CONT'D) rh b
Locating Ca ponents No delineation between major control systems Side by side location of functionally unrelated controls that are identical in appearance Incompatible arrangements of associated displays and controls i
Illogical arrargment of related controls Inconsistent location of the same type of control Performing Sequential or Simultaneous Operations Spatial separation of controls that must be used together I
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- 4. HISTORICAL RECORDS AREA FIGURE I.D.
4-1 FNP CONTROL ROOM ARRANGEMENT
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OUTPUT INDICATOR O
AUTO PussaufrON #ERATORS MAN REMOVABLE COLOR CODED BACKGROUND O
FIGURE I.D.4-6 TYPICAL AUT0/ MANUAL MODULE REMOVABLE GRAPHICS PUSH 90TTON OPERATORS i
FULL OR SPLIT SCREEN ENCRAVED DISPLAY i
REMOVABLE COLOR CODED BACKChDUND
)
l FIGURE I.D.4-7 TYPICAL PUSHBUTTON CONTROL MODULE O
u - - - -
ITEM I.E.4:
COORDINATIW OF PROGRAMS 'IO ANALY'E OPERATING EXPERIENCE (m
r R)
The response to item I.C.5 considers the OPS - custmer interface. I,t is through this interface that OPS will assist each plant owner in the broad area of plant operation. 'Ihe balance of this response to item I.E.4 considers the collection of operating information and its application in the design anc* construction processes.
Offshore Power Systems will receive information from several poten-tial sources, includirg the followirg:
(1) NRC, through I&E Bulletins, etc.
(4) Custmers (5) In-house Preoperational Testing (6) Westinghouse At the present time, industry groups are formulating plans to co-ordinate the collection and evaluation of plant operating experience.
One system, called SEEIN (Significant Event Evaluation and Infor:na-tion Network), uses Licensee Event Reports and Monthly Outage Report, both required by NRC, as primary inpat data.
Offshore Power Systems will establish a focal point for receipt of operating information and subsequent distribution within the organi-zation. This information will be routed to the cognizant design and/or manufacturing disciplines. Design and/or procedural changes will be made as necessary, but, in any event, a written record of the disposition of each item (or group of' similar items) will be main-tained.
/'~T U
49 Revision 1
+ +- --
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M ITEM I.F.1:
EXPANDED QA LIST 7
t
)
LJ The existing body of EC requirements assures that QA lists include a comprehensive set of structures systes and components important to safety. In addition to NRC requiements, Offshore Power Systems utilizes an internal classification system which adds many struc-tures, systems and components to the Quality Assurance list which might not otherwise receive more than standard cocraercial Quality Assurance measures. %e application of both NRC and OPS Quality Assurance requirements in the FNP are outlined below.
We general design criteria for nuclear power plants are contained in 10 cm 50, Appendix A. These criteria provide a broad definition of plant structures, systems and components important to safety. NRC Quality Assurance regulations are contained in 10 CFR 50, Appendix B.
Through regulatory guides the Appendix B Qualit.,r Assurance require-ments are imposed in varying degree, baseo on the daracteristics of the particular structure, system or component concerned. We Regula-(mv) tory Guides discussed below detail the degree of Quality Assurance required for--virtually all of the structure, systems and components which are the subject of General Design Criteria.
Regulatory Guide 1.26 identifies the nuclear plant fluid systems o
which fall into quality classifications A, B, C and D. Offshore s
Power Systems complies with this Regulatory Guide; however, industry safety classifications (1, 2, 3 and Non-Nuclear Safety or NNS) are used in place of quality groups A, B,
C and D.
Offshore Power Systems procedures require Appendix B Quality Assurance measures for all systems and components classified as Safety Class.1, Safety Class 2, Safety Class 3 or NNS.
o Regulatory Guide 1.29 requires that the Quality Assurance Przram of 10 cm 50, Appendix B be applied to each of the y
structures, systems and. components listed in Regulatory Posi-tions 1, 2 and 3. ' The Quality. Assurance measures invoked by
?
tj
~
50-Revision 1
offshore Power Systes.for Floating Nuclear Plant str uctures,
[/
)
systems and components complies with Regulatory Guide 1.29.
G o
Regulatory Guide 1.120 establishes the QA requirements for the Fire Protection System. %ese requirements are unchanged from those of Branch Technical Position APCSB 9.5-1 (Appendix A) which were committed to in offshore Power Systems Report RP06A30, " Floating Nuclear Plant Fire Protection Evaluation",
September, 1977.
o Regulatory Guide 1.143 supplements Regulation Guides 1.26 and 1.29 for Radwaste Systems. Regulatory Position 6 of this guide details an. acceptable Qual _;y Assurance Program for Radwaste Systems. Offshore Power Systems will meet or exceed these requirements in future design and manufacturing activitier.
Offshore Power Systens ergineerirg procedures require the responsible engineers to classify each Floating Nuclear Plant Structure and C)
System usire a pre-defined set of classifications. We set of offshore Power Systems classifications includes several classifica-tions in addition to those defined in Regulatory Guides 1.26 and 1.29. Offshore twer Systems Quality and Reliability procedures establish three quality levels and provide the c, elation between the various engineering classifications and quality levels. Quality Level 1 invokes appropriate portions of full Quality Assurance Program of 10 CFR 50, Appendix B.
Quality Iavel 2 invokes (as a minimum) requirements for procurenent doctraent t. trol, control of non-conforming items and Quality Assurance records. Quality Level 3 requires no Quality Assurance measures beyond standard comercial practice. The following summary indicates the existing Quality
. Assurance Level of those Floatiry3 Nuclear Plant systems which are of particular interest in light of the accident at Three Mile Island. As this sumary indicates, the Floatirg Nuclear Plant Quality Assurance list is very extensive and already includes many of the systems and components which are~ beirg considered by NRC as possible additions to
/^
the Quality Assurance list.
(.)\\
51 Revision 1
SYSTEM QUALITY LEVEL
'(U; All structures, systems and 1
caponents listed in Regulatory Guides 1.26 and 1.29 Main beam, including steam dunp 1 (Note 1)
Main Condansers 2 - tubes 1 - shell
~ Circulatirg Water 2
Condensate Polishing 1 (Note 1)
Condensate - Feedwater 1 (Note 1)
Instrument Air 2
Dnergency Instrument Air 1
Containment Post-Accident Sampling 1
Nuclear Plant Samplirg System 1
Steam Generator Blowdown 1 (Note 1)
- Based on the foregoirg, it is reasonable to expect that the imple-mentation of future additional Quality Assurance requirements will not only ' be feasible but have a minimal impact on the Floating l
Nuclear Plant.
J (1) Quality level 1 applies to main components and flowpaths. Iasser levels may be applied elsewhere in the ' system.
i O.
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7 52' Revision 1 i-m
,~,
y
ITEM I.F-2: -
DETAIIED QUALITY ASSURANCE CRITERIA t
r While it is correct tF *. 10CER50, Appendix B is general in its requirements, other documents such as those below have served to provide considerably greater guidance:
NRC Gray Book (WASH-1283), " Guidance on Quality Assurance Requirements During Design and Procurenent Phase of Nuclear Power Plants."
NRC Green Book ( ASH-1309), " Guidance on Quality Assuranca Requirements Durirg the Construction Phase of Nuclear Power Plants."
Both of these documents were addressed in detail during the review of the Manufacturing License application. Offshore Power Systems comments on the eleven areas identified for possible developnent of t
additional QA/QC guidance are presented in the followire paragraphs.
O As a general cocunent, it should be recognized that FNP design and D
construction is the sole purpose of the OPS organization. Therefore, the entire QA program deals with these activities including the CPS interface with both tht: NRC ar.3 the plant owner.
[1]
Independence of QA/QC Function This principal is recognized in the - OPS QVQC plan. 'Ihe third party concept is a new suggestion which can be implemented if a genuine benefit is likely to result. As presently envisioned, the QA/QC activities (at -least durirg FNP construction) will be carried out in varying degrees by the following parties:
(1) Offshore Power Systems (2) Customer (Resident Personnel)
(3) NRC (Resident Personnel)
(4) ASME (Likely Resident Personnel)
%/
53 Revision 1
(5) American Bureau of Shipping (v)
(6)
U.S. Coast Guard
'Ihis great diversity is self-assuring of OA/QC independence.
gj Procedure Approval Quality-related procedures related to design, construction and installation already require OA approval.
[3] OA Personnel Involvement OPS procedures already involve OA personnel in design and manufacturirg activities important to safety.
[4] OA Criteria for Equipment Classes OPS Ergineerirg and OA procedures now provide for assigning CA requirements on the basis of logical equipnent groupings, such as: structural, electrical and mechanical. Within each grouping there exist a number of pre-defined QA levels which are based on b
importance to safety.
v
'[5] Qualifications for QA/QC Personnel Qualification requirements are presently defined for CA per-sonnel and satisfy the requirements of ANSI N45.2-12 and N45.2-13.
.Similar requirements will be developed for QC personnel prior to their need in manufacturing the first FNP.
.[6] OA Staffing
-OPS will maintain an adequate OA/QC Staff.
[7] OA Program Changes An updated copy.of the Offshore. Power Systems OA Manual is on file in the. NRC ' Reg ion II office. Charges can therefore be revieM by NRC. A more formal system of review for substantive OA; program charges is not objectionable to OPS.
v.
54 Revision 1
-. - - - _ =
~
(8] Other Agency Comparisons
' ()
his is an internal MtC action.
\\ _./
(9] Clarify Reporting Levels The @ organization is clearly defined as is its position in the overall organization. Reporting relationships are likewise clearly specified. These organizational relationships are discussed in Section 17.1 of the Plant Design Report.
[10] As-Built Documents The production of as-built docunents will be throughly treated in the OPS M plan. Maintenance of as-built docunents is the i
responsibility of the plant owner.
[11] Define @ Role in Design The role of M is clearly defined in the body of OPS engineering and % procedures (see Section 17.1.3 of the Plant Design Report for an overview of these requirements). ne collection of O
industry opinions appears to be an internal NRC effort.
v i
55 Revision 1
ITEM II.B.1: REAC'ITR COOUNT SYSTD1 VDflS i
r The FNP will include a reactor vessel head ventirg system which is designed to remove gases from the reactor vessel head via remote-manual operations _ frcan the control rocrn. Instrinnentation for the operation 'of this system is provided by a reactor vessel level instrtsnentation system.
Tne reactor vessel head venting system will discharge into the pres-surizer relief tank in order to accormiodate testirg and potential inadvertent releases of water and steam.
By discharging to the
-pressurizer relief tank, gases in the vessel head are removed from the reactor coolant system and released to the containment through the rupture disc. - Inside the contairraent, hydrogen can be oxidized by means of the post-accident hydrogen recombiners.
The system arrangement provides for ventirg the reactor vessel head by using only safety grade equipnent.
We systern mainly consists of
)
a 1-inch ventirg line with four Safety Class 1, " fail-as-is" isola-tion valves and two Safety Class 2 throttling valves.
To eliminate potential downtime due to isolation valve stem leakage, the system utilizes two normally closed valves.
%e system is designed such that any single active failure will not prevent vessel gas venting nor prevent venting isolation.
We system is capable of rapid disnantlirg for refuelirg and provides the nreseary manual venting functions during vessel filling operations. %c. system connects to the reactor vessel head at the existirs vent pipe.
Persuant to Item II.B.8 the NRC will conduct a rulemaking regarding design features to mitigate severe accidents. As a result, further design features for the control of hydrogen may be required. OPS will implement appropriate requirements in the FNP design once the rulemaking is' completed and the requirements defined.
4
['y L./
56 Revision 1
ITEM II.B.2:
PIJNr SHIEIDING Post-accident release of radioactivity, as described in Regulatory Guide ~1.4, has been used to derive source terms for the current design of the FNP shielding around fluid and ventilation systems that may contain highly radioactive fluids or gases as a result of accidents. _ te existing design includes provision for access to emergency coolant recirculation equipnent for maintenance following a loss of coolant accident, since long term past-accident operation of this equipment must be assured. Followirg is a more detailed stmary discussion of the current FNP post-accident design basis and design features. As part of the detailed design, a comprehensive design review will be conducted to insure that shielding for systems which may contain highly radioactive fluids or gases following an accident is adequate.
Most of the systems which normally interface with the Reactor Coolant System (either directly or indirectly) are isolated frcra the Reactor p)
Coolant System following an accident in which significant quantities N
of radioactivity are released. Release of radioactivity is considered potentially significant if concentrations in the reactor coolant are greater than those associated with 1% failed fuel under normal operating conditions. mose systems which are isolated from the reactor coolant are the followirg:
1.
Gaseous Waste Treatment System (WIG) 2.
Samplirg System (SSR) 3.
Chemical and Volune Control System (CVC) 4.
Boron Recycle Systori (BR9 and 5.
Liquid Waste Treatment System (WrL) III f
(1) In the present FNP design, the Safeguards Area sumps are drained to the Liquid Waste Treatment System. This will be charged such that, follow-ing an accident, liquids collected in these sumps will be pumped back (7
to the contairnent step.
L.Y 57 Revision 1
Se only systems -interfacing with reactor coolant which are not
-f}
isolated are:
a 1.
Safety Injection System (SIS) (for initial coolant injection),
2.
Residual Heat Removal System (RHR) (for coolant recirculation),
and 3.
Containment Spray System (CSS) (for spray injection and recir-culation)
Rese three systems (piping and components) are located within four, separate, shielded safeguards compartments in the FNF.
Shielding thicknesses for spaces in which these systems are located were calculated employing a source derived from Regulatory Guide 1.4.
We source term included 50% of the core equilibrium halogen in-ven%rf and 1% of all other fission products uniformily mixed in the m
containment sump water inventory. Noble gases were rot included in the fluid sources used for design of shielding for these spaces, an assumption which the experience at 'INI indicated to be valid (see also the response to item III.D.3-1).
We sources employed are
- doctraented in Table 12.1.4 of the PDR.
The dose rate criterion for shielding of these systems in safeguards compartments is that the dose in occupied areas outside the shield walls not exceed 3 Ren for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> exposure beginning at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after an accident. Access to these spaces at earlier times is not expected to be necessary.
. h is dose criterion s< 3 Rem for an 8 hoa egsure one day after the accident) 'is -the post-accident shield design criterion for all post-accident work locations on the plant except for the control room and ' emergency relocation area. For the control room and emergency relocation area, the criterion used for shield design is that of
[]
General Design Criterion 19 which is that the dose to personnel V
58 Revision 1
inside those spaces be less than 5 Rem for the duration of the
+o3 C/
accident. Source terms for analysis of the control room and emergency relocation area are based on Regulatory Gui's 1.4 radioactivity relene assumptions.
The RhR, SIS and CSS system components within each safeguards compartment are located in a subcompartment which is isolated from the rest of the safegtards compartment during nomal operation.
Ventilation is provided by a sealed system such that neither supply nor exhaust air lines communicate the subcompartment to the surround-irg space. In the event of an accident resulting in contairment isolation, subcompartment exhaust is lined up to the Annulus Filtra-tion System (AFS). The AFS maintains the subcompartment at a negative pressure, thus assuring that any airborne radioactivity released within the subecopartment is exhausted to the annulus, where it passes through charcoal anc HEPA filters before release to the envir-onment. Because of this unique design, liquid leaks frcn the SIS, RHR p'
or CSS systems will not result in release of airborne radioactivity
(,/
within the surrounding spaces in a safeguards compartment. This configuration is shown pictorially in the response to item III.D.1-1.
Special consideration will be given during final design to post-accident handling of fluids leaking from pumps in the RHR-SIS-CSS subcompartments. In the event of a large leak, recirculation flow from the containment sump to the affected subcompartment can be terminated by closirg the appropriate step isolation valve. These valves are motor operated with the motor outside the shield wall. he operator is connected to' the valve via a reach rod. Manual valve wheels are also provided at the operator so that the valve may be closed even in the event of motor operator failure.
Se FNP has been designed so that post-accident maintenance may be performed on either of the two RHR ptnps by drainirg and flushing the
- RHR equipment. -Drain and flush operations can be performed via reach rod operated valves located outside the shield walls of the RHR pnmp (m.
()
rooms. Airborne activity released to the RHR subcompartment would be 59 Revision 1
swept out by the annulus ventilation system Wich maintains a
. p) negative pressure in the rom. Additionally, the design basis for equipment.important to safety includes a requirement for satisfactory operation followirg post-accident radiation exposure. 'Ihe integrated exposure to safety equipnent, which is calculated using the source term identified above, is a part of the equipnent specification.
To summarize, the existing design @iloso@y for controlling radio-active water and airborne activity following an accident involving core damage is to isolate all systems which could remove radioactive water or air frcan either the contairrnent or the Reactor Coolant System. Systems outside the containment which are needed following an accident for core cooling or contairrnent atmosphere cooling are located within shielded subcompartments, which are part of each separate safeguards compartment. These subcompartments are maintained at a negative pressure and are connected to the annulus following an accident. Source terms specified in Regulatory Guide 1.4 were used for design of shielding for post-accident wrk locations near systems Q
which could potentially contain highly radioactive water.
pL) 60 Revision 1
=
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9
ITEM II.B.3:
POST-ACCIDENT SAMPLING o
Westinghouse has designed a ' post-accident sampling system which meets the NRC post 'IMI requirements. The current FNP post accident sampling system (see PDR Section 9.3.2.2) will be modified to incorporate the Westinghouse system or a similar post-accident samplirg system.
We system employed in the ENP will provide the capability to sample from two reactor coolant hot legs, pressurizer liquid and vapor space, reactor vessel vent, the containment sump contents and the contalment atmosphere durirg and imediately following an accident.
Bis requirement involves the handling and processing of highly radioactive liquid and gaseous samples and analyzirg such samples quickly to provide information important in assessing and controlling the course of the accident. %e FNP system will meet these objectives by piping the samples to a special shielded glove box analysis station where the samples are processed by remote manual and on-line analytical instruments. Dependent upon the gross activity, the samples may be diluted prior to analysis. Specific in-line analyses V
provided are 1) liquid samples - pH, boron concentration, chloride concentration and gross activity, and 2) gas samples - hydrogen concentration and gross activity. Samples may be obtained and removed within one hour after the accident (with a maximtra exposure of 3 rem whole-body) for analysis within two hours for radioactive noble gases, iodines, cesium, ard non-volatile isotopes, for analysis with-in one. hour for boron and for analysis with a shif t for chlorides.
(
)
n./
61 Revision 1
-ITEM II.B.8:
RUIJMAKING CN DEGRADED CORE ACCIDENIS It is anticipated that either a proposed or an interim rule will be Promulgated during the summer of -1980. Offshore Power Systems will
-address the requirements of this draft / interim rule as soon as possible following its publication. 'Ihe response will be in the form of a revision to this topical report.
4 O
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62 Revision 1
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F ITEM II.C.4:
RELIABILITY ENGINEERING p
As a part of. the final design process offshore Power Systems will perfonn reliability analyses for the following srtes as a minimtrn:
o Suberiticality Systems o
ECCS Injection Systems o
ECCS Recirculation Systems o
Shutdown Cooling System o
Containment Spray System o
Safeguards Activation Systems o
Onsite AC Power-o Onsite DC Power o
Instrument Air System o
Safety-Related Cooling Water Systems o
Safety-Related Ventilation and Cooling Systems a
The objectives of these analyses will be to establish overall systems h
reliability estimates and to identify the principle contributions to
%J potential systems failure. Particular attention will be paid to identification of oprator errors, comon modes, single failures and test / maintenance outages which contribute significantly to systems failure probability. 'Ibe results of these analyses will form the basis for appropriate systems design modifications, if required.
'Ihe systems reliability analyses will be subnitted to the NRC within two years after issuance of the Manufacturing License.
3 NJ 63 Revision 1
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ITEM II.D.1:
RCS SAFETf VALVE TEST REQUIREMENTS Safety and relief valve testing is being conducted generically in the EPRI testirg program. The EPRI Program Plan was presented to the NRC on December 17, 1979 and further discussed with the NRC on February 25, 1980. It is presently expected that the EPRI test program will be completed by July 1, 1981. Offshore Power Systems will carefully monitor generic testirg and will demonstrate applicability of the generic test to the Floating Nuclear Plant.
O i
r 1 pd l
64 Revision 1
ITEM II.D.2:
RESEMCH Gi RELIEF AND SAETlY VALVE TEST REQUIRDENTS Offshore Power Systems will verify the applicability of EPRI tests to the FNP and will make - those design charges which the test program shows to be appropriate. Within six months after completion of these tests or issuance of the manufacturing license (whichever is later),
OPS will sutznit a detailed explanation of how recpirements (if any) resultirg frcra tests will be implemented.
O 4
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l 65 Revision 1
l ITEM II.D.3:
RELIEF AND SAFETY VALVE POSITIQ4 INDICATION
,y (v)
Positive indication of pressurizer relief valve position is currently provided in the FNP design.
Such indication is accomplished in the following manner:
1.
Each PORV has indication lights on the control board which are activated by stem-actuated limit switches.
In addition, a position disagreement light / alarm prominently displays a failure of the PmV to achieve the last position comanded.
2.
We temperature downstream of the PORVs and safety valves is displayed on the control board and high temperature alarms are provided.
3.
%e pressurizer relief tank has temperature, pressure and fluid level indication and alams on the main control board.
h 4.
High pressurizer pressure alarms in te Control Room.
a OPS is presently evaluatirg alternate methods to provide safety valve position indication. One such system has been developed and is described below.
Westinghouse has developed an acoustic leek monitoring system that will provide flow indication downstream of the safety valves and thus satisfy the NRC requirements for leakage detection.
The system operates on the principle that turbulent, high pressure flow throtgh an orifice generates an acoustic signa) which is transmitted through-out-the reactor coolant system. ne monitoring system will detect acoustic signals and thus determine valve position. ha ENP will incorporate either the Westirghrause acoustic leak monitor or stem
-mounted limit switches after acceptance by the NRC.
LJ 66 Revision 1
ITEM II.E.1-1: AUXILIARY FEEDOGER SYSTEM EVAlllATION g
(
)
v 1.
Prior to the 'IMI accident Offshore Power Systems had performed a preliminary reliability analysis of the Floating Nuclear Plant's Auxiliary Feedwater System (AFWS). This analysis utilized the same component failure rate data base as the Staff's generic AEWS eval-uation contained in NUREG-0611 and NUREG-0635, and investigated the following three accident scenarios:
a)
Ioss 'of main feedwater with offsite power available.
b)
Loss of main feedwater combined with loss of offsite power.
c)
Ioss of main feedwater combined with total loss of AC power.
For the first two scenarios above, the unreliability of the Floating Nuclear Plant AEWS was found to be in the rarge of 10-5 to 10-4 failures per demand, and for the total loss of AC case in the range
-2 of 10 failures per denand. The difference in reliability between the first two cases and the last one, is due to the fact that during f) total loss of AC power, only the steam driven train is available, and thus no credit can be taken for the redundancies available in the diesel driven trains.
The analysis also revealed, that due to the unique redundancies present in the Floating Nuclear Plant AEWS, independent failures make virtually no contribution to the overall unreliability of the system.
In each of the three cases investigated, the calculated unreliabili-ties were due salely to cx>mmon mode failures. We main contributors in this category were found to be the three normally open, manually operated isolation valves at the exit of the two storage tanks of the AEWS (see Figure II.E.1-1-1). While these valves would be operated very rarely, they could be mistakenly closed (as a result of human error)- prior to a less of main feedwater transient. In the Floating Nuclear Plant design, this incorrect alignment would probably be detected durirg the frequent on-line testiry of the ptnps, via the low pressure alarms in the two suction headers, and by the monitoring (O
/
67 Revision 1
f of individual pump suction pressure indicators during testing, but no
.[
'i credit for such detection was taken in the analysis.
Gi Overall, our review concluded that the Floating Nuclear Plant APdS has above average reliability, as compared to the AFW systems already examined by the Staff. Nevertheless, Offshore Power Systems will re-evaluate the reliability of its AFWS using event-tree and 'ault-tree. logic techniques to determine the potential for AEWS failure under various loss of main feedwater transient conditions, with particular emphasis being given to determining potential failures that oculd result f rom htrnan errors, comon causes, single point vulnerabilities, and test and maintenance outages. We results of this evaluation will be subnitted in appropriate detail within two years of the issuance of a Manufacturing License.
2.
The Floatirg Nuclear Plants Auxiliary Feedwater System is designed in recordance : with the requirements of Standard Review Plan Section 10.4.9. However, a detenninistic review of the system in accordance with -this plan will be carried out and subnitted to the Staff within two years of the issue of a Manufacturing License.
3.-
%e AEW system flow design bases and design criteria have been carefully derived durirg the design evaluation by consideration of the safety-related and non-safety-related functions of the system in the Floatirg Nuclear Plant.
Safety Related Function a)
The AFW system provides feedwater to the steam generators to remove residual heat fran the core and prevent release of reactor coolant through the pressurizer safety valves in the followirg situations:
o Ioss of offsite power o
Ioss of normal feedwater
(~'E.
o.
Malfunction of the Condensate Feedwater System L./
-68 Revision 1 s
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-err i
o Major secondary systen pipe rupture Steam generator' tube rupttre
)
o U
o Control roan evacuation o
Sinking emergency b)
The AEW supplements the IECS flow in removing core residud beat in the event of a snall break LOCA.
c)
The AEW. System is utilized in cooling the reactor coolant down to the cut-in po!nt of the Residual Heat Removal Systan for the sequences listed in a) above, d)
The AEW System maintains the plant at hot shutdown conditions durirg control room evacuation and extended loss-of-offsite power.
e)
- Further clarification of these safety related designs and resultirg system parameters are provided in Section 10.4.6.7 of the Plant Design Report.
NorrSafety-Related Function In addition-to the above functions the Auxiliary Feedwater System is utilized durire normal plant start-up, shutdown and hot shutdown conditions.
As-prt of the final design process, Offshore Power Systems will re-evaluate all of the above requirements, verify the corresponding AFW system functions, and subnit detailed results to the Staff. ' Ibis will.be done within two years of issue of the Manufacturirg Licensa.
p 1:
.&J 69 Revision 1
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ITEM II.E.1-2:
AUXILIARY FEEDWATER Atfl0-START AND FIDW INDICATION
,g
- \\J In the current FNP design, as discussed in Section 10.4.6.7.4 of the Plant Design Report, the four motor driven auxiliary feedwater pumps automatically start on lo-lo level in any steam generator, loss of main feed pump, safety injection signal, or loss of offsite AC power.
The turbine driven ptznp starts autcrnatically on lo-lo level in any two steam generators or loss of offsite power. Automatic initiation signals and circuits for the Auxiliary Feedwater System are Class 1E and can be tested on-line. Manual capability of initiation of the Auxiliary Feedwater System is provided in such a manner that no single failure will result in loss of the system function. No single failure of the autanatic initiation circuitry will prevent manual initiation of the Auxiliary Feedwater System from the Control Room.
Auxiliary feedwater flow channels, with an accuracy of better than the required +10%, will be Class 1E and displayed on the main control board. Each channel of flow instrtsnentation is powered from its
)
respective Class IE instrument power supply.
1v v
71 Revision 1
ITEM II.E.2-1: RELIANCE-CBI ECCS Offshore Power Systems will evaluate relevant operating plant data in order to assess the reliability of the ECCS to perform its intended function. Design changes will be made if shown to be appropriate by these evaluations.
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ITEM II.E.2-3:
IR$ CERTAINTIES IN PERREMMICE PREDICTIONS O
In -response to Item I.C.1 - (Subitem 1), generic mall break IOCA analyses for UHI/ ice containnent plants have been performed and subnitted to the NRC in WAP-9639. These analyses are applicable to the FNP.
Offshore Power Systems will sutrnit the required plant specific mall break IDCA analyses as part of the final safety analysis of the Floating Nuclear Plant.
1 r
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1 73 Revision 1
ITEM II.E.3-1: RELIABILITY OF POWER SUPPLIES FOR NAWRAL CIRCUI.ATION f.\\_)
The FNP provides tha following features which assure a continued supply. of power for the followirs plant components essential to natural circulation flow.
1.
PreFMurizer heaters The total pressurizer heater capacity for the ENP is 1800 IM. _
Four separate backup heater groups (346 IM each) are supplied directly from 4 independent and redundant safety class 480V switchgear buses.
Each bus is supplied fran its respective standby diesel generator following a loss of offsite pwer. The control group (416 IM) is supplied from a non-safety class 480V bus which could be supplied from a diesel-generator bus within several minutes following a loss of offsite power, in the un-likely event that this should become necessary.
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Each independent backup group is large enotgh to maintain natural circulation in the hot standby condition.
The Class - 1E circuit breakers supplying each of the backup groups are tripped open on either a safety injection (SI) or loss of offsite pwer actuation signal.
Se heaters can be manually loaded onto the bus from the main control board after SI is reset and loads required in the initial stages of the incident are ro longer required. Suffi-cient diesel generator capacity is provided to supply the minimtra required number of heaters in the time required. Diesel generator instrtraentation is provided to prevent overloadirg a diesel generator-with these heater loads.
OPS will provide the owner with the necessary procedures for energizing the. pressurizer heaters, including procedures that
.[m[
might be required for load shedding.
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74 Revision 1
eN 2.
Power Operated Relief Valves (PCRV's)
Each MRV is supplied with operating air from a separate Nuclear Safety Class-3 air system which is available followirg 'a loss of offsite power.
Each MRV pilot mlenoid is supplied from independent and redundant 125V DC sources, which are also available following a loss of offsite power.
Se PORV's are controlled frm the main control board. Both PCRV's fail closed on loss of motive or control power.
3.
PCRV Block Valves The PORV block valves are supplied from motor control centers which are readily energized from a correspondirg standby diesel generator following a loss of offsite power.
%e PORV block valves are controlled frm the main control board.
Thus the PORV block valves can also be operated following a loss of
)
,V offsite pwer.
4.
Pressurizer Level Indication Channels All of the - pressurizer level indication channels are derived (and isolated) from their respective protection channels.
The instrtraent loop pwer supplies for these protection channels (including the isolated outputs) are supplied from thnir respective Class _lE Instrtraent buses.
Thus level indication is available ' following a loss of offsite power.
O L,1 75 Revision 1
ITEM II.E.4-1:
CGfrAIl# TENT DESIGN '- DEDICATED PENE1 RATIONS This item is inherently satisfied in the Floatirg Nuclear Plant Design, - because the combustible gas control systems are located inside contairunent.
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O 76 Revision 1
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ITEM II.E.4-2:
CCNTAIl@ENT DESIGN - ISOIATION IEPENDABILITY A
b
'The current Floating Nuclear Plant containment isolation design satisfies all of the provisions of the IRC recommendations as follows:
1.
Phase A isolation (T signal) results in the isolation of all non-essential systems penetrating the containment with the exception of comFonent cooling water lines to the reactor coolant ptznps and the lower compartment which are closed by Phase B isolation (P signal).
Phase A isolation provides for diversity in parameters sensed as well as being autcznatically actuated any time a safety injection signal (S signal) is initiated. Phase A isolation is initiated fran the followirg process variables:
(a)
High steam flow coincident with low steam line pressure or lo-lo TAVG
- v (b) High steam line differential pressure (c)
Low pressurizer pressure (d) High containment pressure (e)
Manual initiation Phase B isolation is initiated from hi-hi containment pressure or manually. Althotgh it is not automatically generated by diverse means, the P signal can only be generated after the T signal, which is diverse, has been initiated. In addition to
-initiating Phase B isolation, the P signal also is used to initiate contairinent spray.
[h.
L) 77 Revision 1
{'s 2.
Offshore Power Systems has.given careful consideration to the systems penetrating the containment which are required to mitigate the consequences of a loss of coolant accident, or any accident calling for containment isolation. %e systems which are required to operate followirg the accidents are as follows:
Safety Injection System
- Residual Heat Removal System (supply lines to cold legs)
Contaiment Spray System (including recirculation strnp lines)
Upper Head Injection System
- Auxiliary Feedwater System The above systems are required to supply cooling and/or make up fluid to the Reactor Coolant System, the containment, and the Main Steam System. %ese systems, or parts of these systems required for post-accident cooling, do not receive any con-tainment isolation signal.
i b
he following systems are rot essential to mitigate the conse-quences of a design basis loss of coolant accident but are considered desirable in assisting in plant recovery from accidents of lower magnitude than a design basis accident. Wey are not part of Phase A isolation, but instead are isolated by the P signal (Phase B isolation).
- Component Cooling Water System (supply and return lines to ICP thermal barrier cooling)
Component Cooling Water System (cooling water flow to the lower compartment fan coolers)
We systems determined to be non-essential are isolated by the T signal (Phase A). They are as follows:
Chemical and Volme Control System
{')
Post-Accident Samplirg System A,/
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g
- Radiation Monitorirg System (contairrnent air sample lines)
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- Nuclear Samplirg System
- Contairinent Ventilation System Post-Accident Containment Ventilation System Liquid Waste Treatment System Service Air System Instrtsnent Air System Emergency Air System
- Ice Condenser Refrigeration System
- Non-Essential Service Water System Reboiler Condensate Return System Reboiler Steam Distribution System Fire Protection Water Spray System
- Safety Injection System (test lines)
- Upper Head Injection System (test lines)
Containment Purge Supply and Exhaust System 3.
All non-essential lines are' properly isolated following the initiation of a containment isolation signal. In addition to the systems which are listed as beirg subject to Phase A isolation, other non-essential systems or lines which penetrate containment have normally closed manual isolation valves, subj ect to administrative control.
4.
Containment isolation reset logic requires deliberate and specific operator action before an isolated line can be reopen-ed. We following control features are provided for containment isolation valves:
a.
_The containment isolation signals override all other autcraatic control signals.
l-b.
We valves will remain in the closed position if the l
initiatirg signal is reset.
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c.
Each ' valve can be opened or closed manually after the 3 [^Y appropriate containment isolation signals are reset.
V d.
Any valves that are nonnally operated in an automatic mode
- (for non-safety functions) are also automatically trans-ferred to manual mode by the isolation signal. This pre-cludes automatic opening of containment isolation valves subsequent to reset of the initiatire isolation signal.
5.
As noted in Table'12.2.1 of the Plant Design Report, containment
~
purge valves are autcznatically closed upon the occurrence of high airborne radiation.
6.
During Floating Nuclear Plant final design the containment high pressure trip point will be reviewed and adjusted downward to the minimtzn compatible with conditions not requiring automatic contalment isolation.
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80 Revision 1 l.
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ITEM II.E.4-4:
CCNTAIl#ENT PURGING vm IV)
Containment purging may be performed by either of two systems (1)
Contalment Pre-access Filtration and Purge System and (2) Post Accident. Containment Venting (Purge) System. %ese systems, which are described in Sections 6.2 arid 9.4 of the Plant Design Report, are designed in compliance with Standard Review Plan 6.2.4 and Branch Technical Position 6-4.
The containment pre-access filtration and purge system provides a continuous purge function at a restricted flow during normal plant operation via an 8 inch diameter supply and 8 inch diameter exhaust penetration. The system is capable of purgirg via a 42 inch diameter penetration; however, the plant owner will be required to resetrict such operation to refuelirg operations or when the plant is in cold shutdown. Procedures will require that the 42 inch diameter purge valves reain closed durirq normal power operation. Each of the two purge lines will be provided with separate containment penetrations, each isolated by two valves in series II)
(qv) he post accident containment venting system provides for hydrogen purge. Se system provides a controlled and filtered containment purge capability by releasing air to the annulus at a maximum rate of 50 SCFM.
Isolation valves are designed to operate against accident pressures and to maintain bubble air-tight closures while performing their i
intended function. We isolation valves have a 2 to 5 second closure time.
The containment pre-access filtration and purge system containment isolation valves are described in Section 6.2.3.3 of the PDR. The 8 inch isolation valves will be included in the operability-assurance plan described in Section 3.9.2.4 of the Plant Design Report.
(1)At present one of the 8 inch purge lines penetrates containment via one of the 42 inch lines (see the Plant Design Report, Chapter 9, Figure N
9.4-6, Sheet 4). An additional 8 inch penetration and isolation valve s~)
will be provided with a contairrnent isolation valve inside and outside the containmen mell. Wis design change provides separation of the two purge functions and maximizes the reliability of the containment isolation function.
81 Revision 1
ITEM II.F.1:
ADD 1TIONM. ACCIDENT MCNITORING INS'IRLMENTATION p}
A.
Containment Pressure:
'Ihe provisions of this recomendation are satisfied in the current FNP design, except that the rarge currently specified for containment
- pressure is 0-18 psig (approx.1.15 times design pressure).
To cmply with the requirement for contairunent pressure monitoring, two additional wide range containment pressure channels will be incorporated into the FNP. These additional channels will rarge from minus 5 psig to 60 psig (4 times design pressure). 'Ihe channels will meet the design requirements of Regulatory Guide 1.97, draft Rev. 2 (Dec. 4, 1979).
B.
Containment Water Level:
As described in Section 6.2.2.7 of the PDR, the Floating Nuclear Plant design does not incorporate a contairrnent strnp as such.
)
U Instead, the containment lower compartment will collect a sufficient volune of water followirg the injection phase of safety injection to allow recirculation.
Redundant safety grade containment water level (wide rarge) measurement is currently provided and displayed in the Control Room. 'Ihe range of these level channels will be increased to cover an elevation equivalent to an 800,000 gallon accumulation, a quantity which includes ice melt and LMI acetzaulator injection.
In addition, Class lE (narrow rarge) level channels will be provided for the local liquid waste treatment system sump at -the 103 foot elevation in accordance with this requirement. These channels will also be used as part of the Reactor Coolant System Leak Detection System.
These-channels will meet the design requirements of Regu-latory Guide 1.89.
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v 82 Revision 1
.C.
Containment Hydrogen Concentration:
r 4
Hydrogen monitors qualified to IEEE-323 and IEEE-344 are not cur-rently available. During FNP final design OPS will select hydrogen monitoring instrumentation which is acceptable to the NRC.
D.
Containment Radiation Intensity (High Range)
The current FNP design for the redundant containment area monitors specifies a rarge of 10-1 to 10 Rad /Hr. This rarge complies with the 7
requirement specified in item 2.1.8.b of NUREG 0578, and Regulatory Guide 1.97, Draft Rev. 2 (December 4,1979.) It should be noted that these detectors for the ENP design are mounted on the outer surface of the steel contaiment but may be considered as "In-contaiment" relative to compliance with this requirement. Se attenuation by the steel shell will be factored into the calibration of the monitors.
Mounting the detectors outside the steel containment serves two safety related purposes:
1) the need for contaiment cable penetra-
\\'
tions is eliminated, and, 2) the monitors will experience less severe postulated accident enviromentional conditions, (i.e., temperature, humidity, and pressure).
E.
Radiation Monitors (Refer also to PDR Sections 11.4 and 12.2):
The current ENP design includes monitors for airborne effluent from three potential release points:
the plant vent, the condenser air ejectors, and the Annulus Ventilation System exhaust vent.
Were are four monitors in the current FNP design which can monitor these three release points:
1)
Plant Vent Particulate Fbnitor 2)
Plant Vent Radiogas Monitor p',w ;
83 Revision 1
3)
Plant Vent Charcoal Cartridge 4)
Condenser Main Air Ejector Monitor
.v-he plant vent monitors (Items 1,
~2, and 3) monitor a sample from either the main plant -vent or the snall annulus ventilation system vent located within the main plant vent.
In order to -comply with this requirement, the plant vent monitoring will be made redundant and will also monitor continuously both release points in the plant vent (i.e.,
both the inner stack for the annulus ventilation re-leases, and the large outer stack).
The radiogas monitor will have 5
an upper detection limit of 10 Ci/cc. We lower end of the range, will be sensitive to a concentration as low as 10 pCi/cc, in order to monitor normal plant releases. %e 12 decades of response will be obtained with a multi range (3 levels) detector.
In order to allow safe collection and analysis of the plant vent charcoal cartridges imediately 'following an accident, provisions will be made in the plant design to do the following (or the equiv-()
alent) :
V 1)-
place the charcoal cartridges in the post accident sampling room and route the sample lines fran the plant vent to the post accident sampling room, which is shielded and habitable immedi-ately followirg an accident, or 2) provide necessary shielding for access to the existing chemistry labs fran the control room, or 3) design - and provide a system for safe remote collection of the cartridges for analysis in the pst-accident sampling room.
We most practical of these alternatives will be selected as the design evolves.
All of the monitor channels will have assured power supplies, independent of offsite pawer (i.e., Class lE or Class 1E Associated Power).
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-6 The condenser air ejector monitor will have a rarge of 10 to 2
10 pCi/cc.
OPS will' participate in and remain knowledgeable of industry efforts to develop methods for monitorirg the concentration of radioactivity in effluents' that might be releared from the main steam safety and atmospheric steam dtmp valves. Once meanirgful methods are available they will be incorporated into the FNP design.
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I 85 Revicion 1
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ITEM II.F.2:
IETECTICN OF INADEQUATE CORE COOLING n
I' V) '
OPS has been evaluatig options developed by the Westinghouse Owners' Group regarding instrumentation for detection of inadequate core cooliry. The preferred option will be selected durirg final design.
Procedures used by the operator to recognize inadequate core cooling will be developed for the instrunentation provided in the final FNP design.
Subcooling Meter An approach being considered by OPS is to provide dedicated, redun-dant, microprocessor-based subcoolirg meter channels with prominent displays on both the Unit Control Console and the Safety Center Panel (Refer to Attachment I.D.4-1 for a description of the FNP Control Board).
Each of these meters would provide a continuous indication of margin to saturated conditions.
He operator could manually select a display of margin to saturation based on either the auc-
/~'N tieneered high incore thermocouple or the auctioneered high loop Thot or Tcold. Auctioneered low reactor coolant system pressure is used for the T calculation by the microprocessor.
Inputs to this would sat system utilize redundant safety grade hot leg and cold leg tempera-tures and reactor coolant system pressure channels. In addition, approximately 8 in-core thermocouple inputs (together with reference junction temperature inputs) would be utilized.
Two setpoints would be utilized to alarm 1) off-normal conditions and
- 2) approach to loss of core coolirg.
Individual sensor channels will l
also be accessible for display.
he subcooling monitor _ hardware will be qualified to IEEE-323 and IEEE-344.
Table lI.F.2-1 provides a summary of tentative design information for the FNP subcoolirg meter.
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8i6 Revision 1 a
TABLE II.F.2-1 INFORMPLTIN REQUIRED W 'IHE SUBCOOLING METER Display Information Displayed Tsat-T, where
'T' is based on (T-Tsat, Tsat, Press., etc.)
either incore or RTD temperatures (Note 1)
Display Type (Analog, Digital, CRT)
Analog (Note 1)
Continuous or on Demand Continuous Single or Redundart Display Redundant Location of Display Unit Control Console and Safety Center Alanns (include setpoints)
(Note 2)
Overall uncertainty ( F)
(Later)
Range of Display 40 F Superheat to 200 F Subcocled
- V Qualifications (seismic, IEEE-344, -323
)
envirorraental, IEEE-323)
Calculator Type (process computer, dedicated Dedicated Digital digital or analog calc.)
If process computer is used specify Not Applicable availability. (% of time)
' Single or redundant calculators Redurdant Selection Logic - (highest T.,
Auctioneered high incore temp.
lowest press.)
or Auctioneered high RCS temp.
vs: Auctioneered low RCS press.
Qualifications (seisnic, IEEE-344, -323 environmental, IEEE-323)
Calculational Technique.
Steam Tables (Steam Tables, Functional Fit, ranges)
A E.-).
87 Revision 1
Input' Temperature (R7D's or T/C's) 8 incore T/C's (2 per quadrant)
(V) 2 Hot Leg RTD's (per loop)
Tenperature (number of sensors 2 Cold Leg RTD's (per loop) and locations T/C ref. junc. RrD's Rarr of temperature sensors Incore T/C's = 150 -2300 F ICS RrD's:
0 -700 F Uncertaingy* of temperature (Later) sensors ( F at 1)
Qualifications (seismic, IEEE-344, -323 environmental, IEEE-323)
Pressure (specify instrument used)
(Note 3)
Fressure (number'of sensors and.
2 (RCS Hot Legs) locations)
Range of Pressure sensors 0-3000 psi Uncertainty
- of pressure (Later) sensors (PSI at 1)
Qualifications (seismic, IEEE-344, -323 envirorniental, IEEE-323)
O Backup Capability Availability of Temp & Press Yes (Note 1)
Availability of Steam Tables etc.
Yes (By Owner)
Training of Operators (By Owner)
Procedures (Later)
- Uncertainties must address conditions of forced flow and natural circulation NOTES:
1.
Individual sensor readouts will also be available, as well as other derived readouts (e.g., temperature differentials, P-Psat based on highest temperature, etc.), utilizing different irdicators.
2.
ho setpoints will be chosen to indicate:
a) off-normal conditions (50 F nominal) and b) approach to loss of core r:coling (later).
3.
Qualified instrtroent will be specified later.
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88 Revision 1
Additional Instrtsnentation m
(v)
OPS is in the process of evaluating the various methods of measuring reactor vessel level that have been investigated by the Westinghouse owners' Group. Le current state of the art appears to favor the use of differential pressure measurment as the best method of detemin-ing vessel leve)
Wis method would utilize sealed reference legs and would rarge frca the vessel top (using an existirg penetration) to the vessel bottom (using an incore instrtsnentation thimble). In addition, taps on the middle of the hot leg pipes would be utilized for level measurement with the Reactor Coolant Pump (s) tripped. All differential pressure measurements would require temperature com-pensation. %e entire level measurement system would be redundant and Class lE, and would be a dedicated system independent of other control or instrumentation channels.
The usefulness of this type of system toward providirg an unambiguous indication of inadequate core cooling er.d an unambiguous indication for vessel ventirg is being evaluated, with attention given to all possible phenomena that could adversely affect the system.
)
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89 Revision 1
ITEM II.F.3:
INS'IRLMENIS FOR MONI'IVRING ACCIDENT CCNDITIONS (Reg. Guide 1.97)
I
)
v OPS has established requirements for Post-Accident Monitoring, includirg a comitment to Reg. Guide 1.97, Rev.
1.
A simnary of the OPS approach to the Draft Revision 2 of Regulatory Guide 1.97 (Dec.
4, 1979) follows. OPS does not expect major problems in complying with the intent of future industry or regulatory require-mants, for reasons given in the response to Item I.D.4.
As required by the Draf t Revision 2 to the Regulatory Guide 1.97, post-accident monitorirg channels will be provided to the FNP control room for the parpose of displaying (or recording, as appropriate) those variables that are required for:
1.
taking preplanned manual operator action 2.
assessirg the process of accomplishirg and maintaining critical safety functions such as reactivity control, core cooling, f )I reacter coolant system integrity, contairraent integrity and radioactive effluent control 3.
monitoring the potential or actual gross breach of barriers to fission product release (including the fuel clad, reactor coolant pressure boundary, and containment) 4.
monitorirg the performance of individual safety systems (e.g.,
safety injection).
-In a3dition, channels will be provided for determining and assessing the magnitude of release of radioactive materials, except for certain variables whose measurement is in the Owner's scope of supply such as monitorirg of radioactivity released to the environment and portable monitors or sampling equipnent.
I
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v 90 Revision 1 l
m-
2e specific channels required for the above functions are stanmarized in Table II.F.3-1.
tis table includes a sumary of the design criteria applicable to each channel.
Se channel ranges described in the table could be expande? as appropriate for future requirements.
For post-accident channels in categories 1,
2, and 3,
described above, the followirg criteria apply:
o Single failure criteria per Reg. Guide 1.53 o
Class lE pwer supply will be provided o
Seismic and environmental qualification per Reg. Guides 1.100 and 1.89 will be provided o
Periodic testing will be provided for in the design.
For post-accident channels in Category 4 (channels that monitor the performance of individual safety systems),. an assured power supply (independent of offsite. power) will be provided, environmental qualification will be provided, and the design will accommodate periodic testirg.
As suggested by draft Revision 2 of Regulatory Guide 1.97, post-accident variables have been selected and will be displayed so as to provide concise, reliable information needed by the operator for post-accident monitorirg.
(3-Q,i 91 Revision 1 l
TABIE II.F.3-1 r-Reg.
Reg.
Reg.
Measured Guide Guide Guide
_y' - Variable Range 1.53 1.89 1.100 Notes Core Exit Temp 150 F-2300 F X
X Note 1 Control Rod Position Full in-Full out X
X Note 2 Neutron Flux 1 c/s-l% power X
X X
~
RCS Hot' Leg Temp 0 F-700 F X
X X
RCS Cold Leg Tap 0 F-700 F X
X X
RCS Pressure 0-3000 psig X
X X
Pressurizer Level Distance between X
X X
taps Degree of subcooling 200 F subcooled X
X X
to 40 F super-heat RCS Loop Flow 0-120% rated flow X
X X
PGV and Safety Valve Positions (see Item III.D.3)
RCS radiation level (Note 3)
(Note 3) X X
Containment Pressure
-5 psig to X
X X
(See Item 60 psig II.F.1)
Contairraent Atmosphere 40 F to 400 F
' Temperature Containment H 0-10%
X X
X 2
Concentration Containment Isolation Full closed-X X
X Valve Position Full open Containment Sump Level (See Item II.F.1) 0 High Rarge Contairrnent 1-10 R/hr X
X X
Area Radiation Steam Generator Pressure 0-1300 psig X
X X
Steam Generator Level Fran tube sheet X
X X
to separators
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92 Revision 1
TABIE II.F.3-1 (cont.)
Rg.
Rg.
Rg.
,-4 Measured Guide Guide Guide
(
Range 1.53 1.89 1.100 Notes Vrriable v
Auxiliary Feedwater Flow 0-110% design flow (Note 4) X X
Main Feedwater Flow 0-120% of design flou X X
X
~ M:in Steam Flow 0-120% of design flow X X
X 2
Condenser Air Ejector 10 to 10 /4.Ci/cc Rad. Monitor
/
Containment Spray Flow 0-110% design flow.
(Note 4) X High Head SI Flow 0-110% design flow (Note 4) X SI Flow 0-110% design flow (Note 4) X Ref. Water Storage Bottom to top X
X X
Tcnk I4 vel Aux. Feedwater Storage Bottom to top X
X X
Tank Level RHR System Flow 0-110% design flow (Note 4) X RHR HX outlet tanp 32 -350 F (Note 4) X Component Cooling Water 32 -200 F (Note 4) X Temp Component Cooling Water 0-110% design flow (Note 4) X Flow Essential Service 32 -200 F (Note 4) X Water Temp Essential Service 0-110% design flow (Note 4) X Water Flow Aux. Raw Wate. Temp 25 -200 F (Note 4) X Aux. Raw Water Flow 0-110% design flow (Note 4) X Essential Raw Water Temp
'25 -200 F (Note 4) X Essential Raw Water Flow 0-110% design flow (Note 4) X Sump levels in
% the level X
~ afeguards areas at which failures -
s would be expected i
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1 93 Revision 1
TABIE II.F.3-1 (cont.)
Reg.
Reg.
Reg.
Measured Guide Guide Guide V'riable Range 1.53 1.89 1.100 Notes Reactor Coolant Drain Bottom to top Tank I.evel Waste Gas Holdup Tank 0-150% design Pressure press.
Safeguards Areas closed-opn (Note 4) X X
Vcntilation Damper Positions Safeguards Areas Space 30 F to 180 F (Note 4) X Tenperatures Status of Class IE As required (Note 4) X Power Sources (Bus voltages, currents, breaker positions)
Status of non-lE Power As required Sources (Bus voltages, currents, breaker positions)
Area Radiation (See Item II.F.1 and PDR Section 11.4)
' ' Airborne Effluent (See Item II.F.1 and PDR Section 11.4)
Radiation Post Accident Sampling &
(See Item II.B.3 and PDR Section 9.3.2)
Analysis Wind Direction 0-360 Wind Speed 1-70 mph Fo r de-termining effluent travel speed &
dilution Outside Temperature
-60 F to 120 F Precipitation Recording Rain Gauge 94 Revision 1
~
ICTES ON TABLE II.F.3-1 1.
Approximately 3-4 measurements per core quadrant will be utilized. Separation of sensor cables will be accomplished to the extent practical.
2.
ho position channels per rod are utilized. Separation of position sensor cables will be acceplished to the extent practical.
3.
Range will be capable of detecting a fuel clad failure which releases 1% of the fuel clad gap and plentra activity into the RCS water. One gama detector will be provided (PDR Section 7.7), with diverse backup using sampling and chemical analysis.
4.
One channel per redundant system is provided. 'Ihe total system instrmentation meets Single Failure Criteria.
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GI l-95 Revision 1 l
ITEM II.G:
ELEC'IRICAL POWER FCR PRESSURIZER RELIEF VALVES, BI4CE VALVES AND IEVEL INDICATION A
1.
Powec Operated Relief Valves (PORV's)
Each PCRV is supplied with operatirg air from a separate Safety Class-3 air system which is available following a loss of off-site power. Each PCRV pilot solenoid is supplied fran indepen-dent and redmdant 125V DC soruces, which are also available followirg a loss of offsite power. The PORV's are controlled from the main control board. Both PORV's fail closed on loss of motive or control power.
2.
PORV Block Valves 4
The NRV block valves are supplied from motor control centers which are readily energized from a correspordirg standby diesel generator following a loss of offsite power. We r,,n v LSek f')
valves are controlled frce the main control board. %us the PORV U
block valves can also be operated following a loss of offsite rower.
3.
Pressurizer Level Indication Channels All of the pressurizer le/e1 indication channels are derived (and isolated) from their respective protection (hannels. ne instrument loop power supplies for these protection channels i
(including the isolated outputs) are supplied from their respective Class lE Instrunent buses. Wus level irdication is available following a loss of offsite power.
/m
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J 96 Revision 1 l
1
ITEM II.H.3:
EVALUATICE AND FEEDBACK PROM MI CIEANUP f]
In April 1979 offshore Power Systerns formed a 21 Task Team to study V
the various accident reports and to temain abreast of developnents.
At about the same time the Westirghouse Owners' Group was formed, an organization with which Offshore Power Systems maintains liason.
Through both the Owner's Group and the mI Task Team, offshore Power Systems will receive and evaluate information from mI cleanup oper-ations. It is anticipated that design requirements might be forth-coming which, for example, would provide for more ready connection of external cleanup equipnent following an accident. Because of the standardized ENP design, an external cleanup system could be skid-mounted. and available for use with any FNP. We FNP design is not expected to preclude such features, and, because of its mobility the ENP might provide additional cleanup flexibility.
( )\\
b 97 Revision 1
i 4.
l ITEM II.J.3:
MMSGEMENT FGL IESIW AND CGIS'IRUCTIN a
i 1he response to this' item-is combined with the response to item I.B.1-1.
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ITEM II.K.1:
I&5: BULIJ!l TINS
(
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The item nebering within item II.K.1 (which is used below) corre-s_-
sponds to that used in NUREG-0660, pages II.K-1 and II.K-2. 'Ihese items, nm bered (1) through (7) address those I&E Bulletin require-ments which are not covered elsewhere in NURrT. -0660.
ITEM II.K.1 (1) : SAFETY-REIATED VALVE POSITIONS Tb the extent that this item may apply to the manufacturing license application, it is addressed in the response to item I.D.3. Primary responsibility for the item rests with the plant owner.
ITEM II.K.1(2):
This item does not apply to the Manufacturing License Application.
ITEM II.K.1 (3): PRESSURIZER IIVEL 'IRIP The Floating Nuclear Plant design already satisfies the required condition. As stated in Section 6.3.2.2.1 of the Plant Design Report, the injection node of emergency core cooling is initiated by the safety injection signal ("S" Signal). This signal is actuated by any one of the following:
- 1. Iow Pressurizer Pressure
- 2. High Containment Pressure
- 3. High Differential Pressure Between Any 'IWo Steam Lines
- 4. High Steam Line Flow Coincident with Either Iow T r Low Steam AVC Line Pressure
- 5. Manual Actuation ITEMS II.K.1 (4 ), (5), (6), (7):
These items do not apply to the mnufacturing License Application.
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99 Revision 1
ITEM II.K.3:
BtX.IETINS AND ORDERS TASK PERCE FINAL RECOP91ENDATIONS T
'Ihe item numberirg within Item II.K.3 (which is used below) corre-
')
sponds to that used in Table C.3 of Appendix C of NUREG-0660.
ITEM II.K.3 (1):
This item does mt apply to the manufacturing license application.
ITEM II.K.3 (1) : REPORT ON OVERALL SAFETY EFFECT OF PORV ISOIATION SYSTEM OPS will complete necessary studies within the required two years after issuance of the Manufacturing License and implement any requirements resulting from those studies prior to shipment of the first ENP.
ITEM II.K.3 (3) : REPORTING OF SAFETY VALVE AND PORV FAILURES Reportiry of safety valve and PmV failures durirg plant operation is an Owner responsibility. OPS will report challeiges and failures occurirg durirg FNP testirg at the Blount Island manufacturing facility.
ITEMS II.K.3 (4), (5) :
These items do mt apply to the manufacturing license application.
G 100 Revision 1
ITEM II.K.3 (6): INS 1RUMEIRATION 10 VERIFY NA1VRAL CIRCUIATION i
Instrmentation to verify natural circulation will be provided in the s
PNP. Studies to determine the need for additional instrmentation are being conducted by the Westinghouse Owners' Group, as part of their analyses of inadequate core cooling (refer to Item II.F.2). Addi-tional instrumentation, if found to be necessary by these studies, will be provided in the ENP.
ITEMS II.K.3 (7), (8):
These items do not apply to the manufacturing license application.
ITEM II.K.3 (9): PID C0ffIROLLER MODIFICATION The pressure interlock setpoint will be raised to the same value as that of the control bistable, per the Westinghouse recornendation.
iv ITEM II.K.3 (10) : REAC10R 1 RIP ON 1URBINE 1 RIP ABOVE 50% POWER This issue need not be resolved in conjunction with the manufacturing license application. At issue is the proposal by certain licensees to reset the anticipatory reactor trip (or turbine trip) such that the trip will be operable only when plant power is at an elevated level.
Floating Nuclear Plant trip systems are capable of set point adjust-ment, and the plant owner will be able to take advantage 'of a favorable determination.
ITEM II.K.3 (ll): CCt#ROL USE OF PORV SUPPLIED BY CCt#ROL COMPONEtRS, INC.
The PORVs are supplied to OPS by Westinghouse. Currently Westinghouse does not supply Control Components, Inc. PORVs.
i-V J
101 Revision 1
(
)
ITEM II.K.3 (12): REACitR 1 RIP CR 1URBINE 1 RIP v
As stated in the Plant Design Report, Section 7.2, the Floating Nuclear Plant is provided with a reactor trip upon turbine trip.
ITZMS II.K.3 (13) through (29):
These items do rot apply to the manufacturing license application.
ITEM II.K.3 (30): REVISED SMALL-BREAK IDCA ME'niODS TO SHOW CCMPLIANCE WITH 10CFR50, APPENDIX K The present Westirghouse snall break evaluation model used to analyze the FNP is in conformance with 10CFR50, Appendix K.
Ibwever,
Westinghouse will address the specific NRC items contained in
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NUREG-0611_ in a model change scheduled for completion by 1/1/83.
G' ITEM II.K.3 (31): PIANT SPECIFIC CAIEUIATIONS TO SHOW COMPLIANCE WITH 10CFR50.46 The present Westinghouse snall break evaluation model and snall break analysis for the FNP are in conformance with 10CFR50, Appendix K and 10CFR50.46. As stated in response II.K.3-30, Westingrouse plans to subnit a new small break analysis to NRC for review by 1/1/83. If the results of the new Westinglouse nodel (and subsequent NRC review and approval) indicate that the present snall break analysis for FNP are not in conformance with 10CFR50.46, a new analysis using the new and approved model.will be subnitted to the NRC within six a.onths af ter 1 issuance of the Manufacturing License or within six months following completion of the generic evaluations, whichever is later.
m.t 102 Revision 1
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t j-these items do not apply to the manufacturing license application.
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ITEM III.A.1-2: EMERGENCY SUPPORT FACILITIES
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The Onsite Technical Support Center (EC) for the ENP consists of the V
supervisor's office and visitors area adjacent to the Control Room.
This center is provided with the same degree of shieldirg, environ-mental control, missile protection and security as the Control Room.
This center uses the same ventilation system as the Control Room and also. utilizes the Control Room radiation monitoring equipment.
Necessary connunication between the TSC and both the Control Rocra and onsite operational support center will be provided. Offsite com munications will be provided by the owner. As outlinM below, plant status can be readily obtained in the TSC during normal as well as emergency operation. Necessary "as-built" documentation will be filed in the EC or elsewhere within the shielded control building.
T.ie TSC is directly adj acent to the Control Roan and access is through a doorway directly into the Control Room. Additionally, a glass window in the comon wall between the TSC and Control Roor..
provides for easy observation of recovery activities. For these
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reasons, the instrtraentation requirenent for the TSC is minimized.
V Offshore Power Systems will provide a CRT terminal to access data frcrn the plant computer. The specific instrtraentation required in the TSC will be determined during final detailed. design of the FNP.
OPS believes that the FNP concept provides unique advantages re-garding as-built documentation, including the following:
a.
greater level of detail on drawirgs (dimensioning, pas *. numbers, etc.) because of the manufacturing concept.
b.
greater consistency and coordination among as-built doctinents, since OPS is ultimately responsible for all as-buil t docu-mentation for the FNP.
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FNP units and their docunentation would be virtually identical,
. allowing use of other mits for full-scale studies regarding recovery operations.
We Dnergency Relocation Area (at Elev.100' and 109' in the control building) beneath the Control Room will be the Onsite Operational Support Center. his ' area is designed to the same criteria for shieldirg, missile protection and envirorrnental controls as the Control Room. Dnergency storage facilities and communications equipnent for onsite operational support are provided. We Dnergency Relocation Area is safely accessible from the Control Room via a stairway which is enclosed within the shielded control building.
We Near-Site Dnergency Operations l'acility will be provided by the plant owner.
l A V
105 Revision 1
)
ITEM II.F.3 INS'IRLMENIS ITR MCEITORING ACCIDENT CONDITIONS (Reg. Guide 1.97) g A
ld OPS has established requirements for Post-Accident Monitoring, includirg a comitsnent to Reg. Guide 1.97, Rev.
1.
A stmary of the OPS approcch to the Draft Revision 2 of Regulatory Guide 1.97 (Dec.
4, 1979) follows. OPS does not expect major problems in complying with the intent of future industry or regulatory require-4 ments, for reasons given in the response to Item I.D.4.
As required by the Draft Revision 2 to the Regulatory Guide 1.97, post-accident mor'itorirg channels will be provided to the FNP control room for the parpose of displaying sor recording, as appropriate) those variables that ute required for:
1.
taking preplanned manual operator action 2.
assessirg the process of accomplishirg and maintainirg critical safety functions such as reactivity control, core cooling, p
reactor coolant system integrity, contairsuent integrity and
'V radioactive effluent control 3.
monitoring the potential or actual gross breach of barriers to fission product release (including the fuel clad, reactor coolant pressure boundary, and containment) 4.
monitorirg the perfonnance of individual safety systems (e.g.,
safety injection).
In addition, channels will be provided for determining and assessing the magnitude of release of radioactive materials, except for certain variables whose measurement is in the Owner's scope of supply such as monitoring of radioactivity released to the environment and portable monitors or sampling equipment.
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l' 106 Revision 1
ITEM III.D.1-1: RADIATION SOURCE CCNIROL OlflSIDE CCNTAIl#4ENT g-O
%e ENP design contains several features which act collectively to minimize the leakage of radioactivity from systems outside contain-ment. These features provide separate, shielded safeguards compart-ments W11ch are watertight with controlled atmosphere that are interconnected with the contaiment annulus during accident cordi-tions, thus preventing any spread of radioactivity to other parts of the plant. These features, which are stmarized below, are further discussed in Section 11.6 of the Plant Design Report.
We responses in Items III.D.1-3 and III.D.3-1 provide additional information relevant to pst-accident s>urce control outsile containment.
Most systens which interface with reactor coolant, either directly or indirectly, will be isolated for accidents which release significant radioactivity to the coolant. The exceptions are:
1.
Residual Heat Removal System (RHR)
O 2.
Safety Injection System (SIS)
V 3.
Containment Spray System (CSS)
Ninerous design features are incorporated in the FNP including de-tection which minimize the potential for the spread of contamination due to. leakage.
These include careful component selectton, proper orientation of valve stems on normally closed valves, and use of valve backseats and/or piped valve leakoffs. Failure analyses and reliability evaluations of safety class systeus serve to identify potential leakage paths durire the design stage.
Besides the features designed to prevent the leakage of radioactivity the potential for the release of radioactivity to the environment is further reduced ' by the provision of secondary isolation and control within sealed safeguards compartments which contain the RHR, SIS, and CSS systems..%e sealed compartments, depicted on Figure III.D.1-1-1, provide a secondary barrier which, together with the' automatic activ-
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ation of the annulus air filtration system discussed in the Response V
107
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to Item III.D.1-3, prevent the spread of airborne radioactivity with-( N, in the plant.
Filtered air discharged to the envirorsnent by the t
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annulus air filtration system is limited to that necessary to main-tain a negative pressure within the sealed compartments.
During normal plant operation the safeguards compartments and contaiment annulus spaces are maintained under a negative pressure with filtered exhaust systems. Exhaust is transferred to the annulus air filtration system on an "S" signal.
At present, the Safeguards Area sump pumps start on sump level signal dischargirg to the floor drain tank of the Liquid Waste Treatment (WIL) System. We Safeguards Area sump flow path will be modified as shown schematically on Figute III.D.1-1-1 so that any accumulated fluid is pumped tack to the containment sump during the post-LOCA recirculation phase. This will be accomplished by utilizing the recirculation phase initiation signal -(low level on WST) to change the position of the air operated valves on the safeguard sump ptrnp discharge. We normally open valves to WrL will be closed and the
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normally closed valve to contalment will be opened. In this manner, no uncontrolled discharge could be made from the safeguards Area to the Auxiliary Buildirg durirg the post-LOCA recirculation phase.
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ITEM III.D.1-2: RADIQhCTIVE GAS MMOGMDir
)
Offshore Power Systes will follow the NRC-sponsored study which will assess the feasibility and desirability of alternatives for the storage and disposal of noble gases released in containment in an accident.
For reasons outlined below, the present design of the Floating Nuclear Plant will not preclude the application of noble gas process-ing, should.such requirements be imposed.
Following an accident, fission product noble gases are initially retained inside containment. Subsequently the noble gases can either be released to the envircrrnent or processed for long-term storage.
We latter option requires that there be a pipe (or duct), along with valving, through which process equipent can connunicate with the containment interior. To be usable the pipe must terminate in a loca-tion which is accessible in the post-accident environment. Isolation
["N valves must be operable under the same conditions. Should the need for such a termination be established, the 8 inch diameter contain-ment purge lines (presently provided in the ENP) could be modified without difficulty to provide the required function. One suitabl e location for a termination is the fuel building at elevation 154 feet. This location is provided with (1) good access to the loading dock, including mechanical handling equipent; (2) adequate floor space for temporary equipent and (3) a charcoal-filtered ventilation exhaust system' to the plant vent.
Examples of-processing options are high pressure or cryogenic storage. Because of the very low probability of eventual use and because gas processing would not be started for some time following an accident, it should not be a requirment to install processing equipent at every plant. Rather, each plant could provide an appro-priate termination for connection of external equipent. Compressors, i
refrigeration equipent, tanks and the like can be pre-assembled into
(^')
skid-mounted. subsystems which are easily transported to the site of
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110.
Revision 1
ITEM III.D.1-3: VENTIIATICBI SYSTDtS AND FILTER CRITERIA 73 (G)
De ventilation systems on the FNP are designed such that adequate provisions are made for post-accident filtration of radioactivity in the exhausts and that acceptable collection efficiencies of radio-iodine adsorbers are maintained during accident conditions.
Se design philosophy for controlling airborne radioactivity is to prov1de a flow of air within the building areas in the direction of increasirg potential radioactive contamination.
Potentially con-taminated areas are maintained under a negative pressure with respect to adjoinirg areas. Clean areas, such as the control room and other habitability areas, are maintained under a positive pressure in order to exclude the infiltration of airborne contaminants. High perform-ance air filters are used in all ventilation systems which introduce outside air. Exhaust air which could contain radioiodine is processed through high efficiency air filters, including charcoal adsorbers.
Such air, once filtered, is exhausted frc.m the plant via a central h) plant vent stack equipped with a radiation monitor-recorder.
V Followire an accident, most of the systems which could transport radioactivity outside containment are isolated. Systems outside the contairment which must comunicate with contairment (in order to provide core cooling and containment cooling) are located within a shielded subecmpartment in each of the safeguards compartments. These subcompartments are maintained at a negative pressure and ete connected to the annulus following an accident. The annulus air filtration system, described in detail in Section 6.4 of the Plant t bu Ker>rt, takes suction on these subcompartments via the annulus and remains in continuous operation,following the accident. %is aspect of ventilation system design is further described in the response to Item III.D.1-1. In general, the annulus air filtration system maintains a slightly subatmospheric pressure in the spaces surrounding the containment. Air is exhausted through charcoal filters thus minimizirg the quantity of radioactive leakage reaching n
I i
V 111 Revision 1
the enviroment. Exhaust flow is limited to that necessary to fm maintain a subatmos@eric pressure.
IV)
De remaining ventilation systems which are maintained in continuous post-accident operation are those serving the control room and emergency relocation areas. These systems autmatically operate in an air recirculation mode under accident conditions with the intake of outside air limited to that required to pressurize the spaces.
Radiation monitors are used for selecting the source of outside air between two different intake locations on the outside of the plant.
We control room and emergency relocation area ventilation systems employ charcoal adsorbers while operating in the closed (recircu-lation) mode. All outside air passes through charcoal adsorbers before enterirg the rom.
In addition to the continuourly operating ventilation systems there are other systems which are maintained on standby status following an accident. %ese are the following:
1.
Post-accident contairment venting (purge) system 2.
Counting room, sampling room, chem hood exhaust, charcoal air filtration 3.
Fuel handling exhaust
%ese systems are automatically shut down following an accident and are subsequently activated if the need occurs. Charcoal adsorbers are used in these systems with the exception of the post-accident contalment' venting -(purge) system, in which case air is discharged to the annulus where the annulus air filtration system performs the necessary filtration. Ductirg and housirgs on high efficiency air
- filtration systems are of all.-welded construction and are subject to
-periodic leak testing.
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l 112 Revision 1
As a further safeguard against the spread of contaminated air into rN clean areas, filtered exhausts are discharged to the plant vent V).
t system which provides multiple booster fans at the base of the central plant vent stack. The plant vent system is designed to maintain the high efficiency air filtration systems (including - the discharge side of the system fans) and feeder ducts under a negative pressure.
The charcoal adsorber air filtration systems described in this response are designed in accordance with Regulatory Guide 1.52 and include provisions for charcoal filter isolation during normal operation ard for surveillance testirg. These features, provisions and testing assure that acceptable collection ef ficiencies of radioactive iodine adsorbers are maintained durirg accident condi-tions. In-service testing to the requirements of Regulatory Guides 1.52 and 1.140 is the responsibility of the plant owner.
Detailed descriptions of the normal and post-accident ventilation systems are presented in Sections 6.2, 6.4 and 9.4 of the Plant V
Design Report.
In general there exists sufficient space and fan capacity to permit additional charcoal filtration, should this be required in the future. Because of the extent of charcoal filtration presently provided, it is unlikely that future requirements will result in significant ENP design changes.
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1 ITEM III.D.2-3:
LIQUID PADMhYS DOSE ANALYSIS
/7
, O Radiological consequences of release to the liquid pathways, includ-ing the effects of interdiction, at a Floating Nuclear Plant site have been exhaustively analyzed and are reported in the following l
docments:
Of fshore Power Systems Report 22A60, " OPS Liquid Pathway Generic Study," June 1977 i
NUREG-0440, " Liquid Pathway Generic Study," USNRC, February 1978 NURB3-0502, " Final Enviromental Statement, Related 1.o Manu-facture of Floating Nuclear Plants by Offshore Power Systems,"
Part III, USNRC, December 1978.
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ITEM III.D.3-1: WGtKER RADIATION PROTECTION
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Offshore Power Systes believes that the present design of the Floating Nuclear Plant is adequate, notwithstanding the events at WI, to support the initial issuance of the Manufacturing License. At a very early stage in the Manufacturing License application worker protection features meeting the requirements of Regulatory Guide 8.8 and Chapter 12 of the Standard Review Plan were included in the plant design. Occupational radiation protection is discussed in Section 11.6 and in Gapter 12 of the Plant Design Report. Other relevant features are discussed in this report in response to items II.B.2, II.D.1-1 and III.D.3-4. 'Ihe fission product source term used in the design of FNP shieldirg for accident conditions generally exceeds the estimated fission product release during the mI accident; compari-sons of stanp water and airborne activities are as follows:
A.
Activity in sump water: On a total @oton energy basis, the sump water fission product activity asstzned for FNP shielding design a roximately two times that found in the sump water at B.
Airborne Activity in Containment: Floating Nuclear Plant Shieldirg design is based on the release of 100% of the core noble gas inventory, 50% of the iodine inventory and 1% of the remainirg fission product inventory. In contrast it has been estimated (2) that only about 50% of the core inventory of Xenon-133 was released to the contairrnent at mI. _0ther specific comparisor's with MI are not available; however, the preceeding asstrnptions are thought to be conservative.
(1)
Based on sump water activities reported in NUREG-0600, p. II-3-64.
j (2)
Presentation by J.A. Gieseke, R.S. Dennirg and H. Jordan of Battelle 1 Coltrnbus laboratories, at USA / ERG Core Melt Research Information Excharge Meetirg, April 11, 1980.
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j As thel events of 1MI are.further considered in rulmakirg and other activities, it is possible, even likely, thac additional ocetpational protection will be required for accident releases. However, because of the design features already included in the Moating Nuclear
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Plant, the impact of new requirements is expected to be mall.
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III.D.3-3:
IN-PIMrr MCNITORING V
offshore Power Systems has designed a comprehensive Radiation Manitor System (INS) for the Floatirg Nuclear Power Plant which provides adequate monitoring for a broad range of plant conditions. Included as part of the INS are both area monitors and airborne monitors.
he area and airborne monitors assist in assuring that occupational radiation exposures to operatirg personnnel are kept as low as reasonably achievable.
Area monitors detect the ambient gamma radiation exposure in selected areas of the DJP. Airborne monitors supplement area monitors in selected areas of the MJP by sampling the atmosphere to detect the concentration of significant radionuclides in particui;cte and/or gaseous form.
In addition to the above general functions, certain FES channels provide indication and/or control for specific functional purposes.
These special INS functions include:
(1) Airborne monitoring of the containment atmosphere for in-contairrnent reactor coolant leakage detection.
(2) High level area monitoring of the containment for following the radiological course of a loss-of-coolant accident.
(3) Area nonitoring of the control room and emergency relocation area to provide automatic switchover of the ventilation systems for these areas to their emergency mode of operation if a high radiation level is detected.
This is an ergineered, safety feature br the FNP designed to insure habitability of these areas following postulated accidents which could result in a significant release of radionuclides to the plant environs.
(4) Area monitoring of the nomal and :lternate outside air intake ducts of the main control room ventilation system.
%ese f]
channels provide supplementary information to the primary wind
' C 117 Revision 1
direction instrtsnentation to allow the operator to verify that
{V]
the least contaminated intake is utilized for ventilation followiry postulated accidents which could result in a signif-
.icant release of radionuclides to the plant environs.
A total of twenty-eight area monitors and six airborne menitors are included in the mS.
'Ihese monitors are described in detail in the Plant Design Report, Sections 12.1.4 and 12.2.4.
-1 The current range of the redmdant containment area monitors is 10 7
to 10 Rad /Hr. In order to comply with Recomendation 2.1.8.b of 0
0 NUREG-0578, this range will be changed ' to 10 to 10 Rad /hr. It should be noted that these detectors for the ENP design are mounted on the outer surface of the steel containment but may be considered as "In-contairraent" relative to compliance with this recomendation.
'Ihe attentuation of the steel shell will be factored into the calibration of the monitors. Mountirg the detectors outside the steel containment serves to increase their reliability because the need for (O
contairunent cable penetrations is eliminated and the monitors will f
V experience less severe postulated accident environmental conditions, i.e., temperature, htraidity, and pressure.
During the final design of the ENP, OPS will review the locations of all area and airborne monitors to ensure the adequacy of the design.
With regard to improved in-plent iodine monitoring, offshore Power Systems will provide space on the ENP for countirg rooms and labora-tories where analyses of radioiodine concentration can be performed.
The location of these spaces and support systems design are such as to permit personnel occupancy for times required to perform necessary analysis followirg accident conditions includirg those specified
'.n NRC position 2.1.8a of NUREG-0578. Shielding will be provided to i
. ensure a low background in the countire roan. Ventilation with clean air at a pressure higher than surrounding spaces will be provided for the countirg room to minimize bac., ground airborne contamination in this region. Capability for purging of entrapped noble gases from A) i'O 118 Revision 1
charcoal samples using either clean air or nitrogen will be provided
~3-in the laboratory area. Residual noble gases will be routed to and (d
vented from the plant stack.
Sampling methods, counting equipnent and other laboratory analytical equipnent will be specified and procured by the plant owner. '1he gamma ray spectrometer is a commercially available method for discriminatirg betweest ~ residual noble gases and radiciodine adsorbed on the charcoal filters in the atmospheric sampling devices. OPS will recomend to the plant owner that such equipnent be procured for analysis of the charcoal filters used for sampling of areas within the facility. OPS will also recomend to the utility owner that portable sampling - devices be procured and available for sampling occupied spaces for radioiodine followirq accidents.
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l 119 Revision 1
III.D.3-4: C0KIROL ROOM IRBITABILITY 7~.s
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%e Floating Nuclear Plant design will accorrvnodate required operating personnel in a safe state of occupancy for the duration of postulated accidents. Wis is accomplished by providing areas within the plant which are protected from external hazards including radiation and toxic gases and 5hich will contain the necessary life support sup-plies to prculde for personnel until replenishment or evacuation can be -affected. %ese areas, the Control Room and the Dnergency Reloca-tion Area, are safeguarded by filtered ventilation systems and by biological shielding.
In Mdition, the Dnergency Relocation Area (ERA) will provide the facilities necessary for occupancy of the Floating Nuclear Plant by the operating crew.
%ese facilities include food supplies and food preparation equipnent, medical supplies, sleeping accommodations, and cor:vnunications equipnent.
Biological shieldirg for the Control Rom has been designed to comply with General Design Criterion 19 (i.e., 5 rem ga::ina whole body dose, m
30 ran thyroid dose, and 30 rem beta skin dose) for the duration of an accident.
Postulated accidents analyzed include the loss-of-coolant accident, fuel handlirg accident, main steam line rupture and gas decay tank rupture.
Detailed results of the Habitability Analyses are given in Section 6.5 of the Plant Design Report in compliance with Standard Review Plan 6.4.
Control Rocrn shieldirg, designed to attenuate the direct radiation from fission products within the containment, consists of a 2 foot thick concrete roof and concrete wall on the side facirg the con-tainment building, and one foot thick walls on the sides not facing the containment.
%e back wall of the control roan consists of a 1-1/8 inch thick steel plate missile shield. Wese walls extend from the 172' elevation down to the 100 foot elevation, which is the level of the - lower floor of the containment building.
%ese walls thus enecr p3ss the Control Roan, the Process Rack Roctn, the Cable pull area, and the ERA. %e four rooms are separated by a floor / ceiling which is a 2-hour' fire barrier.
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Source terms for the radiation analysis are based on applicable Regulatory Guide assumptions.
For the loss-of-coolant accident, Regulatory Guide 1.4 asseptions were used.
%e accident operational mode for the ventilation system is initiated autmatically by a high radiation alarm on any of four monitors:
1) the air particulate and/or gas monitors in the plant vent stack, and,
- 2) the area monitors in the Control Room and Emergency Relocation Area.
Basically the accident operational mode consists of closing all Control Roan and ERA exhaust ports, maintaining a positive pressure, and recirculating the internal atmosphere. %is positive pressure prevents inleakage of potentially contaminated air from surrounding spaces. Independent ventilation systems serve the ERA and Control Roan. Each system also continuously recirculates a considera-ble quantity of air thorugh the filters to further reduce the iodine activity within the ERA and Control Roan. A detailed description of the ventilation system is given in Section 9.4.1 of the Plant Design O
Report.
b The ventilation systems for the control room and GA have dual intakes which are physically separated. Dual intakes allow outside air ' to be drawn from a region where the concentration of radio-activity is relatively low following an accident. %e preferred air intake is-automatically selected in response to a wind direction controller to assure that the preferred intake is on-line continu-ously. Radiation detectors will be used in the ventilation air
-intakes as a precautionary measure to indicate any measurable levels of activity and confirm the correctness of the chosen air intake. We operator can override the automat'c feature. Outside air is brotsht in through charcoal filters at a maximtra rate of 100 cm as necessary to. maintain a positive pressure of 0.25" water pressure.
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% results of an extensive wind tunnel measurements program (Refer-
- g-s ence 1) employing scale models of two Floating Nuclear Power Plants located within a scale model of a typical breakwater were used to determine locations for the alternate control room ventilation
' intakes and to determine the atmospheric dispersion factors at the intakes used in accident analysis.
We Floating Nuclear Plant has been reviewed against the requirements of Regulatory Guide 1.78 and 1.95 and Standard Review Plans 2.2.1, 2.2.2, 2.2.3 and 6.4. As stated in the Safety Evaluation Report, the 1
FNP design meets the applicable requirements.
b(~N d
(1)" Wind Engineering Study of Atmospheric Dispersion of Airborne Materials Released frcn a Floating Nuclear Power Plant", R. N. Meroney, et. al.,
Colorado State University, August,1974.
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.J 122 Revision 1
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