ML19346A415
| ML19346A415 | |
| Person / Time | |
|---|---|
| Site: | Atlantic Nuclear Power Plant  |
| Issue date: | 06/11/1981 |
| From: | Haga P OFFSHORE POWER SYSTEMS (SUBS. OF WESTINGHOUSE ELECTRI |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| References | |
| FNP-PST-050, FNP-PST-50, NUDOCS 8106190291 | |
| Download: ML19346A415 (142) | |
Text
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1 FNP-PST-050 ,
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PDDR 3RIGINAL
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June 11, 1981 Ys g
Ms. Elinus- Adensam, Chief Licensing Branch No. 4 Q " [Q ;"
U.S. Nuclear Regulatory Commission % 3 7920 Norfolk Avenue ',
Bethesda, MD 20852 '
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Dear Ms. Adensam:
Re: Docket STN50-437. Offshore Power Systems Resconse to Post-TMI Requirements Transinitted herewith are twenty-five copies of draft Appendix C to the Plant Design Report, which appendix contains the Offshore Power Systems responses to the post-TMI requirements for pending Construction Permit and Manufacturing License applications. At the conclu-sion of the technical review and uoon resolution of Staff concerns, if any, Offshore Power Systems will file Appendix C formally in Amendment 28 to the Plant Design Report.
Offshore Power Systems is prepared to assist in any way the Staff may desire ir. order to complete the review of these responses and expedite production of the final supplement to the Floating Nuclear Plant Safety Evaluation Report.
Yo' s very truly,
(
P. B. Haga CC: V. W. Campbell A. R. Collier f fiel Enclosure 81061'90 N [
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i APPENDIX C
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Amendment 28
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I TABLE OF CONTENTS
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V APPENDIX C RESPONSES TO POST-TMI REQUIREMENTG Subsection 10 CFR 50.34(e) Subject Page Introduction C-1 (1) (i) Probabilistic Risk Assessment C-3 (1) (ii) Auxiliary Feedwater Evaluation' C-9 (1) (iii) RCP Seal Damage C-13 (1) (iv) LOCA from PORV Failure C-14 (1) (v)-(1) (xi) Applicable to BWR Only -
(2) (i) Applicable to cps Only -
(2) (ii) Plant Procedure Improvement C-15 (2) {iii) ' Control Room Design C-16 (2) (iv) Safety Parameter Display C-45 (2) (v) Bypassed and Inoperable Status Indication C-47 (2) (vi) RCS Vents C-54 (2) (vii) Radiation Design Review C-58 (2) (viii) Post-Accident Sampling C-63 (2) (ix) Hydrogen Control System C-68 (2) (x) SV and RV Qualification C-70 (2) (xi) SV and hV Position Indication C-71 (2) (xii) Auxiliary Feedwater Automatic Initiation /
Flow Indication C-73 (2) (xili) Pressurizer Power Supplies C-74 (2) (xiv) Contairunent Isolation Systems C-76 (2) (xv) Containment Purge / Vent Systems C-80 C-1
TABLE OF CONTENTS (continued)
D V- Subsection 10 CFR 50.34(e) Subject Page ,
(2) (xvi) Applicable to B&W Plants Only -
(2) (xvii). Containment Instrumentation C-82 ,
(2) (xvili) Core Cooling Instruaentation C-86 (2) (xix) Post-Accident Instrumentation C-90 (2) (xx) Power for PORV, Block Valves, Level Instrumentation C-91 t
(2) (xxi) Applicable to BWRs Only -
(2) (xxii) ,(xxiii) Applicable to B&W Plants Only -
(2) (xxiv) Applicable to BWRs Only -
(2) (xxv) Post-Accident Support Facilities C-93 (2) (xxvi) - Leakage Reduction Outside Containment C-95 (2) (xxvii) Radiation N nitoring C-100 (2) (xxviii) Control Room Habitability C-103 (3)(i) Experience Feedback C-107 (3) (ii) Quality Assurance List C-109 (3)(iii) Quality Assurance Program C-113 (3) (iv) Dedicated Containment Penetrations C-122 (3) (v) Degraded Core Matters C-123
. (3) (vi) External Hydrogen Recombiners C-136 (3) (vii) Management of Design and Construction Activities C-137 O
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, APPENDIX C
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RESPONSES TO POST-TMI REQUIREMENTS
. Appendix C provides the Offshore Power Systems' (the Applicant's) responses to the post-MI requirements for pending Construction Permit and Manufac-turing License Applications. %ese requirements were issued initially for coment III in the form of a proposed paragraph (c) to be added to 10CFR 50.34. We basis for the technical requirements set out in the proposed rule making is NUREG-0718(2) ,
In preparation for a Commissioner meeting on my 27, 1981, the Staff made a number of changes to the text of the proposed rule. These changes, which reflect the latest thinking of both the Staff 'i the O)mmissioners, is contained in the public document SECY-81-20D( }.
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At the time of filing Amendment 28, a final rule had not yet been adopted by the Comission. However, during the May 27, 1981 meeting, the Comis-sioners instructed the Staff to conduct their review of the pending Construction Permit and Manufacturing License applications on the basis of the rule most recently proposed by the Staff for Commission approval. We responses contained in their appendix, therefore, address the prop > sed rule as it appears in SECY-81-20D. Each response is preceded by a restatement of the relevant section of the proposed rule. We alpha-nuneric designator (for example II.B.8) appearing at the end of such each restatement is the section of NUREG-0718, Appendix B, frm which the technical content of the rule is drawn.
(1) Federal. Register (46FR18034), March 23,1981.
(2) USNBC Licensing Requirements for Pending Applications for Construction Pennits and Manufacturing License, NUREXI-0718, March 1981.
(3) Memorandtsn from W. J. Dircks to the Comissioners, SECY-81-10D, May 18,1981, Enclosure 3.
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REGUTATION 10CFR50.34 (e) (1) (i) f
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Subject:
Probablistic Risk Assessment To satisfy the following requirements,. the application shall provide sufficient information to describe the nature of the studies, how they are to be conducted, estimated subnittal dates, and a program to ensure that the results of such studies are factored into the final design of the facility:
Perform a plant / site specific probabilistic risk assessment, the aim of which is to seek such improvements in the reliability of core and contain-ment heat removal systems as are significant and practical and do not impact excessively on the plant. (II.B.8) 0FFSHORE POWER SYSTEMS RESPONSE Introduction Following is a description of the proposed risk / reliability program for the ,
Floating Nuclear Plant includirg an outline of f.he program scope, methodol-ogy, and schedule. %e objective of the program will be to identify improvements in the reliability of core and contaiment heat removal functions as are significant and practical and do not impact excessively on the plant.
Project Scope r
The FNP risk / reliability program wi'l be similar in scope to the Interim Reliability Evaluation Program (IREP) being performed by the NRC on several operating plants. Individual accident sequences and their probabilities will be analyzed to identify the initiatirg events and plant systenv' component failures *ich are dominant contributors to the potential for core damage. The initiatiry events to be analyzed will be determined during the initial phase of the study. As a minimun, the systems listed in Table C-1 will be analyzed to determine if system modifications are appropriate ,
and could significantly reduce overall plant risk.
Methodology ;
The approach to be used in the program will employ event tree / fault tree methodology similar to that used in Wash 1400 and other comprehensive plant !
O riex stodies. *e m ser taexe inve1ved ere discessed eetew.
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(,) Initiating Event Selection A list will be established of initiating events which, together with system failures, have the potential for causiry core damage. This will be accom-plished through a screening of the accidents and transients identified in P[R Chapter 15, WASH 14 , and other studies to identify the basic set of initiating events requiring operation of the key safety systems for core protection and release mitigation. The frequency of these initiating events will be estimated based on available data including WASH 1400, EPRI NP-801, and, pertinent plant-specific information.
Event Tree Development For each type of initiating event, an event tree will be constructed, identifying the systems required to mitigate the event and the expected effect on ability to maintain core and containment integrity given success or failure of each system involved. The full event tree will be reduced to reflect system interdependencies and required sequences of operation. Note p that 'he event trees will address certain non-safety systems such as offsite power and the power conversion system. ,
System Failure Modes and Effects Analysis (FMEA)
For each safety system involved, a FMFA will be conducted to identify and tabulate component and comon cause failures and their effect on system operability for each initiating event. We EMEA will provide docmentation of the basis for inclusion or exclusion of specific failure modes in the system fault tree analysis. Failure modes will include mechanical and electrical faults, operator error, maintenance or testing outages, etc.
Particular attention will be paid to ptential comon cause failures which could disable multiple cmponents. Comon cause failure mechanims to be investigated include environmental factors, operator or mair*enance errors, passive failures and systs interactions.
System Fault Tree Analysis Using the FMEA as input, fault trees will be constructed for each safety p systs identifyirg the failures (basic events) and their logical combina-V tions sich will result in system unavailability (top event) . We fault C-4
tree will be analyzed to determine the minimal cut sets and failure combi-C nations which are the dminant contributors to systen unavailability. Using the appropriate component failare data, a quantitative assessment of ovec-all systen unavailability and of the dminant cut ' sets will be performed.
Data Base Deve '_
A component failure mun base for use in system fault tree analysis will be developed from recognized reference sources including WASH 1400 a;;f IEEE 500. In addition, prototype-specific failure data will be requested from vendors of selectal cmponents (e.g., diesel generators) . The data base will identify the types of components and estimated median failure rates on demand and, where appropriate,- per hour of continuous operation. Error ranges will be assigned to each median value to reflect the mcertainty in the data base. The data base developnent task will include methodologies to adjust failure data to account for varying testing and surveillance strategies. Test and maintenance unavailability contributions will be included based on preliminary technical specifications and typical nuclear plant operating and maintenance procedures.
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Human error rates will be estimated for required or corrective actions by control roan operators and for maintenance or testirg operations which are included' as failure modes in the system fault trees. Available human error and performance data, .' ncludirg those provided by NUREG/CR 1279, will be used.
Containment Responsa Analyses Based on studies of core melt phenomenology and containment transient calculations which evaluate the interrelated physical processes taking place within containment, containment event trees will be constructed for spccific categories of core melt event sequences. These plant event categories will cover all inputs from the plant event tree analysis. We event tree techr.ique permits tracking of the containment response to a cagraded core accident. W rough careful definition of each node, the containment event tree will be applicable to all core melt accident
, sequences. Additionally, the framework is provided to evaluate the relative C importance to risk of related physical phenmena (such as hydrogen burning)
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and to . evaluate the relative merits of various preventive and mitigative (m) features.
Accident Sequence Probabilities The unavailability or failure probability of each system will be calculated by inputting the appropriate failure rate data into the systs fault tree analysis. 'Ihe various accident sequences, as represented by the branches on the event trees will then be quantified by inputtirg the system failure probabilities determined from the quantitative fault tree analysis. Each individual accident sequence will be classified according to release category and the total probability of a given release category will be obtained by the sumation of all accident sequence probabilities assigned to that category.
Uncertainty Analysis Quantitative results will be reported in terms of point values of a probability distribution function, includirg expected (mean) or median (50th percentile) value and upper (90th percentile) and lower (10th percentile) uncertainty bounds. These point values will be detennined based on a propagation of component failure data, including error ranges, through the fault trees and event trees. The uncertainty propagation will be performed using standard statistical distribution functions (e.g. log-normal) or numerical (e.g. Monte Carlo) techniques.
Sensitivity Analysis The results of the study will be reviewed to determine the relative importance to risk of the various accident sequences and to identify those which are the dominant contributors. Within those sequences, the signifi-cant systs and cmponent failure modes will be determined. Cm parisons with existing risk studies, including WASH 1400, will be made to identify arxl explain any significant differences.
The sensitivity of the results to assumptions regarding component or common cause failures will be evaluated by varyirg the assmed failure rates and determining the resultant effect on system failure rates and overall
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Schedule r] OPS will perform the risk / reliability analysis on a time scale such that
.results from the . evaluation can be factored into the design. We plant risk / reliability program outlined above will be completed within two years after receipt of the Manufacturing License. Since significant additional design work- on the Floating Nuclear Plant is not anticipated during the i next few ye irs, results of the analysis can readily be factored into the plant design.
Application of Results to Final Design There are currently no established regulatory requirements or acceptance criteria for judging the acceptability of quantitative system reliability j analyses. Thus the need for implementing changes in design or in opera-ting, testing, or maintenance procedures to achieve improvement in system p reliability will be based on judgemental criteria which are not directly related to licensing requirements. Wese acceptance critaria will be established during the program and will include both quantitative and qualitative assessments of potential design changes taking into account impact on plant cost, schedule and availability.
Following completion of the base line reliability analysis, the results will be reviewed and various options available for improvement in reli- !
ability will be evaluated with respect to die established acceptance l criteria. Recomendations will be made regarding changes in the design or in recommended plant procedures and the reliability analysis will be <
revised to reflect those selected for implementation. t t
Routine design changes will also be evaluated on an ongoing basis. A determination will be made regarding the effect of any proposed design change on the reliability analysis results. If the change is expected to ;
affect reliability, the reliability anabrsis will be revised and the results reviewed for acceptability and need for further modifications as
', described above. In this manner, the Risk / Reliability Program will be .<ept [
current with respect to design modifications and a mechanism will be in place to evaluate reliability-related changes for acceptability as the O
v design is finalized.
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- g.e TABIE C-1 Systems On Which Reliability Evaluations Are To Be Performed ,
- 1. -Auxiliary Feedwater
- 2. Essential Raw Water
- 3. Residual Heat- Removal .
- 4. . Diesel Generators (including support systems)
- 5. Upper Head Injection
- 6. Essential Service Water
- 8. Ccmponent Cooling Water
- 9. Safety Injection ,
, ' 10. Safeguards Compertment Ventilation
- 11. -Control Building HVAC
- 12. Ice Condenser i 13. Electric-Power (lE Po ur supplies, buses & breakers)
- 14. Protection System
, 15. Radiation Monitoring (1E channels) ;
- 16. Annulus Air Filtration 1
- 17. Dnergency Irstrument Air
- 19. Containment Isolation i
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REUUIATION 10CFR50.34(e) (1) (11) 7 V
Subject:
Auxiliary Feedwater Evaluation To satisfy the following requirements, the application shall provide ,
sufficient information to describe the nature of the studies, how they are to be conducted, estimated subnittal dates, and a program to ensure that ;
-the results of such studies are factored into the final design of the '
facility:
I Perform an evaluation of the proposed auxiliary feedwater systems (AEWS),
to include (applicable to PWR's only): (II.E.1.1)
(A) A simplified AEWS reliability analysis using event-tree and fault-tree logic techniques.
(B) A design review of AEWS.
(C) An evaluation of AEWS flow design bases and criteria.
OFFSHORE POWER SYSTEMS RESPONSE Prior to the mI accident Offshore Power Systems had performed a prelim-inary reliability analysis of the Floating Nuclear Plant's Auxiliary Feedwater (AEW) System his analysis utilized the same component failure
{} rate data base as the Staff's generic AEW System evaluation contained in NUREU-0611 and NUREU-0635, and investigated the following three accident scenarios:
a) Loss of main feedwater with offsite power available.
b) Ioss of main feedwater conbined with loss of offsite power.
c) Ioss of main feedwater combined with total loss of AC power.
Fc - the first two scenarica above, the unrei tability of the Floating Nuclear Plant AEW System was found to be in the range of 10-5 to 10-4 failures per demand, and for the total loss of AC case in the range of 10-2 tailures per demand. We difference in reliability between the first two cases ard the last one, is due to the fact that during total loss of AC power, only the steam driven train is available, and thus no credit can be taken fot the redundancies available in the diesel driven trains.
Overall, our review concluded that the Floating Nuclear Plant AEW System
. O) w has above average reliability, as compared to the AEW systems already l
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examined by the Staff. Nevertheless, Offshore Power Systems will re-c3 evaluate the reliability of its AEW System using event-tree and fault-tree
'U logic-' techniques to determine the potential for AEW system failure under various loss of main feedwater transient conditions, with particular ;
emphasis being given to determining potential failures that could result from htsnan errors, comon causes, single point vulnerabilities, and test and maintenance outages. The results of this evaluation will be sutznitted !
in appropriate detail within two years of the issuance of a Manufacturing License. (See the response to 10CFR50.34(e) (1) (1) .
The Floating Nuclear Plants Auxiliary Feedwater System is designed in accordance with the requirements of Standard Review Plan Section 10.4.9.
However, a deterministic review of the system in accordance with this plan will be carried out and sutznitted to the Staff within two vears of the issue of a Manufacturing License.
f The AFW system flow design bases and design criteria have been carefully derived during the design evolution by consideration of the following safety-related functions of the system in the Floating Nuclear Plant.
a) The AEW syst . provides feedwater to the steam generators to remove residual heat from the core and prevent release of reactor coolant through the pressurizer safety valves in the following situations:
o Ioss of offsite power o Ioss of normal feedwater o Malfunction of the Condensate Feedwater System o Major secondary system pipe rupture o Steam generator tube rupture ,
o Control room evacuation o Sinking emergency b) The AEW supplements the ECCS flow in removing core residual heat in the event of a small break IDCA.
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c) The AFW Systen is utilized in coolirg the reactor coolant down to the cut-in point of the Residual Heat Reecval System for ~ the sequences listed .in a) above.
d) The AEW System maintains the plant at hot shutdown conditions during control rom evacuation and extended loss-of-offsite power.
As part of the final design process, offshore Power Systems will re-evaluate the above requirements, verify the correspondirg AEW system functions, and subnit detailed results to the Staff. 'Ihis will be done within two years of issue of the Manufacturing License.
As noted below, the present design of the ENP Auxiliary Feedwater System generally satisfies the recomendations contained in the staff position paper entitled, "NTCP Acceptance Criteria, Task II .E.1.1, Auxiliary Feedwater (AEW) System Evaluation." The one possible exception is the suction piping arrangement, which will be examined as a part of the probabilistic risk assessment required by 10CFR50.34 (e) (1) (i) . (See item b, below) .
a) 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 t
level in any steam generator, loss of main feed pump, safety injection signal, or loss of offsite AC Fower. The turbine driven pmp starts automatically 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 lE and can be tested on-line.
Manual capability to actuate the Auxiliary Feedwater System is provided in such a manner that no single failure will result in loss of the systen function. No single failure of the automatic acutation circuitry will prevent manual actuation of the Auxiliary Feedwater System from l the Control Room. I b) The arrangement of AEW suction piping (including isolation valves) will be examined as a part of the probabilhtic risk assessment required by U 10CER50.34(e)(1)(1) . Changes in suction line design which are found to C-11 I
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l contribute significantly to systein availability will be made during FNP ic) final design. Offshoce Power Systems expects that staff recomendation (4) (b) will be implemented on this basis. However, Offshore Power Systems - prefers to await the study results before proposing detailed changes to the AEW suction piping arrangements.
c) Operation of the turbine-driven AEW subsysts is initiated, monitored and controlled by instrumentation which will continue to operate for a period in excess of two hours followirg coincident loss of both the offsite and onsite AC power supplies.
d) The AEW system is housed in areas which are protected frm t;rnado missile damage in accordance with Regulatory Guide 1.117. Ebrther, the AEW system is designed to seismic Category 1 requirements in accordance with Regulatory Guide 1.29. Werefore, additional pump suction protec-tion fra tornado or earthquake damage is not required.
e) Redundant level indications and low-level alarm functions will be A
() provided for the AEW storage tanks. We low level alarm point will be set so as to provide a minimm of 20 minutes w rning to loss of suction, assumirs operation of the turbine-driven pmp (the pmp with the highest flow rate) . AEW storage tank level and alarm will be provided in the control room.
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l REGULATION 10CFR50.34(e) (1) (iii) m
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RCP Seal Damage To satisfy the following requirements, the application shall provide sufficient information to describe the nature of the studies, how they are to be conducted, estimated submittal dates, and a program to ensure that the result ; of such studies are factored into the final design of the facility:
Perform an evaluation of the potential for and impact of reactor coolant pmp seal damage followirg small-break IIXA with loss of offsite power. If damage cannot be precluded, provide an analysis of the limiting small-break loss-of-coolant accident with subsequent reactor coolant pmp seal damage.
(II.K.2.16 and II.K.3.25)
OFFSHORE POWER SYSTEMS RESPONSE For the RJP, reactor coolant pump seal injection is provided by the- CVC chargirg pmps. In the event of a loss-of-offsite trwer (LOOP), a "B" or blackout signal is generated. 'Ihe "B" signal automatically starts the chargiry pumps on redundant Class 1-E buses to provide BCP seal injection flow at 10 seconds af ter the IOOP. In addition, the OCW pumps are loaded on
, at 20 seconds after the LOOP to provide thermal barrier coolirig. Thus it may be seen that the reactor coolant pump seals are adequately provided with both seal injection and thermal barrier cooliry which will preclude seal damage and a subsequent increase in seal leakage.
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i REGUIATION 10CFR50.34 (e) (1) (iv)
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Subject:
LOCA from PORV Failure To satisfy the following _ requirements, the application shall provide sufficient information to describe the nature of the studies, how they are to be conducted, estimated subnittal dates, and a program to ensure that the results of such studies are factored into the final design of the facility:
Perform an analysis of the probability of a mall-break loss-of-coolant accident (LOCA) caused by a stock-open power-operated relief valve (PORV) .
If this probability is a significant contributor to the probability of small-break _ LOCA's fra all causes, provide a description and evaluation of the ef feet on small break IDCA probability of an automatic PORV isolation systs chat would operate when the reactor coolant system pressure falls af ter the PORV has opened. (Applicable to PWR's only) . (II.K.3.2)
OFFSHORE POWER SYSTEMS RESPONSE An analysis tus been performed by Westinghouse and reported in W.AP-9804 dated February 1981, which estimates the probability associated with a small break loss-of-coolant accident (LOCA) caused by a stuck open power h~J operated relief valve (PORV). The WCAP concludes that a significant reduction in the frequency of a small break IDCA, due to a stuck open PORV, has been achieved by plant modifications made subsequent to 'IMI . Specif-ically, the probability is estimated to be 2 x 10
-6 per reactor year for Westinghouse plants. Since the probability of a small-break LOCA from all causes is approximately 1 x 10-3 per year, FORV failure is not a signifi-cant cont.ributor. Based on historical plant data, no failures of a PORV to close tuve occurred in domestic Westinghouse plants.
In additio:., the consequences of a small break LOCA caused by a transient-related PORV opening ani failure to reclose has been analyzed in K.AP-9601 dated June 1979. Even vith conservative licensing basis assmptions, no core damage is expected for this compound transient as reported in WCAP-9601.
REGULATION 10CFR50.34 (e) (2) (11)
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- Plant Procedure Improvement To satisfy the following requirements, the ' application shall prov'ide sufficient information to demonstrate that. the required actions will be satisfactorily completed by the operating license stage. This information is of the type customarily required to satisfy 10CFR50.35(a)(2) or to address unresolved generic safety issues.
Establish a program, to begin during construction and follow into operation,- for integrating and expanding current efforts to improve plant procedures. The scope of the program shall include emergency procedures, reliability analyses, human factors engineering, crisis management, operator trainirg, and coordination with INPO and other industry efforts.
(I.C.9) 0FFSHORE POWER SYSTEMS RESPONSE Each plant owner will be responsible to the NRC for the preparation and updatirg of plant operating procedures. Offshore Power Systems will assist 4
plant owners in discharging this responsibility by serving as a clearing-house for -important information derived from Floating Nuclear Plant in-
! service experience, design developments and experience gained during testing.
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REI3UIATION 10CFR50.3i(e) (2) (i11) v
Subject:
Control Room Design To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. Wis information is of the type customarily required ta satisfy 10CER50.35(a) (2) or to address mresolved generic safety issues.
Provide, for Comissim review, a "ontrol rom design that reflects state-of-the-art human ractor principles prior to mmitting to fabrication or revision of fabricated control rom panels and layouta. (I.D.1)
OFFSHORE POWER SYSTEMS RESPONSE offshore Power Systems will provide a control room design that reflects state-of-the-art human factor principles by the application of design criteria $1ch will assure the ability of the control room operators to prevent anticipated transients frcra developiry into accidents and to cope with accidents should they occur. We control room design will be provided to the NRC prior to comittirg to fabrication of the control rom panels.
O The ENP Control Room design bases, criteria and general functional re-ouirements are presently adequate for meetiry all of the concerns identi-fled after the 'IMI accident. A sumary of the relevant design bases, criteria, functional specifications / descriptions for the FNP Control Room is presented in the attached excerpts from the Control Room Specification.
Further description of safety systs status monitority (includirg descrip-tion of the light and alarm sequences for control board modules) is presented in the response to 10CFR50.34 (e) (2) (v) . The design bases, criteria, and functional specifications for the FNP Control Room meet or exceed Draft IC of IEEE-566 and Draf t 3 of IEEE-567. OPS does not expect problems in implementing forseeable future requirements, such as the human factor seguirements to be published as NUREI] 0700, because:
- 1. OPS practice for control rom design evolution includes a fo rmal, ongoing multi-discipline review process (including the use of a full-scale mockup) . Any significant changes will be implemented with q that process. Final design review will use the methods beiry developed by Sandia and NRC.
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- 2. The design concept is flexible (modular construction and modular l display software will be used).
- 3. The status of detailed design is such that schedule constraints are not i expected.-
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l EXCERPFS MM 1HE FNP CCNIT40L ROOM SYSTDI SPECIFICATIO4S n.,
1.0 Control Room Design Bases and Criteria 1.1. Equipment Design Basis
1.1.1 Classification
The control room panels, their foundations and supporting structures are classifial as follows:
Panels that lave Class lE equipment or circuits, are classified as Clast 1E, Seismic Category I.
Panels that Inuse no Class 1E equipment or circuits are classified as Class NIE Seismic Category II.
r] The classification, lE or NIE, of the individual display and control devices and their associated wirinj are as designated in their respective schematic connection diagrams. The PNP Instrtsnent List will provide a composite listing of all the devices nounted on the pinels and include, amon) other infomation, the classification.
1.1.2 Environment
The main control panels and all equipment nounted therein shall be designed to operate under the followin) atmo-spheric conditions:
O Tmperature S C to 50 C (40-122 F)
Pressure 1 Atm litanidity 10 - 90% nit A)
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NOTE: We Control Building Air Conditioning System will
-) maintain the control room at 75+ 2 F and a relative hurr.idity of 40 - 70% under all postulated plant conditions.
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
i he basic criteria for the design, fabrication and testing of the main control panels are derived fran the application of IEEE 279 and IEEE 384 to the Class lE equipment and circuits contained therein. -
i a
The main mntrol panels must provide sufficient support and n
physical protection to its class lE equipnent and circuits
(.j to enable them to perform their essential functional requirements before, during, and after motion conditions, up to and including design basis notion conditions.
t The panels shall be so designed that, at the frequencies and accelerations of the floor resulting from design basis j.
motions, they do not amplify the forces beyond the level at I which the equipment contained therein is qualified to j function properly. To meet this requirement, the panel shall be designed with sufficient cigidity so that no I natural frequencies or resonances can exist at a frequency l less than (later) Hz. Welded stiffeners, diagonal braces [
and thick plata skin shall be used singly or in any combination thx t .11 satisfy this rigidity requirement.
The panel d ?si gn shall be seismically qualified in accordance with the requirements of IEEE 344.
O C-19
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%e panel design shall. include provisions for securely
() mounting the board to its supports. A dynamically equiva-lent support shall be used in the seismic testing of the panel.
Plug-in or slide mounted equipment 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 equipment or . cir-cults. ,
The internal structural desic,.. of the main control panel shall provide for the physical separation of redundant .
Class lE circuits and equipment, as required by IEEE 384 so that no single credible event t in prevent the proper functioning of any Class lE system. W e required separation shall be achieved by an adequate air space or a fire-retardant barrier between redundant Class lE circuits and equipnent. The circuit wirirn shall be supported in a manner that will assure maintenance of the air space throughout the design life of the panel. The required separation shall be maintained from the point of entry of the circuit into the panel to the final tennination on the surface mounted devices. Non-Class 18 circuits and equipnent shall likewise be separated fran all Class lE 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. We structural framework arx! surfaces of the panel shall be fabricated of steel or altr.iinun stock. Any nonmetallic V
C-20
components and devices should be manufactured frca self-extirx3uishing material as defined by ASTM Std. D635-1972.
(v). - .
. Paints or other applied surface preparations shoeld contribute only nominally to the total combustible potential of materials or components in or on the panel.
Consideration should be given to the release of toxic or corrosive gases and denso snoke 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 *. Quality levels shall be achieved through the specification of requirements known to promote high quality.
All control and instrtunent wiring shall have sufficient mechanical strergth, current capacity, thermal rating and insulation characteristics to meet the circuit and installation requirements established by plant design.
Wire and cable insulation shall be flame retardant with self-extirguishirg nonpropagatirg characteristics. Cc.:-
sideration shall be given to the potential release of toxic or corrosive gases and dense smoke and their possibk-effects upon personnel and equipment. All wire and ca. ale installed within the main control panels must be capable of meeting the flame test requirements of IEEE 383.
i '
- Mean Time Between Failures: 20,000 hr. -
Mean Time To Repair: 30 min.
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v C-21 l l
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1.2 Operator-Interface Criteria
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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 unler normal and upset conditions and present the in such a manner as to enhance his mder-standing of the plant status and raluce the probability of an operational error.
1.2.2 Information and Control Requirements:
The determination of Wich 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 systs ergineer. This process is
()
O formalizad by evolution of the following controlled documents:
- 1. Process System Specifications
- 2. Instrument Block Diagrams
- 3. Control Ingic Diagrams
- 4. Schematic Connection Diagrams Throughout this design process, each system is analyzed frm an operating point of view as well as frm a design point of view. We control systems engineering group is responsible for the systes integration as well as for the l application of control and display hardware.
l i
In the process systs specifications, the information and control requirements are defined in functional terms. Wese functional requirements are further defined in the Instru-(mv) ment Block Diagrams and Control togic Diagrams and then C-22
converted to ha rdware requirements (i.e. indicators,
_,3
( ) lights, switches, etc.) in Schematic Connection Diagra,ms, v
We I&C hardware requirements for all the FNP systems are consolidated into a single document - the FNP Instrument List. 21s document is a computerized list of all the I&C ha rdwa re provided on the FNP and includes sufficent information to completely describe each item. Included in the bank of information is the ;nountirg location for each item.
1.2.3 Arrangement Requirements:
The configuration and arrangement of the control center panels ard the placement of indication and control devices on the panels shall be based on the following:
The FNP 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 acccaino-
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dating additional operating personnel during periods of high activity when it is desirable to relieve the burden of the lead operator.
In multi-ogrator situations, the following organizations will be assumed:
The lead operator will be singularly responsible for the safe conduct of control center operations. The secord 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. The 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)
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C-23
w s whenever an abnormal situation is noticed. 'Ihe lead operator. will likewis+ be responsible for keeping the assistant operator (s) informed of the plant status.
Functional Areas -
The control center shall be subdivided into distinct operatirg areas. The functional requirements defined for each area determine the major criteria for the allocation of display and control devices within the control center.
These operating areas are designed to provide for a separ-ation of safety-related systs and auxiliary a:d supporting devices from those required by the operators to nonitor and control the plant urder normal condP ions. This method of device allocation af fords a reduction in number of displays that the operator must observe under normal conditions and consequently, a reduction in the probability of misinter-pretation anri erroneous action.
\
V The operating areas to be incorporated in the ENP control center design are as follows:
Normal Operations Area The normal operations area is the primary control location for the Floating Nuclear Plant under hot non-upset condi-tions. 'Ihe display and control devices located in this area will be the m%imtrn required for the operator to perform the following:
- 1. to assess the status of the plant and its systems at any time l
- 2. to be alerted to abnormal situations and changes in l plant status l t
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C-24 l l
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, f- 3. to maintain the plant in a safe hot shutdown condition l ')
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- 4. to maneuver the plant fra a hot shutdown condition to full power operation
- 5. to manually initiate safety systems (on a system level)
Safety System Operating Area This 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
,a tj 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 This area is allocated to display and control devices needed to perform auxiliary or supportirg functions that j are required infrequently. (An example m uld be the devices that are used only duriry heatup, cooldown, cold sh stdown and refueling.)
A C-25
s Historical Records Area i\
V his area is allocated to the devices thich are required to provide hard copies of computer stored data.
4 Area Arrangement The arrangement of the operating areas within the control center will be consistent with the followim criteria:
- 1. We normal operations area will be centrally located ard 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 from the normal operations arec and not be in a separate enclosure.
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- 3. We historical records area shall be located apart frczn the operatirg areas.
- 4. We supervisor's office shall be located as to give him a visual comand of all control center activities.
1.2.4 Htrnan Enginer: ring Requirements:
The following human engineerirg considerations shall be factored into the design of the ENP Control Center.
Anthropometric Considerations -
We control panels will be designed to permit 5 to 95 percentile (in height) operators to read or reach all indicators and controls from a standing position in front f>)
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%/ of the auxiliary panels and from a seated or standing C-26 l
position in front of the consoles. %e 5th percentile
- (v) operator is 5'4" tall and the 95th percentile operator is 6'4" tall .
Task Analysis -
We assignment of controls and displays to the functional areas and 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:
- l. the operating nodes during which the operator needs the device,
- 2. how of ten the operator uses the device when it is *
>O needed, t 4 V i
- 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 i operator take corrective action.
i Other Considerations -
During the last few years a number of human engineering reviews of existing nuclear power plant control panels have been conducted by the Electric Power Research Institute, W Research Laboratories and others. The results of these i studies shall be used to develop a checklist for the ENP design to ensure that the typical deficiencies noted in Table C-2 are avoided.
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C-27 i
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i 1.2.5 Panel Mock-Up: i A full-scale control center mock-up will be constructed and ,
j used to evaluate the human engineering aspects of the ENP !
design. %e evaluation will include " walk-throughs" of the l FNP operating procedures to ensure that no; operational i 2
problems are overlooked. l i i i .
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2.0 ff1CTIONAL DESCRIPTION
(') Figure C-1 shows the preliminary ENP control room layout and the
. locations of the various functional areas. The correlation of equip-ment with these areas is as follows:
(1) NORMAL OPERATIONS AREA - - UNIT CONTROL CONSOLE
, (2) SAFETY SYSTEMS OPERATING AREA - SAFETY CENTER (3) INFREQUENT OPERATIOl AREA - SECONDARY CONTROL CENTER (4) HIS1DRICAL RECORDS AREA - COMPUTER OPERATORS CONSOLE 2.1 Unit Control Console:
The Unit Control Console (UCC), shown in Figure C-2, is a compact modula rized console f rom which all normal plant operation is conducted. It includes all the displays ano controls necessary to brirg the unit frcan hot shutdown to rated power (and back to hot shutdown) and fo r controlling and monitoring load changing f'} operations. The UCC design permits one-man operation while providing space for two.
1 1
The UCC provides three computer generated visual displays (CRT's).
i These displays, together with their associated keyboards, provide the operator with all the information he needs to assess the status of
- the plant and its systems at any time. The center CRT contains a i process overview in which all key parameters are continuously updated i
in a single display. The left hand CRT contains an alarm display showing the status of all points in alarm. The right hand CRT is used to display parameters for any selected systen in detail.
The remainder of the UCC is arranged in stations that are dedicated to those portions of the followin3 systens used durirg normal i operation:
- 1. Rod Control System i Rod Position System
,A 2. Nuclear Instrumentation System C-29
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,_ 3. Reactor Coolant System
(
y) 4. Chemical and Voltane Control System
- 5. Feedwater and Condensate Systems and Auxiliary Feedwater
- 6. Main Steam System
- 7. Turbine 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.
2.2 Auxiliary Panels 2.2.1 Safety Center:
The Safety Center, shown in Figure C-3, provides for the monitoring, control and testing of the FNP protection and engineered safety features systems. We panel is arranged in stations that present a logical flow of information to the operator. We left most station provides the displays and controls associated with the Reactor Protectiori System (SSPS) and those displays and recorders required for Post-Accident Monitoring. The stations located intnediately to the right contain the component level displays and controls for the ESF systems. These include:
- 1. Upper Head Injection
- 2. Safety Injection
- 5. Essential Service Water
- 6. Essential Raw Water Also included in this area are the CRT and keyboard and ' !
systm level binary status displays required by the Protection and Engineered Safety Feature Availability and Test System.
C-30
(v) To the right of tle ESP stations will be located the indicators and contro.s associated with IE support systens including:
i
- 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:
The Secondary Control Center, shown in Figure C-3, contains all of the required control and indicators that are not located on the UCC or the Safety Center. We arrangement of the systens generally follows the order in which they are used in bringing the mit to power operation. We 5
controls and indicators located here are primarily those ,
that are used only during refueling, heatup, cooldown and ,
maintaining cold shutdown.
2.2.3 Component Arrangement:
t r
The controls and indicators required for the operation of l each individual systen will be integrated into a comon work station. We arrangement of components within the wark .
station will follow the placement of the controlled compon-ents in the actual process system. For complicated systems or those used infrquently, a graphic display will be pro-vided above the work station. A typical example of this arrangenent is shown in Figure C-4.
! f VO C-31
fm 2.2.4 Annunciators:
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= Annunciators will be located along the upper sloping por-tion of the safety center panels. These will be conven-tional tardwired alarm points and will be used as backups in the unlikely event that the computer generated alarm display is inoperative. Wis use of annunciators will be restricted to alarming 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 C-1, is centered at the
! computer operators console. At this cor. sole an operator will be able to obtain a hard copy of CRT displays, computer calculations, test results, etc.
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C-32
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3.0 EQUIPMENT DESCRIPTION
, G 3.1 Unit Control Console The Unit Control Console (UCC) will be a free standing sit-stand
. console as defined by IEEE-27. Figure C-2 shows the general size and shape of the console.
The UCC will be provided with front and rear removable panels for access to internally mounted equipment and cable terminations Equipment will be mounted within the panel with ~a view towards maximum accessibility for testirg and maintenance.
The 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 rouns.
V Horizontal raceways will be provided for supporting cable along the 1ergth 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.
The 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 equipment will be by the most direct route consistent with the following division separation requirements.
- 1. 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.
d C-33
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_ 2. Within one foot of the panel surface the division separation may (a) 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 dien, due to operational considerations, it is desirable to mount equipnent belorgirg to different divisions at adjacent locations. We justification for the one and one-half inch separation will be provided through analysis ard testirg. The analy-sis will show that ro single credible event, with the exception of an internally generated fire, could prevent the proper functioning of a 1E system. We testing will demonstrate that, with the low energy 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. The panels will be provided with rear access doors and removable front panels.
Provisions for cable entry and routing are similar to those described in section 3.1.
3.3 Instrument Modules:
The majority of the discrete display and control functions required on the panels will be accanplished usirg a modular system of irstru-mentation. We nodules are all of the same height and their widths are multiples of a fixed modular dimension. The basic modules in-clude: an indicator module; and auto / manual module; a p2shbutton module; and a recorder module.
A typical indication module, shown in Figure C-5, has two vert.' .1 displays. Each of the displays will provide one percent reading l h) i v accuracy and will be scaled in engineering mits.
C-34
l T l The auto / manual module, shown in Figure C-6, contains four wJ backlighted pushbuttons and an edgewise indication. The auto / manual F
module together with an indicator module will be used to perform auto-manual control ftxtions. In this application, one ' of the displays will be used for tae measured variable and the other for the set point.
The pushbutton control module, shown in Figure C-7, is the primary binary control and indicatirg means. The module can contain up co six backlighted pushbutton operators and each button can be splo to display two messages.
A recorder nodule will be used eenever a hard copy record of a process variable is required and the plant computer cannot be used to provide it. 'Ihe recorder modu'.e 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 te removed with the circuit active without affectirg the state of the o ntrolled component or parameter.
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- C-35
TABLE C-2 i \
V TYPICAL HUMAN ENGINEERItG DEFICIENCIES Reading Indications Reconsnended viewing distance exceeded Meter. design causes glare and improper viewing angle Control design obscures position setting Reaching Controls Functional reach exceeded Workirg posture leads to accidental activation Activating Controls Inconsistent direction of novement relationships between control and associated display Violation of operator expectation of direction of movement of control Nomenclature that violates operator expectation Use of same nomenclature for different functions Different nomenclature used for functionally identical controls Interpretirg Coding 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 coding Interpreting Llarns No differentiation of the severity of alarms Nuisance alarms A
C-36
TABLE C-2 (CONT'D) ,
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(d Incatirg Caponents
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No delineation between major control systems
' i Side by side location of functionally unrelated controls that :
are identical in appearance j
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Incompatible arrangements of associated displays and controls F i
Illogical arrangement of related controls j
, Inconsistent location of the same type of control f i
Performing Sequential or Simultaneous Operations !
Spatial separation of controls that must be used together j I
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C-41 TYPICAL f10DULE ARRANGEf1ENT FIGURE C-4
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C-43 TYPICAL AUT0/fiANUAL t10DULE FIGURE C-6
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C-44 TYPICAL PUSHBUTTON CONTROL MODULE FIGURE C-7
_, REGUIATION 10CFR50.34 (e) (2) (iv)
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Subject:
Safety Parameter Display
- To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the ' required actions will be satisfactorily mmpleted by the operating license stage. his information is of the type custmarily required to satisfy 10CER50.35(a) (2) or to address mresolved generic safety issues.
Provide a plant safety parameter display console that will display to operators a minimum set of parameters defining the safety status of the plant, capable of displayirg a full range of important plant parameters and data trends on demand, and is capable of indicating When process limits are beirn approached or exceeded. (I.D.2)
OFFSHORE POWER SYSTEMS RESPONSE The ENP will be equipped with a safety parameter display system (SPDS) which will be designed to the following criteria:
- 1. Input parameters will be selected and the parameter data processed in q such a way as to yield concise, reliable indication of the status of h the following safety functions: core cooling, reactor coolant sub-coollry, reactivity control, control of primary coolant irc/entory, coolant temperature and pressure control, and containment of radio-activity.
- 2. The display (output) format will be designed in accordance with human ergineerirg principles so as to provide visability ard ease of infor-mation interpretation.
- 3. Diversity, redundancy, and error-checking will be utilized to assure reliable safety status indication.
- 4. Display devices will be prominantly located on the Safety Center and output information will also be available for display on a CRT in the Unit control Console. (See Figures C-1, C-2 and C-3 for the basic Control Rom ard panel arrangements) . Provisions for duplicate infor-
.. mation display in the Dnergency Operations Facility (Plant Owners
) Scope) and the Technical Support Center will be provided.
C-45 I
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l f' 5. The design provisions for safety-related display instrumentation described in Section 7.5 of the PDR will be utilized so as to comple- l s
y ment the SPDS. ;
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i i l 6. 'Ihe design of the SPDS will be consistent with NURM-0696, Febrmry (
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C-46
7-REUIATION 10CFR50.34(e) (2) (v) h
Subject:
Bypassed and Inoperable Status Indication To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. 'Ihis information is of the type custemarily required to satisfy 10CFR50.35(a) (2) or to address mresolved generic safety issues.
Provide for automatic indication of the bypassed and operable status of safety systems. (I .D.3)
OFFSHORE POWER SYSTEMS RESPONSE Since as early as 1974, Offshore Power Systems has been comitted to total conformance with Regulatory Guide 1.47 (Plant Design Report Section 7.1.7) .
The Floating Nuclear Plant presently includes significant provisions for status monitoring; these are outlined in the followiry paragraphs.
Assurance of proper operation and/or positioning of safety-related equip-m ment (includirg equipment in ergineered safety features supportiry systems) during all operating activities is provided by:
Main Control Board (MCB) Display Features: including position / status indicating lights, position / status disagreement indication, availsbility indication, and system level bypass indication. These features which meet or exceed Regulatory Guide 1.47, are as follows:
Position / Status Indicating Lights (Backlit Pushbutton) (PIL)
Backlit red (open) and green (closed) pushtuttons indicate actual valve position fran limit cwitches on the valve. The rushbutton is part of the MCB module for that valve.
Backlit red (on) and jreen. (off) pushbuttons indicate breaker or contacter status frm approptlate auxiliary contacts. The pushbutton is part of the MCB module for that component (pump, fan, etc.)
These position / status signals are also inputs (through isolation devices)
(3) to the Plant Computer Systems.
C-47
( ,
) Valve Position Indicating Lights (Lights Only) (PIL*)
This valve p)sition signal is also an input to the ?lant computer Systems.
Position / Status Disagreement Light / Alarm (Backlit Pushbutton) (PDL)
A backlit alarm indication / acknowledgement pushbutton (normally extinguish-ed) 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 indicatirg light flashes. This backlit pushbutton is part of the PCB module for that equipment.
Both of these flashing lights are acknowledged by this pushbutton, changing the alarm indication pushbutton fran flashing to steady, and the comanded PIL from flashing to extinguished. 'Ihe steady alarm indication light is rot extirguished until the comanded and the actual equipment state are in agreement.
Availability Light / Alarm (Same Backlit Pushbutton as PDL above) (AVL) p b) If the equipnent is removed from service (i.e., if motive power is un-available or locked out) either deliberately or due to failure, the backlit alarm indication / acknowledgement pushbutton (the same device actuated by the PDL) flashes in conjunction with an audible alarm.
For equipnent removed from service, this alarm signal is also an input (through an isolation device) to the Plant Computer System. 'Ihe Plant Computer System flashes a system level display (BYP) on the MG indicating that the appropriate system ESF train is bypassed.
System Level Bypass Indication (BYP)
An engraved backlit window, prominently displayed to the operator, is provided for each division of each major Safety Subsystm (e.g., SIS, RHR) .
- indicates " Lights Only", see Table C-3.
a C-48
This window flashes tenever any of the following conditions (within. the
( ) scope of the window) indicates a bypass of a protective action:
a) Motive pwer mavailable to an ESP actuation device (for example, an MOV, power unavailable to the reversirg contactor), due to deliberate bypass or circuit failure. his condition is derived from " AVL" signal . (AVL /BYP) b) Valve p sitioned so as to create a bypass of a protective action. Wis condition is derived from actual valve position. (PIL/BYP) c) Window activated manually by operator from MCB, responding to informa-tion received through administrative control. (ADM/BYP)
If two redundant divisions of any subsystem are concurrently placed in a bypass mode (due to any of the above -inputs), the second division window would flash and an audible alarm wuld occur. Acknowledgement of the first division level bypass causes the first window to change from flashirg to p) g steady, until the bypass is cleared. Acknowledgement of the second (con-J current) division level bypass silences the audible alarm, but leaves the second window flashing mtil one of the bypass conditions is cleared.
The plant cmputer systes perform the combination and sequence logic that is required to control the system level bypass indication windows. We position / status inputs to the ccmputer that are derived from Class IE control circuits are isolated in accordance with Regulatory Guide 1.75.
The bypass indication system meets or exceeds the requirements of Regula-tory Guide 1.47. Additional design criteria for the bypass indication systm are provided in Section 7.5.1 of the PDR.
System Level Monitor Indication (MON)
An engraved, backlit window, prominently displayed to the operator, is provided for each division of each major safety subsyste (e.g., SIS, CSS) .
This window flashes, in conjunction with its corresponding PDL light (s),
ob C-49
whenever any equipnent (within the scope of the window) has failed to respond to an ESF signal.
v control Circuit Design Features: In addition to the display features described above, circuit design features are provided to assure proper alignment of equipment. tese 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 equipment is in service (i.e., while motive power is available to i t) , its control priorities are assigned such that ESF signals will always override non-EST signals (with the exception of electrical and mechanical circuit protecticn features which must override ESF signals in order to prevent component damage) .
Failure Mode of Actuation Device (FM)-
Removal of an air operated or solenoid operated valve from service (i.e.,
id removing motive power) will cause the valve to move to the safe position.
Administrative Control Input (Manual) to Bypass Indication System (ADM)
%e system level bypass indication (BYD) can be manually input by the operator through administrative control. Cbmputer software supplements plant administrative controls by tracking these manual inputs (together with non-manual inputs), determining the system level effects, and pro-viding appropriate displays.
- Table C-3 illustrates the specific application of these design features to the generic types of FNP equipnent that could be incorrectly operated. to table indicates which of the FNP control and display design features provide direct defense against
a) The effects of mispositioned circuit breakers or contactors b) %e effects i,f mispositioned valves, or p -
b C-50
w
) c) Undetected mispositionirg of equipment for various canditions of plant operation and for various types of equipment.
Considered in the table are: ;
f a) The nature of the safety system bypass (deliberate vs. inadvertant)
L b) The plant operating mode : periodic test, maintenance, etc.) l c) The engineered safety features systems rode (standby vs. active) l j
d) The type of safety ' equipnent (circuit breaker, motor operated valve, l
, hand operated valve, etc.) i 1 !
Table C-3 applies only to ENP components which are important to safety. 'Ihe table does not include other types of design features (e.g., process ,
alarms) that in some cases would further enhance safety.
i Additional Control Roan design infomation is contained in the response to 10CER50. 34 (e) (2) (iii) .
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C-52
TABIE C-3, Sheet 2
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M 1. Valves are locked in safe pasition, and are under adminis-trative control.
M 2. " Operator error" includes failure to recognize a valve that is left improperly positioned (for power operation) following startup.
NOTE 3. Safety-related hand operated process valves that have the capability of significantly degradirg a protective action if left mispositioned are subject to the following cri-teria:
a)' If normally operated nore frequently than once per year with the plant at power, shall be locked in the safe position mder administrative control. In addi-tion, remote position indication shall be provided.
(~\
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b) If normally operated at startup, shutdown and/or refueling, shall have the provisions of paragraph a),
c) If only operated for non-routine maintenance or repair (e.g., to isolate a punp or heat exchanger for repair) with the plant at power, shall be locked in the safe position under administrative control.
m 4. 'Ihis 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.
l.
m 5. Where trore than one design feature provides defense, the j most prominent one is listed first.
O N.)
C-53 j
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- REGULATION 10CFR50.34 (e) (2) (vi)
- )
Subject:
RCS Vents v
To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will' be satisfactorily completed by the operating license stage. %is information is of the type customarily required to satisfy 10CFR50.35(a) (2) or to address unresolved generic safety issues.
Provide the capability of high point venting of non-condensible gases from the reactor coolant system and other systems that may be required to maintain adequate core cooling. Systems to achieve this capability shall be capable of being operated from the control room and their operation shall not lead to an unacceptable increase in the probability of loss-of-coolant accident or an unacceptable challenge to containment integrity. (II.B.1)
OFFSHORE POWER SYSTEMS RESPONSE The FNP will include reactor vessel haad and pressurizer vent systems which are designed to remove gases from the reactor coolant systen via remote manual operations from the control room. We reactor vessel head and pressurizer vent systems are completely independent systems which provide the capability of venting separately the reactor vessel head and the U pressurizer. We reactor vessel head and pressurizer vent systems will discharge into a well ventilated area of the containment in order to ensure adequate dilution of combustible gases.
The reactor vessel head vent system flow diagram is shown in Figure C-8.
The system arrangement provides for venting the reactor vessel head by using only safety grade equipment. We system mainly consists of 1-inch vent piping with four Safety Class 2 " fail closed" isolation valves. To eliminate potential downtime due to isolation valve seat leakage, the systan utilizes all normally closed valves. The isolation valves are powered from redundant Class 1-E buses. We system is designed such that any single active failure will not prevent vessel gas venting nor prevent venting isolation. We system is capable cf being dismantled with relative ease for refueling, and provides the necessary manual venting functions during vessel filling operations. All piping and equipment between the orifice and the discharge point is Safety Class 2.
Ov C-54
E The pressurizer vent system flow diagram is shown in Figure C-9. The system
'n
( )~ arrangement provides for venting the pressurizer by using only safety grade
-wJ equipnent. The system consists of 1-inch vent piping with four Safety Class 2 " failed closed" isolation valves. % eliminate potential downtime due to isolation valve seat leakage, the system utilizes normally closed valves.
The isolation valves are powered from redundant Class 1-E buses. %e system is designed such that any single active failure will not prevent pressuri-zer venting nor prevent venting isolation.
The systen connnects to the Nuclear Safety Class 2 pressurizer vapor sample line, which is normally filled with steam. We pressurizer venting system is of Nuclear Safety class 2 up to the discharge point to the containment.
The contribution of these vent systems to the probability and consequences of small-break WCA will be evaluated durirg the probabilistic risk assessment discussed in the response to 10CFR50.34 (e) (1) (i) .
[
t C-55
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l C-56 FIGURE C-8 i
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FIGURE C-9
REGUIATION 10CFR50.34 (e) (2) (vii)
,m
'( )
Subject:
Radiation Design Review To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage.1his information is of the type custmarily required to satisfy 10CFR50.35(a) (2) or to address unresolved generic safety issues.
Perform radiation and shieldirg design reviews of spaces around systems that may, as a result of an accident, contain TID 14844 source term radioactive materials, and design as necessary to permit adequate access to imp;rtant areas and to protect safety equipment from the radiation environ-ment (II.B.2)
OFFSHORE POWER SYSTEMS RESPONSE Post-accident release of radioactivity, as defined in Regulatory Guide 1.4, has been used to derive source terms for the existirg design of the ENP shielding around equipment in systems that may contain highly radioactive fluids or gases as a result of accidents. The existirg design includes provision for access to energency coolant recirculation equip ent for o maintenance followirg a loss of coolant accident, since lorg term post-() accident operation of this equipment must be assured. Following is a more detailed sumary discussion of the current FNP post-accident design basis and design features. As part of the detailed final design of the ENP, a emprehensive design review will be conducted in accordance with NUREU-0737 to insure that shielding for systems which may contain highly radioactive fluids or gases followirg an accident is adequate to meet dose rate cciteria for vital areas, or potentially vital areas, on the plant. Should the additional review so indicate, design modifications will be implemented to permit adequate post accident access or to protect safety equipnent from the radiation environment.
Preliminary review of the vital areas on the plant indicates compliance with NUREG-0737. Those vital areas are: the control room, the Technical Support Center, and the post-accident samplirs area.
The FNP control rom has been designed to meet General Design Criterion 19, p assuming continuous occupancy over 30 days following a Design Basis bv )
C-58
accident. Se average whole body gamma dose rate in the control room is <5 A mrar/hr., which is -less than- that required by item 3a of NUREG-0737. We r i V detailed dose analysis for the FNP control room is given in Section 6.5 of the Plant Design Reprt. @e FNP onsite Technical Support Center (See the response to 10CER50.34 (e) (2) (xxv) which will be located adjacent to the Control Room, is provided with the same degree of radiological protection as the control room. We post-accident sampling area (see the response to 10CER50.34(e) (2) (viii)), which will be located between the control room and the shield building, will also- be provided with radiological protection such that GDC 19 criteria will be met.
On the ENP, controls for actuation of the post-LOCA hydrogen control system, containment isolation reset controls, manual ECCS alignment, motor center controls, vital instrument panels and emergency power supply actuation are all located inside of the Control Building.
Most of - the systems which normally interface with the Reactor Coolant Systm (either directly or indirectly) are isolated from the Reactor
,o s Coolant System following an accident in which significant quantities of d radioactivity are released. (Rielease 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.) Those systems which are isolated from the reactor coolant system are the following:
- 1. Gaseous Waste Treatment System (WIG)
- 2. Sampling Systen (SSR)
- 3. 01emical and voltine control System (CVC)
- 4. Boron Recycle System (BRS), and
- 5. Liquid waste Treatment System (WrL)III (1) In the existing EUP design, the Safeguards Area sinnps are drained to the Liquid Waste Treatment System. We design will be changed such that, following an accident, liquids collected in these sumps will be pumped back to the containment sunp. (See the response to 10CFR50.34 (e) (2) (xxvi) .
~
v C-59
%e only systems interfacing with reactor coolant which are not isolated r3 are:
w]
- 1. Safety Injection System (SIS) (for initial coolant injection),
- 2. Residual Heat Removal Syten (RHR) (for coolant recirculation), and
- 3. Containment Spray System (CSS) (for spray injection and recirculation)
These three systems (piping and components) are located within four, separate, shielded safeguards compartments in the FNP.
Shielding thicknesses for spaces in which these systems are located were calculated employing a source derived in accordance with Regulatory Guide 1.4. The source term includes 50% of the core equilibriun halogen inventory and 1% of all other fission products uniformly mixed in the containment sunp water inventory. Noble gases are not included in the fluid sources used for design of shielding for these spaces, an assumption which is justified for recirculated depressurized cooling water. The sources employed are documented in Table 12.1.4 of the PDR.
D.
U The criterion for shielding of these systems in the safeguards compartments is that the dose in potentially occupied areas outside the shield walls shall 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. mis dose criterion (<3 Rem for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> exposure one day af ter the accident) is the post-accident shield design criterion for all post-accident wark locations on the plant except for the vital areas previously discussed. In addition, the FNP is designed with an Emergency Relocation Area (as part of the Control Building) which is located at the 100' elevation. The Emergency Relocation area is provided with the same degree of radiological protection as the Control Room and is designed to accomodate personnel safely throughout the course of an accident. We Onsite Operational Support Center is located in the Dnergency Relocation Area (see the response to 10CFR50.34 (e) (2) (xxv) .
The RHR, SIS and CSS system components within each safeguards compartment are located in a subcompartment which is isolated from the test of the c safeguards compartment during nonnal operation. Wese systems are the only U)
(
C-60
ones Wich are likely to ' oontain post-accident radioactivity. Ventilation
[] ,
is provided by a sealed system such that neither supply nor exhaust air s
's lines 'coninunicate the subcompartment to the surrounding space. In the event of an accident resulting in containment isolation, subcompartment exhaust is. lined up to the Annulus Filtration System (AFS). 'Ihe AFS maintains the subcompartment at a negative pressure, thus assuring that any airborne radioactivity released within the subcompartment is exhausted to the annulus, where it passes through charcoal and HEPA filters before release to the environment. Because of this unique design, liquid leaks from the SIS, RHR or CSS systems will not result in release of airborne radioactivity within the surrounding spaces. 'Ihis configuration is shown pictorially in the response to 10CFR50.34(e) (2) (xxvi) .
Special consideration will be given during final design to post-accident handling of fluids which may leak from pmps in the RHR-SIS-CSS subcom-partments. In the event of a la rge leak, recirculation flow from the containment stanp to the affected subcompartment can be terminated b/
closing the appropriate sump isolation valve. 'Ihese are notor operated valves with the motor outside the shield wall. Manual valve wheels are also
- v' provided at the motor so that the valve may be closed even in the event of motor operator failure.
The FNP has been designed so that post-accident maintenance may be per-formed on either of the two RHR pumps 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 ptrap rooms.
Airborne activity released to the 9HR subcompartment would be removed by the annulus ventilation systen which maintains a negative pressure in the subcomp.:tment. Additionally, the design basis for equipment important to 4
safety ir.cludes a requirement for satisfactory operation following post-accident radiation exposure.
Source terms tased on Regulatory Guide 1.4 for release to the containment are given in Section 12.1 of the PDR. Doses and dose rates outside the shield building as a function of time after the accident (based on those q source terms) are given in Section 15.4 of the Plant Design Report.
t #
%J^
C-61 i l
l
[~') Environmental. qualification conditions for WP equiFant are given in Table 3.11-1 of the Plant Design Report. The radiation doses listed are based on source terms developed in accordance with Regulatory Guide 1.4. During the detailed design of the plant, a more detailed listing will be prepared.
To sunnarize, the existirg design philosophy for controllirg raatoactive water and airborne activity following an accident involving core damage is to' isolate non-essential systems which could transport post-accident radioactivity outside containment. Systems outside the containment which are needed followirg an accident for core coolirg or containment atmosphere cooling are located within shielded subcompartments, which are part of each separate safeguards empartment. These subcampartments 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 work locations near systems which could potentially contain highly radioactive water. Vital areas on the WP have been designed t'o meet the requirements of NURD3-0737.
f C-62
- REI3ULATION 10CFR50.34 (e) (2) (viii) 7m
.(v
)
Subject:
Post-Accident Sampling To satisfy the following requirements, the application shall provide sufficient information to derconstrate - that the required actions will be satisfactorily completed by the operating license stage. mis information is of the type customarily required to satisfy 10CFR50.35 (a) (2) or to address m resolved generic safety issues.
Provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain TID 14844 source tenn radioactive materials without radiation exposures to any individual exceeding 5 rem to the whole-body or 75 rem to the extremities. Materials to be analyzed and quantified include certain radionuclides that are indicators of the degree of core damage (e.g . , noble gases, lodines and cesitzns, and non-volatile isotopes) , hydcogen in the containment atmo-sphere, dissolved gases, chloride, and boron concentrations. (II.B.3)
OFFSHORE POWER SYSTEMS RESPONSE Westinghouse has developed a design for a Post-Accident Sampling System (PASS) that .will provide for the acquisition and analysis of the required liquid and gas samples with a minimtza of radiation exposure to personnel in V compliance with the NURD3-0737 requirements. We present FNP system (as described in Section 9.3.2.2 of the PDR) will be modified to incurporate the Westinghouse or a similar post-accident sampling systen. The Westing-house system, with some modifications for the FNP design, is described below.
The modified PASS will permit obtaining liquid and gas samples within i hour of an accident condition which releases an assumed TID 14844 source term radioactive materials without radiation exposures to any individual exceeding 5 rem to the whole body or 75 rem to the extremities. It will provide for. remote manual collection and p ocessing of liquid and gas samples with maximtzn practical use of online instrumental methods of analysis. Provisions are also included for remote dilution of liquid and gas samples and for degassing of liquid samples. Additional shielding will be added as necessary to satisfy the dose limitation.
m
'l )
v C-63
~
a The followirg la-contairenent sample points are included:
73
-\ l o Eample points on two reactor coolant hot legs o Pressurizer 1iguid sample o' Pressurizer vapor space sample o Reactor vessel vent sarple o Containment sump sample o Upper reactor compartment at outlet of two air recirculation fans o Iower reactor compartment at three ice condenser door locations The in-containment sampling portion will be designed to minimize the amount of sample collection and analysis equipment located inside the containment and thus nuximize accessibility to the components. Motive force for the reactor coolant samples will be supplied by either system pressure (if available) or by redundant positive displacement pumps in-containrr.ent. 'Ihe containment step sample will be pm ped by means of the same pasitivo displacement pumps. A single bellows pump will supply the driving force for the containment atmosphere samples ard will be located outside containment.
im
\
I
Isolation valves in the system will be remote-manual, solenoid-operated with electrical p)wer beirn supplied frm redundant Class 1-E buses. The liquid and gas sample lines may be flushed with demineralized water and dry nitrogen respectively frcn outside containment. The flush liquids as well
~
as excess sample flows are returned to either the containment sump or the pressurizer relief tank (for liquids) or to the containment atmosphere (for gas). Men the system is being used under normal plant operating condi-tions, the excess liquid samples ard flush solution may be routed to the waste holdup tanks of WrL.
The samples descr" a above are piped to a special sample analysis station where samples are icessed both by remote manual and online techniques.
The sample analysis station will consist of a concrete cubicle located close to the containment outside wall. The cubicle provides capability for remote, on-line analysis of some parameters. All of the sample handling equipnent and analytical monitors for remote analysis will be located
. Inside the cubicle with controls and readouts located on a remote panel. A
{}1
\_-
C-64
-glove box within the cubicle provides the capability to obtain grab samples
( ) for analysis. Shielding will .be provided sufficient for short-term access to the cubicle for sample removal.
~
The following types of grab samples will be obtainable from the glove box.
o Undiluted pressurized liquid sample o Undiluted containment atmosphere sample o Diluted liquid sample o Diluted containment atmosphere sample
~
o Degassed undiluted liquid sample The following chemical analyses will be performed on-line by flow-through instrumentation remotely. controlled and monitored.
o Oontainment atmosphere hydrogen concentration o Liquid soluble boron concentration o Liquid pH
(]
\
o Liquid soluble chloride concentration
"/ o -Liquid and gas gross activity MODES OF OPERATION The liquid and gas sampling systems are independent. Each is designed so that a sample of liquid or gas can be remotely processed and analyzed in one of several different ways. Prior to analyzing a liquid or a gas sample, its gross radioactivity level can be checked by circulatirg the sample through a radiation tronitoring station, the excess sample being returned to the containment. The methods by which samples are obtained and
- analyzed are outlined below.
Undiluted Samples (Liquid or Gas)
To obtain an mdiluted liquid or gas sample, the sample point is selected by appropriate valve lineup, a sample panp is started and the sample stream is circulated out of containment, through a shielded sample vial, and back
-(9 G'
to the containment. The sample vial is then isolated remotely ard the
}I c-65
. entire piping syste external to the sample vial is purged with flush water l ) or dry nitrogen as appropriate. %e sample vial is then manually removed fra the system, put into a shielded transport cask inside the shielded glove box, and . renv ,ed for of f-site analysis. tis sample may be analyzed for certain r alonuclides which are indicators of the degree of core damage (e.g., noble gases, iodines, cesiums and non-volatile isotopes) .
Diluted Samples (Liquid or Gas)
A. diluted sample of liquid or gas is obtained in much the same way as for the undiluted sample except an additional dilution operation is performed.
Once the mdiluted sample has been drawn int.o a calibrated sample vial, isolated, aM the systen flushed, the sample vial is lined up to a re-circulating loop containing a known volume of pure diluent (water or nitrogen) ard the sample is mixed with the diluent by a circulating ptinp.
The ratio of initial sample volume to dilution system volume is predeter-mined by calibration (typically 1:100). The required order of dilution is obtained remotely by successively isolating the sample vial, purging and refilling the dilution loop and remixirg by circulation. A radiation C monitor is used to verify the number of dilution cycles required. We sample vial is finally isolated and removed as for the undiluted sample.
Depressurized/ Degassed Liquid Samples A liquid sample is degassed by collecting it in a snall stripping column which is part of a recirculating loop containirg a shielded sample vial and a pump. We liquid is recirculated while nitrogen is bubbled through the strippiry column. This system operates at atmospheric pressure. The depressurized, degassed sample is then either collected as before in the sample vial ard removed for analysis or the liquid sample is circulated to ,
one of the remotely-operated flow-through analyzers included in the sampliry station.
Remote Soluble Boron Analysis The degassed and depressurized liquid sample is circulated through a flow-through analyzer which measures the boron concentration by a neutron attenuation technique. F.eadout is on the remote Sample System Control Panel.
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f Remote Soluble Chloride Analysis The degassed and depressurized liquid sample is circulated through a cell containirg a specific-ion electrode systen for chloride. Readout is on the renote Sample System Control Panel.
Remote Solution pH Analysis The degassed and depressurized liquid sample is circulated through a cell containirg a hydrogen-ion electrode systen. Readout is on the remote Sample System Control Panel.
Remote Gross Radiation Level The gross radiation levels of either liquid or gas samples is determined by a wide range radiation monitorirg station on the respective sample inlet lines. Wese monitoring stations will be located outside containment, and the readout is on the remote Sample System Control Panel.
Remote Aanospheric Hydrogen Concentration The concentration of hydrogen in the containment atmosphere is determined by passing the atmospheric sample stream through a thermal conductivity hydrogen monitor located outside containment. The gas stream then returns to the containment. Readout is on the remote Sample System Control Panel.
Handling of Excess Liquid Samples and Flush Samples The excess liquid sample and flush volumes could be routed either to the pressurizer relief tank or stunp inside cor'tainment or to the plant waste holdup tank in the auxiliary building. We latter would be used only when radiation levels were low as durirg normal plant operations. The excess gas and nitrogen purge would be discharged into the containment atmosphere.
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O C-67
REI3UIATION 10CFR50.34 (e) (2) (ix)
(
Subject:
Hydrogen Control System To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily cx>mpleted by the operating license stage. 'Ihis information is of the type customarily required to satisfy 10CER50.35 (a) (2) or to c.ddress mresolved generic safety issues.
Provide a systs for hydrogen control that can safely accormodate hydrogen generated by the equivalent of a 100% fuel-clad metal water reaction.
(II .B.8)
OFFSHORE POWER SYSTE.MS RESPONSE Offshore Power Systems will install a Hydrogen Ignition System (HIS) in the containment of the Floating Nuclear Plant capable of handling hydrogen generated by the equivalent of a 100% fuel-clad metal water reaction. 'Ihe HIS enploys districuted hydrogen ignition sources located throughout the ,
containment building. Tne ignition sources are of the thermal element (or glow plug) type. Shortly af ter activation these glow plugs reach tempera-S tures which are adequate to reliably ignite combustible hydrogen / steam / air (d mixtures.
OPS will incorporate the results of industry and NRC research programs, such as AIF-IDCOR, EPRI, Sandia, Livermore, Fenwall, etc., which will demonstrate the ignition characteristics of these glow plug igniters.
Within two years after receipt of the manufacturing license, design details, describing the Hydrogen Ignition System, will be provided to the NBC for review; these design details, includirg test data and analyses, will illustrate that the hydrogen control systems will perform in the manner required by the NRC position.
The HIS consists of approximately 62 glow plug igniter assemblies located in 31 distinct locations in the containment buildirg. Each igniter assembly consists of a glow plug and a control p)wer transformer similar to those used for Sequoyah and McGuire Nuclear Stations. The glow pity and trans-former are mounted in a sealed metal box housing tich employs heat shields to limit the tmperature rise inside the box and a drip shield to reduce (a) v C-68
direct moisture impingement on the thermal elment. We igniter boxes are
-$ seismically mounted to prevent damage to Category 1 equipment.
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Each designated containment location has two igniter assemblies powered from separate emergency power trains. We igniters are powered from 120 VAC buses. In the event of loss of offsite power the igniter assemblies will be supplied p)wer from the emergency diesel generators. Cables of the two divisions are physically and electrically separated.
Glow plug igniters are located throughout the containment to promote hydrogen burning in all areas prior to reaching hydroegn concentrations sufficient to threaten containment integrity. We glow plugs within each compartment or subcompartment are located near the ceiling since, if there is any nonr-uniform distribution of hydrogen, the higher concentrations would be expected to exist there. Mounting the assemblies near the ceiling also minimizes interference with the operation of equipment and provides some degree of physical protection for. the protruding glow plug. Se igniters are located in the incore instrument tubing chase, pipe chases, stear generator enclosures, pressurizer enclosure, instrument room, below the operating floor, in the ice condenser upper plenum below the top deck doors, and in the containment dome.
The Hydrogen Ignition System will be automatically actuated in accident situations which have- the potential for the generation of excessive quantities of hydrogen. To effectively perform their intended function the HIS igniters must be energized and at operating temperature before sig-nificant amounts of hydrogen are releaswl to the containment atmosphere.
Therefore, actuation of the HIS will occur once the potential for excessive hydrogen generation is established and will not be dependent upon a mea-surement of the hydrogen concentration inside containment. Signals for automatic initiation of the HIS and the setpoints will be selected during the developnent of syste design details. Each division of the HIS can also be manually actuated from the control room.
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v C-69
4 REGULATION 10CFR50.34 (e) (2) (x)
( j
Subject:
SV and RV Qualification v
To satisfy the following requirements, the application shall provide sufficient information - to demonstrate that the required actions will be satisfactorily completed by the operating license stage. %is information is of the type customarily required to satisfy 10CFR50.35(a) (2) or to address unresolved generic safety issues.
Provide a test program and associated model developnent and conduct tests to qualify reactor coolant system relief and safety valves and, for PWR's,
- PORV block valves, for all fluid conditions expected under operating conditions, transients and accidents. . Consideration of anticipated transients without scram (A WS) conditions shall be included in the test program. Actual testing under AWS conditions need not be carried out until subsequent phases of the test program are developed and not before issuance of an AWS rule. (II.D.1)
OFFSHORE POWER SYSTEMS RESPONSE Safety and relief valve testing is being conducted generically in the EPRI testing program. We EPRI Program Plan was presented to the NRC on December 17, 1979 and further discussed with the NRC on February 25, 1980. It is h
G presently expected that the EPRI test program will be completed by July 1, 1981. Of fshore Power Systems will carefully monitor generic testing and will demonstrate applicability of the generic test to the Floating Nuc'. ear Plant.. We effect of as-built relief and safety valve discharge piping on valve operability will be analyzed and the piping and supports will be designed for loads resulting from design basis transients and accidents.
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v C-70
1 RB3UIATION 10CFR50.34 (e) (2) (xi)
Subject:
SV and R/ Position Indication J-To satisfy' the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. Wis information is of the type customarily required to satisfy 10CER50.35(a) (2) or to address mresolved generic safety issues.
Provide direct indication of relief and safety valve position (open or closed) in the control room. (II.D.3)
OFFSHORE POWER SYSTESM RESPONSE 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 powered frm vital instrument buses. In addition, a position disagreement light / alarm prominently (O ; displays a failure of the PORV to achieve the last position concanded.
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- 2. We temperature downstream of the PORVs and safety valves is displayed on the control board and high temperature alarms are provided.
- 3. We pressurizer relief tank has temperature, pressure and fluid level
^
indication and alarms on the main control board.
- 4. High pressurizer pressure alarms in the control Room.
Offshore Power Systes is presently evaluatirg alternate methods to provide safety valve Fosition indication. One such system has been developed and is described below.
Westinghouse has developed an acoustic leak monitoring system that will provide flow indication downstream of the safety valves and thus satisfy the NRC requirements for leakage detection. We system operates on the i
u-C-71
~ ._ - . . . _ . . _
principle that turbulent, high pressure flow through an orifice generates an acoustic signal which is transmitted' throughout the reactor coolant
\ ' system. 'Ihe monitoring system will detect acoustic signals' and thus determine valve position. The FNP will incorporate either the Westinghouse ,
acoustic leak monitor .or stem mounted limit switches af ter acceptance by
-the NRC.
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REGULATION 10CFR50.34 (e) (2) (xii)
,,y
Subject:
Auxiliary Feedwater Automatic Initiation / Flow Indication
. ( s) x.
To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily cmpleted by the operating license stage. This information is of the type customarily required to satisfy 10CFR50.35 (a) (2) or to address unresolved generic safety issues.
Provide automatic and manual auxiliary feedwater (AEW) system initiation,
- and provide auxiliary feedwater systs flow indication in the control room.
(Applicable to PWR's only) (II . E.1. 2)
OFFSHORE POWER SYSTESMS RESPONSE Auxiliary Feedwater System auto-start provisions are detailed in the response to 10CFR50.34(e) (1) (ii) .
Auxiliary feedwater flow channels, with an accuracy of better than the required +10%, will be displayed on the main control board. Each channel of flow inr*.rumentation is powered from its respective associated Class lE ;
instrument p wer supply.
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C-73 l l
REGULATION 10CFR50.34(e) (2) (xiii)
)
Subject:
Pressurizer Power Supplies q,/
To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. Wis information is of . the type customarily required to satisfy 10CER50.35 (a) (2) or to address unresolved generic safety issues.
Provide pressurizer heater power supply and associated motive and control power interfaces sufficient to establish and maintain natural circulation in hot standby conditions with only .onsite pwer available. (Applicable to PWR's only) (II . E. 3.1)
OFFSHCRE POhER SYSTEMS RESPONSE The ENP design provides the following features which assure a continued supply of power for the followirg plant components essential to natural circulation flow.
- 1. Pressurizer heaters O)
( The total pressurizer heater capacity for the ENP is 1800 W. Four separate backup heater groups (346 W each) are supplied directly from 4 independent and redundant safety class 480V switchgear buses. Each bus is supplied from its respective standby diesel generator following a loss of offsite power. %e control group (416 W) is supplied from a non-safety class 480V bus which could be supplied frcrn a diesel-generator bus within several minutes following a loss of offsite power, in the unlikely event that this should become necessary.
Each independent backup group is large enough to maintain rutural circulation in the hot standby condition.
We Class 1E circuit breakers supplying each of the tackup groups are tripped open on either a safety injection (SI) or loss of offsite power actuation signal.
t wj C-74
%e heaters can be manually loaded onto the bus from the main control ;
(~T' '
board after SI is reset and loads required in the initial stages of the s
is incident are no longer required. Sufficient diesel generator capacity is provided to supply the minimum required nmber of heaters in the time required. Diesel generator instrumentation is provided to prevent overloading a diesel generator with these heater loads.
OPS will provide the owner with the necessary procedures for energizing the pressurizer heaters, including procedures that might be required for load shedding. ,
- 2. Power Operated Relief Valves (PORV's)
Each PORV is supplied with operating air from a separate Safety Class-3 air system which is available follcwing a loss of offsite power. Each PORV pilot solenoid is supplied from independent and redundant 125V DC sources, which are also available following a loss of offsite power.
%e PORV's are controlled from the main control board. Both PORV's f3 fail closed on loss of motive or control power.
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- 3. PORV Block Valves The PORV block valves are supplied from stotor control centers which are readily energized f rom a corresponding standby diesel generator following a loss of offsite power. We PORV block valves are controlled '
from the main control board. %us the PORV block valves can also be operated following a loss of offsite power.
- 4. Pressuriner Level Indication Channels All of the pressurizer level indication channels are derived (and isolated) from their respective protection channels. We instrment loop power supplies for these protection channels (including the isolated outputs) are supplied from their respective class lE Instru-ment buses. %us level indication is available following a loss of fm offsite power.
. \v)
C-75
_RB3UIATION 10CFR50.34 (e) (2) (xiv)
[].
Subject:
Containment Isolation Systems C/
To satisfy the following requirements, the application shall provide sufficient- information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. 'Ihis information is of the type - custmarily required to satisfy 10CFR50.35(a) (2) or to address mresolved generic safety issues.
Provide containment isolation systems that: (II.E.4.2)
(A) Erisure all non-essential systems are isolated automatically by the containment isolation system, (B). for each non-essential penetration (expect instrument lines) have two isolation barriers in series, J
(C) .do not result in reopening of the containment isolation valves on resettirg of the isolation signal, (D) utilize a containment set point pressure for initiating containment
- isolation as low as is empatible with normal operation, I
(E) include automatic closing on a high radiation signal for all systems that provide a path to the environs.
4 (O OFFSHORE POWER SYSTEMS RESPONSE The ENP containment isolation system, described in Section 6.2 of the Plant Design Report satisfies the acceptance criteria of Standard Review Plan 6.2.4. Containment isolation system features specifically required by this rule are addressed below:
A. Phase A isolation (T signal) results in the isolation of all non-essential systems penetratirg the containment with the exception of component cooling water lines to the reactor coolant pumps and the lower cmpartment fan coolers which are closed by Phase B isolation (P signal) .
Phase A isolation provides for diversity in parameters sensed as well as beirq autmatically actuated any time a safety injection signal (S (n) v C-76
signal) is initated. Phase A isolation is initiated from the following i
process variables:
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(1) High steam flow coincident with low steam line pressure or lo-lo T,yg.
(2) .High steam line differential pressure (3) tow pressurizer pressure (4) High containment pressure (5) Manual initiation Phase B isolation is initiated from hi-hi containment pressure or manually. Although it is not automatically generated by diverse means, the P signal can only be generated after the T signal, which is i diverse, has been initiated. In addition to initiating Phase B isolation, the P signal also is used to initiate containment spray.
Of fshore Power Systems has given careful consideration to the systems penetrating the containment Wich are required to mitigate the conse-p quences of a loss of coolant accident, or any accident callirg for
(- containment isolation. We systems which are required to operate followirg the accidents are as follows:
- Safety Injection System Residual Heat Removal System (supply lines to cold legs)
Contairinent Spray System (includiry recirculation stznp 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 Systen, the containment, and the Main .eam System. Wese systems, or parts of these systems required for Fost-accident cooling, do not receive any containment isolation signal.
l The following systems are not essential to mitigate the consequences of a design basis loss of coolant accident but are considered desirable in y assisting in plant recovery from accidents of lower magnitude than a 1
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-design basis accident. '1 hey are not part of Phase A isolation, but L]j
[ instead 'are isolated by the P signal (Phase B isolation) .
Component Cooling Water System (supply and return lines to RCP thermal barrier cooling)
Component ~ Cooling Water System (cooling water flow to the lower compartment fan coolers)
The following systems have been determined to be non-essential and are isolated by the T signal- (Phase A):
- Chemical and Volume Control System
- Post-Accident Sampling System
- Radiation Monitoring System (containment air sample lines)
- Nuclear Sampling System
- Containment Ventilation System
- Post-Accident Containment Ventilation System
- Liquid Waste Treatment System
- Service Air Systen - Instrument Air System D -
Emergency Air System
- Ice Condenser Refrigeration System
- Non-Essential Service Water System i - 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
- 8. All non-essential lines are properly isolated with two barriers in series following the -initiation of a containment isolation signal. In addition to the systens which are listed as being subject to Phase A isolation, other- non-essential systens or lines which penetrate containment have normally closed manual isolation valves subject to administrative control.
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- v C-78
7, C. Containment. isolaiion reset logic requires deliberate and specific
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operator action before an isolated line can be reopened. We followireg control features are provided for containnent isolation valves:
-(1) The containment isolation signals override all other automatic control signals.
(2) The valves will remain in the closed position if the initiating signal is reset.
(3) Each valve can be opened or closed manually af ter the appropriate containment isolation signals are reset. -
(4) Any valves that are normally operated in an automatic mode (for non-cafety functions) are also automatically transferred to manual mode by the isolation signal. his prec1t des automatic opening of containment isolation valves subsequent to reset of the initiating isolation signal.
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t d D. During Floating Nuclear Plant final design, the containment high . pres-sure trip point will be reviewed and adjusted downward (if necessary) to the minimum compatible with conditions not requiring automatic containment isolation.
l E. Systems that provide an open path from the containment to the environs are the containment purge - supply and vent systems. In the present design, isolation valves in these systems are automatically closed on high radiation. This design will be upgraded to provide closure on a safety grade high radiation _ signal.
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l I
l
~ REGUIATION 10CFR50.34 (e) (2) (xv,
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G/
Subject:
Containment Purge / Vent Systems To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. Wis information is of the type customarily required to satisfy 10CFR50.35(a) (2) or to address mresolved generic safety issues.
Provide a capability for containment purging / venting designed to minimize the purging time consistent with AIARA principles for occupational expo-sure. Provide and demonstrate high assurance tb1 the purge system will reliably' isolate mder accident conditions. (II.E.4.4)
OFFSHORE POWER SYSTESMS RESPONSE Containment purging is performed by either of two systems (1) Containment Pre-access Filtration and Purge System or (2) Post Accident Containment Venting (Purge) System. Rese systems, which are described in Section 6.2 f
and 9.4 of the Plant Design Report, are designed in ccanpliance with Standard Review Plan 6.2.4 and Branch Technical Position 6-4.
) The Containment Pre-access Filtration and Purge System provides a continu-v ous purge function at a restricted flow during normal plant operation via an 8 inch diameter supply and 8 inch diameter exhaust penetration. %e system is capable of purging via a 42 inch diameter penetration; how2ver, the plant owner will be required to restrict such operation to refueling operations or when the plant is in cold shutdown. Procedures will require that the 42 inch diamater purge valves remain closed during normal power operation. Each of the two purge lines will be provided with separate containment penetrations, each isolated by two valves in seriesIII .
(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 9.4-6, Sheet 4) . An additional 8 inch penetration and isolation valve will be provided with a containment isolation valve inside and outside the containment shell. Wis design change provides separation of the two purge functions and maximizes the reliability of the containment isolation function.
(3 L)
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The Post Accident Containment Venting Systen provides for hydrogen purge. l
('T te system provides a controlled and filtered containment prge capability t
L) by releasirq air to the annulus at a maximun rate of 50 SCFM.
Isolation valves are designed to operate against accident pressures and to !
maintain bubble air-tight closures while perfonning their intended func- f tion. We isolation valves have a 2 to 5 second closure time. %e contain- {
ment pre-access filtration ard purge system containment isolation valves f are described in Section 6.2.3.2 of the PDR. We 8 inch isolation valves l
will be included in the operability assurance p? m described in Section !
3.9.2.4 of the Plant Design Report.
f C-81
REGULATION 10C*R50.34(e) (2) (xvii)
(3 I t.
Subject:
Containment Instrmentation To satisfy the following requirements, the application shall provide sufficlent information to demonstrate that the requitM actions will be satisfactorily completed by the operating license stage. %is information is of the type customarily required to satisfy 10CER50.35(a) (2) or to address mresolved generic safety issues.
Provide instrumentation to measure, record and readout in the control room:
(a) containment pressure, (b) containment water level, (c) containment hydrogen concentration, (d) containment radiation intensity (high level),
and (e) noble gas effluents at all potential, accident release points.
Provide for continuous sampling of radioactive iodines and particulates in gaseous effluent's from all potential accident release points, and for onsite capability to analyze and measure these samples. (II.F.1) 0FFSHORE POWER SYSTEMS RESPONSE A. Containment Pressure:
To comply with the requirement for containment pressure monitoring, two additional wide range containment pressure channels will be incorporat-
/Q ed into the FNP. %ese additional channels will range from minus 5 psig to 60 psig (4 times design pressure) . The channels will meet the design requirements of Regulatory Guide 1.97, Rev. 2 (Dec.1980) . We present instrument rarge 0-18 psig.
B. Containment Water Level:
As described in Section 6.2.2.7 of the PDR, the Floating Nuclear Plant design does not incorporate a conventional containment smp as such.
Instead, the containment lower compartment will collect a sufficient volme of water followirg the injection phase of safety injection to
, allow recirculation. Redundant .:afety grade containment water level (wide range) measurement is currently provided and displayed in the Control Room. We range of these level channels will be increased to cover an elevation equivalent to an 800,000 gallon accmulation, a quantity Witch includes ice melt and UHI accmulator injection.
C-82 3 - m 9 q -:- e -.+ -
In addition, Class IE (narrow range) level channels will be provided for the local liquid waste treatment systen sunp at the 103 foot elevation in accordance with this requirement. ihese channels will also be used as part of the Reactor Coolant System Leak Detection System.
These channels will meet the design requirements of Regulatory Guide 1.89.
C. . Containment Hydrogen Concentration:
A continuous indication of hydrogen concentration in the containment atmosphere will be provided in the Control Room. Measurement capability will be provided over the range 0% to 10% hydrogen concentration under both positive and negative ambient pressure. Hydrogen monitors which can perform this function are presently available but have not yet been qualified to IEEE-323 and IEEE-344. During FNP final design Of fshore Powr Systems will select hydrogen monitoring instrumentation which is acceptable to the NRC.
D. Containment Radiation Intensity (High Range)
/T 1
The current FNP design for the redundant containment area monitors specifies a range of 10-1 to 107Rad /Hr of gamma radiation. %is range complies with the requirement specified in NUREG 0737 and Regulatory Guide 1.97, Rev. 2. It should be noted that these detectors for the FNP design are mounted on the outer surface of the steel containment but may be considered as ";n-containment" relative to compliance with this requirement. The attenuation by the steel shell will be factored into the calibration of the monitors. Interference from non-containment sources will be eliminated by proper shielding of the detectors. The monitors will be located such that they are widely separated, view a a
large fraction of the upper compartment and have an unobstructed view down to the operating deck. Although the detectors will be mounted high on the containment dome, they can be readily accessed for maintenance using platforms and ladders already included in the, ENP design for inservice inspection of the containment shell. Mounting the detectors Q
O -
C-83
l outside the steel contairment serves two safety related purposes: 1)
, the need for containment cable penetra lons is eliminated, and, 2) the
%) monitors will experience less severe postulated accident environ-mentional conditions, (i.e., temperature, humidity, and pressure) .
E. Airborne Radiological Effluent Monitors (Refer also to PDR Sections 11.4 and 12.2):
The current ENP design includes monitors to detect airborne effluent from three potential release points: the plant vent, the condenser air ejectors, and the Annulus Ventilation System AFS) exhaust vent. Also included is a passive plant vent charcoal cartridge for iodine detection.
Two pairs of radiogas and particulate monitors are used to monitor the plant vent and the AFS vent. Each particulate monitor consists of a high-range and low-range channel. Each of these monitor pairs can be selected to the following three sample locations
- 1) Plant vent
- 2) AFS vent
- 3) Containment atmosphere Following an accioent ("S" signal) one pair of monitors is assigned automatically to continuously monitor the plant vent and the other is assigned to continuously monitor the AFS vent. This arrangement satisfies the requirements of NURED-0737 and Regulatory Guide 1.97, Revision 2.
A radiogas monitor will have an upper detection limit of 10 4 C1/cc of Xe-133. The lower end of the range, will be sensitive to a concentra-tion as low as 10- Ci/ce, in order to monitor normal plant releases.
The 12 decades of response will be obtained with a multi range (3 levels) detector. 'Ihe high range particulate monitor will be replaced by a passive filter cartridge that can be removed for analysis. The low ;
range prticulate monitor in the present design will be retained.
U' C-84
'In order to allow safe collection and analysis of the plant vent
()
j3 charcoal 'arx! filter cartridges durirg armi imediately followirg an accident, provisions will be made in the plant design to do the followirg (or the equivalent):
- 1) place the high rarge cartridges in the post accident sampling room and route the sample lines fram the plant vent to the post accident sampling room, Wilch is shielded and habitable imediately follow-Irg an accident, or I
- 2) design and provide a system for safe remote collection of the cartridges for analysis in the post-accident samplirn room.
The mota practical of these alternatives will be selected as the design Golves. All effluent monitoriry channels will have assured. p wer supplies, independent of offsite pwer (i.e., Class IE or Class 1E Associated Power) . .
The current FNP design includes a noble gas monitor on the condenser
- air ejector discharge. The monitor will be upgraded to meet the requirements of NUREG 0737 and Reg. Guide 1.97, Rev. 2 includirg increasiry the rarge to 10 -6 to 10 5 C1/cc of Xe-133.
In order to meet the new requirement for nonitoring releases from the atmospheric steam dtsnp valves or the main steam safety valves, four 4
monitors will be added to the plant, one for each main steam line.
These monitors will view the main steam line upstream of the main steam stop valve and will have a range of 10 -1 to 10 3 Ci/cc of steam.
- \g C-85
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p- m RfDULATION 10Cm50.34 (e) (2) (xvi 11)
[l
Subject:
Core Coolirg Instrmentation O
To satisfy the following requirements, the application shall provide suf'icient information to demonstrate that the required actions will be
-satisfactorily completed by the operating licenae stage. 'Ihis information ,
is of the type custamarily required to satisfy 10cm50.35(a)(2) or to address mresolved generic safety issues.
Provide instruments that provide in the control room an unambiguous indication of inadequate core cooling, such as primary coolant saturation meters in PWR's, and a suitable combination of signals frm indicators of coolant level in the reactor vessel and incore thermocouples in PWR's and BWR's. (II .F.2)
OFFSHORE POWER SYSTEMS RESPONSE OPS continues to evaluate options developed by the Westinghouse Owners' Group regardirg instrumentation for detection of inadequate core cooling.
Of the options presented in NUROG 0737, the one preferred will be selected ducirn final design. Procedures used by the operator to recognize in-adequate core cooling will be developed based on the instrumentation provided in the final FNP design.
Subcooling Meter An approach being considered by OPS is to provide dedicated, redundant, microprocessor-based subcoolirg meter channels with prominent displays on both the Unit Control Console and the Safety Center Panel (Refer to the attachment in the response to 10Cm50.34 (e) (2) (iii) for a description of the ENP Control Board) . Each of these meters would provide a continuous indication of margin to saturated conditions. The operator could manually select a display of margin to saturation based on either the auctioneered high incore thermocouple or the auctioneered high loop Thot r T eold*
Auctioneered low reactor coolant system pressure is used for the Tsat calculation by the microprocessor. Inputs to this systs would utilize redundant safety grade hot leg and cold leg temperatures and reactor coolant syste pressure channels. In addition, approximately 16 in-core l l thermocouple inputs (together with reference junction temperature inputs)
-would be utilized.
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-Iko setpoints would be ' utilized to alarm 1) off-normal conditions and 2)
) approach to loss of mre moling. Individual sensor channels will' also be (O accessible for display in the control room.
Table C-4' provides a sumary of tentative design information for the ENP subcoolirg meter.
Additional Instrumentation s
Offshore Pawer Systems is in the process of evalmting the various methods of neasurirg reactor vessel level that 'have been investigated by the Westinghouso owners' Group. We current state of . the art appears to favor the use of differential pressure measurement as the best method of deter-mining vessel level. mis method would utilize sealed reference legs and would range frm the vessel top (usiry ~ an existirg penetration) to the vessel bottom (using an incore instrumentation 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 compensation. The entire level V 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 providing an unambiguous indication of inadequate core cooliry and an unambiguous indication for vessel venting is being evaluated, with attention given to all possible phenomem that could adversely affect the system.
\
v' i
C-87 l
,r%' TABIE C-4 Q]
INIORMATIN REQUIRED W THE SUBCOOLING MEh3 Display Information Displayed Tsat-T, dere 'T' is based on (T-Tsat, Tsat, Press., ate.) either incore or RTD temperatures (Note 1)
Display Type (Analog, Digital, CRT) Analoj (Note 1)
Continuous or on Demand Continuous and on demand Sirgle or Redundant Display Redundant location of Display Unit (bntrol Console and Safety Center Alarms (include setpoints) (Note 2)
Overall uncertainty ( F) (Later)
Range of Display 4v F Superheat to 200 F Su'xooled C
()\ Qualifications (reismic, environmental)
IEEE-344, -32's Based on NUREI3 0737 j Calculator Type (process computer, dedicated Dedicated Digital 1 digital or analog calc.)-
If process computer is used specify Not Applicable availability. (% of time)
Sirgle or redundant calculators Redundant Y
Selection Lo31 c (highest T., Auctioneered high incore temp.
lowest press.) or Auctioneered high RCS temp.
vs: Auctioneered low RCS press.
Qualifications (seismic, IEEE-344, -323 environmental) Based on NUREG 0737 Calculational Teu que ' Steam Tables (Steam Tables, Ebnctional Fit, ranges) 4 O
i o C-88 ;
Input
/h Temperature (R1D's or T/C s) 8 8 incore T/C's (2 per quadrant)
V 2 Hot Leg R1D's (per loop)
Temperature (number of sensors 2 Cold Leg R1D's (per loop) and locations T/C ref. junc. RrD's Pange of temperature sensors Incore T/C's = 150 -2300 F RCS RrD's: 0 -700 F Uncertainty
- of temperature sensors (Later)
( F at 1)
Pressure (specify instrument used) (Note 3)
Pressure (number of sensors and 2 (RC3 Hot Legs) locations)
Range of Pressure sensors 0-3000 psi Uncertainty
- of pressure (Later) sensors (PSI at 1)
Qualifications (seismic, IEEE-344, -323 environmental) Based on NURm 0737 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 N7TES:
- 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 dif ferent indicators.
- 2. Two setpoints will be chosen to indicate: a) off-normal conditions (50 F nominal) and b) approach to loss of core cooling (later) .
- 3. Qualified instrument will be specified later.
v C-89 a_ _ _ -. - - _ _ . - , _
. _REGUIATION 10CFR50.34 (e) (2) (xix) v-
}
Subject:
' Post-Accident Instrtmentation To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the- requi ed actions will be satisfactorily completed by the operating license stage. Wis information is of the type customarily required to satisfy 10Cm50.35(a) (2) or to address mresolved generic safety issues.
Provide instrumentation adequate for monitorirq plant conditions following an accident that includes core damge.
OFFSHORE POWER SYSTEMS RESPONSE Of fshore Power Systems has comitted that the WP design for Post-Accident Monitority will comply with Regulatory Guide 1.97 Revision 1. The present FNP design includes much of the instrumentation required to meet Revision
- 2. Those recomendations of Revision 2 not already in the current design will be incorporated or a suitable alternate will be provided for those items that challenge the state--of-the-art. Design information for alternate instrumer.tation and justification of its adequacy will be sutaitted for NRC O review pr.f or to equipnent procurement.
D
\
U C-90
r REGUIATIch 10CFR50.34 (e) (2) (xx)
Subject:
Power for PORV, Block Valves, Level Instrisnentation To satisfy the following requirements, the application shall provide
'^
sufficient information to demonstrate that the - required actions will be satisfactorily completed by the operating ' license stage. Wis information is of the type customarily required to satisfy 10CFR50.35(a) (2) or to address unresolved generic safety issues.
Provide power supplies for pressurizer relief valves, block valves, and level indicators such that: (A) level indicators are powered from vital buses; (B) motive and control power connectiou to the emergency power sources are through devices qualified in accordance with requirements applicable to systems important to safety and (C) electric power is provided from energency power sources. (Applicable to PWR's only) . (II.G.1) 0FDSHORE POWER SYSTEMS RESPONSE (A) Pressurizer Level Indication Channels All of the pressurizer level indication channels are derived (and isolated) from their respective protection channels. We instrument loop power supplies for these protection channels (including the O isolated outputs) are supplied from their respective Class lE Instru-1]
ment buses. Thus level indication is available following a loss of offsite power.
(B) , (C) Motive and Control Power Sources and Connections '
(1) Power Operated Relief Valves (PORV's)
Each PORV is supplied with operating air from a separate Safety Class-3 air system which is available following a loss of offsite power. Each PORV pilot solenoid is Class lE and is supplied from independent and redundant Class 'lE 125V DC sources, which are also available following a loss of offsite power. We PORV's are controlled from the main control board. Both PORV's fail closed on loss of motive or control power.
U C-91
4'
, I (2) PORV Block Valves '
i The PORV ~ block valves 'are supplied from qualified ' motor control t centers which are readily energized fra a- correspondirq standby ;
diesel generator following a loss - of offsite power. %e PORV ,
I block valves are controlled frm the main control board. Thus the PORV block valves can also be operated following a loss of offsite power. !
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_RE_UULATICd 10CFR50.34 (e) (2) (xxv)
Subject:
Post-Accident Support Facilities (V')
To ~ satisfy the following ' requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. %1s information is of the e/pe customarily required to satisfy 10CER50.35 (a) (2) or to address unresolved generic safety issues.
Provide a Technical Support Center, an onsite Operational Support Center, and an Dnergency Operations Facility. (III.A.l.2) 0FFSilORE' POWER SYSTEMS RESPONSE Emergency Respnse Facilities will be designed in accordance with guidance provided in NURID-0696.
The Onsite Technical Support Center (TSC) for the ENP consists of the supervisor's office, shift technical advisor area (previously the visitors area) and computer room, a3jacent to the Control Room (see Figure C-1, and additional space in the Dnergency Relocation Area separated from the O Operations Support Center) . %is center is provided with the same degree of V shieldiry, environmental control, missile protection and security as the Control Room. %is center uses the same ventilation system as the Control Rom ard also utilizes the Control Rocrn radiation monitorirg equipnent.
Necessary comunication between the TSC and to _ . the Control Room and Onsite Operational Support Center will be provided. Offsite comunications will be provided by the owner. As outlined below, plant status can be readily obtained in the TSC durirg normal as well as emergency operation.
Necessary "as-built" documentation will be filed in the TSC or elsewhere within the shielded control building.
The supervisors office and shift technical advisors area are directly adjacent to the Control Roczn and access is through a doorway directly into the Control Room. Additionally, a glass window in the common wall between these areas and the Control Room provides for easy observation of recovery activities. .Fbr these reasons, the instrumentation for the EC is enhanced and redundancy requirements minimized. Offshore Power Systeu will provide O
C-93
m CRT terminals to access data fra the control room. The -specific instru-
^
j mentation required in the N will be determined during final detailed
)
'd design of the FNP.
OPS believes that the FNP concept provides unique advantages regarding as-built documentation, includirg the following:
- a. greater level of detail on drawings (dimensioning, part numbers, etc.)
because of the manufacturirg concept.
- b. greater consistency and coordination among as-built documents, since
- c. FNP units and their documentation would be virtually identical, allowirg use of other units for full-scale studies regardirq recovery operations.
3 The Ehergency Relocation Area (at Elev. 100' and 109' in the control s
v/ building) beneath the Control Rom will be the Onsite operational Support Center. This area is designed to the same criteria for shielding, missile protection ard environmental controls as the Control Room. Emergency storage facilities and comunications equipment for onsite operational support are provided. The Ehergency Relocation Area is safely accessible from the Control Room via a stairway which is enclosed within the shielded control building.
The Near-Site Ehergency Operations Facility will be provided by the plant owner.
O C-94 m
REGULATION - 10CFR50. 34 (e) (2) (xxvi)
(
)_
Subject:
Leakage Reduction outside Containment To satisfy 'he following requirements, the application shall provide sufficient iaformation to demonstrate that the required actions will be satisfactorily completed by the operating license stage. %is information is of the type customarily required to satisfy 10CER50.7aa) (2) or to address mresolved generic safety issues.
Provide for leakage control and detection in the design of systems outside containment that contain (or might contain) TID 14844 source term radio-active materials following an accident. Applicants shall sutmit a leakage control program, including an initial test program, a schedule for re-testing these systems, and the actions to be taken for minimiziry leakage from such systems. We goal is to minimize potential exposures to workers aM public, and to provide reasonable assurance that excessive leakage will not prevent the use of systems needed in an emergency. (III.D.l.1)
OFFSHORE POWER SYSTEMS RESPONSE All reactor plant systems sich could contain TID 14844 source term radioactivs materials followirg an accident are isolated frm the Reactor Coolant System except for the following:
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- 1. Residual Heat Removal System (RHR)
- 2. Safety Injection System (SIS)
- 3. Containment Spray System (CSS)
A major design feature has been incorporated into the Floating Nuclear Plant which significantly reduces the potential for exposure to workers and to the p2blic following a design tesis accident. We feature involves incorporation of the above safety systems within separate safeguard compartments which are shielded, tich have 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire walls and 3 of which are watertight. The safeguards compartments are provided with a controlled atmosphere sich is connected to with the containment annulus durirg accident conditions to prevent the spread of radioactivity to other parts of the plant and reduce release to the environment. Wis feature is further discussed in Section 11.6 of the Plant Design Report.
O C-95
These three ~ safety systens will also incorporate various leak testing, leak
^ reduction and/or collection features, including:
()
- 1. Welded / seamless piping system;
- 2. Pumps with mechanical seals;
- 3. f.ow point drains from the piping system and drains from equipment such as peps, leakoff frm valves, etc. with double isolation valves, are routed to sumps;
- 4. High point vents with a single isolation valve and pipe caps or double isolation valves;
- 5. Pressure test connections for temporary (or local) instrumentation witir a single isolation valve and pipe caps or double isolation valves;
- 6. Leak offs from lantern rings piped to the sump for valves 4 inches ard larger.
- 7. Packless metal diaphragm valves utilized for 2" and snaller valves.
O To detect leakage between systems, the following provisions are incor-h porated in the design:
- 1. Radiation monitors with alarm annunciation in the control room are provided for the Essential Service Water systen which removes heat from the RHR, SIS and GS, thereby enabling detection of radio-active fluid on the non-radioactive fluid side due to heat exchanger tube leaks;
- 2. Flushing connections provided fo r post-accident RHR system maintenance have both an isolation valve ard a blind flange to prevent leakage. In addition, relief valves piped to the sump are provided for protection against overpressurization. (See response to 10CFRSO.34 (e) (2) (vii) for description of post-accident RlR maintenance.)
In addition to the features designed to prevent -the leakage of radio-activity, the potential for the release of radioactivity to the envirornnent is further redaced by the provision of secondary isolation and control j V
C-96
s
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e within sealed safeguards compartments which contain the RHR, SIS, and CSS systems. %e sealed _' compartments, depicted oa Figure . C-10, provide a (y
secondary barrier Alch, together with the automatic activation of the annulus air filtration system prevent the spread of airborne radioactivity within the plant. Filtered air ' discharged to the environment by the annulus air ' filtration system is limited to that necessary to maintain a negative pressure.
. During normal plant operation the safeguards compartments and containment annulus spices 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 peps start on smp level signal and discharge to the floor drain tank of the Liquid Waste Treatment (WrL)
System. %e Safejuards Area sump flow path will be modified as shown schematically on Figure C-10, to provide an alternate discharge line to the containment sump. We signal for acuating the alternate discharge path will n be selected during final design. This new configuration will prevent uncontrolled discharge from the Safeguards Area to the Auxiliary Building following an accident.
The coolant leakage control and detection program relative to the system's potentially containirn a TID 14844 source term following an accident are:
- 1. Design period. Incorporation of the design features discussed herein will minimize potential leakage from the safety systems.
- 2. Initial test period. Hydrostatic and operational testing of the safety syntems will be performed prior to the acetsnulation of any radioactive contamination. Safeguards area leak testing will be performed during the initial test period.
- 3. Operational period. Periodic safeguards area leak tcsting vill assure integrity of the three systems. We utility owner / operator will be responsible for developing a preventative maintenance V .
1 C-97
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4 i
- program lwill be utilized to reduce leakage from sources outside of a
- contairinent to as-low-as-practical. The preventative maintenance
- program will be developed.to determine leak rate at startup and at r - regular. intervals thereafter. ,
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FIGURE C-10 l
REGUIATION 10CFR50.34 (e) (2) (xxvii)
(.
Subject:
Radiation Monitoring (v)
To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage. %is infore.ation is of. the type customarily required to satisfy 10CFR50.35(a) (2) or to address unresolved generic safety issues.
Provide for monitoring of inplant radiation and airborne radioactivity as appropriate fo r a broad range of ' routine and accident conditions.
(III.D.3.3)
OFFSHORE POWR SYSTEMS RESPONSE Of fshore Powr Systems has designed a comprehensive Radiation Nnitoring System (RMS) for the Floating Nuclear Power Plant which provides adequate monitoring for a broad range of plant conditions. Included as part of the RMS are both area monitors and airborne monitors.
The area and airborne monitors assist in assuring that occupational
/7 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 FNP. Airborne monitors supplement area monitors in selected areas of the ENP by sampling the atmosphere to detect the concen-tration of significant radionuclides in particulate and/or gaseous form.
In addition to the above general functions, certain RMS channels provide indication and/or control for specific functional purposes. Wese special RMS functions include:
(1) Airborne monitoring of the containment atmosphere for in-containment reactor coolant leakage detection.
(2) High level area monitoring of the containment to follow the radi-ological course of a loss-of-coolant accident. (See also the response to 10CFR50.34 (e) (2) (xvii) .)
b U[
C-100
(
[ (3)' Monitoring of the control room and mergency relocation area to
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) provide automatic switchover of the ventilation - systems for these O areas to their emergency mode of operation if a high radiation level is detected. %is is an engineered safety feature for the ENP designed to insure habitability of these areas following postulated accidents sich could result in a significant release of radionuclides to the plant environs.
- (4) Monitoring of the normal and alternate outside air intake ducts of the
)
l main cuntrol room ventilation systs. These channels provide supple-mentary information to the primary wind direction instrumentation to allow the operator to verify that the least contaminated intake is utilized for ventilation following postulated accidents Wich could result in a significant release of radionuclides to the plant environs.
A total of twnty-eight area nonitors and six airborne n;nitors are included in the RMS. These monitors are described in detail in the Plant
(~~T Design Report, Sections 12.1.4 and 12.2.4.
V During the final design of the FNP, OPS will review all area and airborne monitors to ensure the adequacy of the design, location and rarges, including a determination of which monitors must meet the requirements of l Regulatory Guide 1.97, Revision 2.
With regard to improved in-plant iodine monitoring, Offshore Pow r Systems will provide space on the FNP for counting rooms and laboratories where analyses of radiciodine concentration can be perfo rmed . %e location of these spaces and support systems design are such as to permit personnel occupancy for times required to perform necessary analysis following accident conditions. Shieldiry will be provided to ensure a low background in the counting room. Ventilation with clean air at a pressure higher than surrounding spaces will be provided for the countiry room to minimize
- background airborne contamination in this region. Capability for purging of entrapped noble gases from charcoal samples usirs either clean air or A
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C-101
nitrogen will be provided in the laboratory area. Residual noble gases will
,m be routed to and vented from the plant stack.
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Sampling methods, counting equipment and other laboratory analytical equipment will be provided by the plant owner. We gama ray spectroueter is a comercially available method for discriminating between residual noble gases and radioiodine adsorbed on the charcoal filters in the atmospheric sampling devices. OPS will recomend to the plant owner that
- such equipment 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 radiciodine following accidents, i
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O C-102
RIIIUIATION 10CFR50.34 (e) (2) (xxviii)
(w ,-
)
Subject:
Control Room Habitability To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the required actions will be satisfactorily completed by the operating license stage.1his information is of the_ type customarily required to satisfy 10CFR50.35(a) (2) or to address triresolve.d generic safety issues.
Evaluate potential pathways for radioactivity and radiation that may lead to control room habitability problems under accident conditions resulting in a TID 14844-source term release, and make necessary design provisions to preclude such problems. (III.D.3.4)
OFFSHORE POWER SYSTEMS RESPONSE The Floating Nuclear Plant design accommodates required operating personnel in a safe state of occupancy for the duration of postulated accidents. This is accomplished by providing areas within the plant which are protected from external hazards including radiation and toxic gases. These areas, the Control Room and the Emergency N1ocation Area, are safeguarded by filtered ventilation systems and by biological shielding. The Dnergency Relocation
( I Area (ERA) provides the facilities necessary to support the operating crew, e.g., food supplies and food preparation equipnent, medical supplies, sleeping accommodations, and communications equipment.
Biological shielding for the Control Room has been designed to comply with General Design Criterion 19 ( i . e . ,. 5 rem gama whole body dose,' 30 rem thyroid dose, and 30 rem beta sk.in dese) for the duration of an accident.
Postulated accidents analyzed include the loss-of-coolant accident, fuel handling accident, main steam line rupture and gas decay tank rupture.
Detailed results of the habitability analyses, Wilch are in compliance with Standard Review Plan 6.4, are given in Section 6.5 of the Plant Design Report.
Control Room shielding, designed to attenuate the direct radiation fro n fission products within the contairrnent and those leaked from containment, consists' of a 2 foot thick concrete roof and concrete wall on the side m facing the containment building, and one foot thick walls on the sides not
[v)
C-103 L
facing the containment. %e back wall of the control room 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. Rese walls thus encompass the Control Room, the Process Rack Room, the Cable pull area, and the ERA. We four rooms are separated by a floor / ceiling Wich is a 3-hour fire barrier.
Source terms for the radiation analysis are based on applicable Regulatory Guide assumptions. For the loss-of-coolant accident, Regulatory Guide 1.4 assmptions were used.
Protection from airborne radioactivity is provided by the control building ventilation system operatirg in the post-accident mode. We post-accident mode is initiated automatically 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 Relo-cation Area. Basically the post-accident operational mode consists of closing all Control Room and ERA exhaust ports, providing controlled intake to maintain s 'ositive pressure, and recirculating the internal atmosphere.
s This positive pressure prevents inleakage of potentially contaminated air from surroundirg spaces. Independent ventilation systems serve the ERA and Control Room. Each system also continuously recirculates a considerable quantity of air through the filters to rer.iove any potential iodine activity within the ERA and Control Room. A o. ' ailed description of the ventilation system is given in Section 9.4.1 of the Plant Design Report.
The ventilation systems for the control room and ERA hwe dual intakes which are physically separated. Dual intakes allow outside air to be drawn from a region Were the concentration of radioactivity is relatively low following an accident. The preferred air intake is automatically selected in response to a wind direction controller to assure that the preferred intake is on-line continuously. Radiation detectors will be used in the ventilation air intakes as a precautionary measure to indicate any mea-surable levels of activity and confirm the correctness of the chosen air intake. We operator can override the automatic feature. Outside air is
, brought in through charcoal filters at a maximm rate of 100 CEN as n necessary to maintain a positive pressure of 0.25" water pressure.
C-104
7 g ne results of an extensive wind tunnel measurements program (1) employing V scale models of tw> Floating Nuclear Power Plants located within a scale model of a typical breakwater were used to determine locations for the alternate contro} room ventilation intakes and to determine the atmospheric dispersion factors at the intakes used in accident analysis.
De pathways for internal contamination at the Control Room at 'IMI-2 were:
(a) lack of adequate control room access control, (b) access by contami-nated personnel, (c) doors that were left open, and (d) the inability to accurately monitor the c sntrol room atmosphere in the recirculation mode.
The FNP control room will not have the difficulties listed in (a), (b) and (c), above, because as the plant will be provided with a dedicated Tech-nical Support Center (TSC) and an onsite Operational Support Center to be used as staging areas for mergency support personrel. 'No radiation area monitors are provided inside the control room to ir .cate possible control room airborne contamination at all times. Portable iodine monitors also r will be available to control room personnel.
(3.
)
The 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 FNP design meets the applicable requirements.
Section 9.3.7 of the Plant Design Report states that there are no toxic gases stored on the FNP. We only hazardous chenicals used and stored on board are sodium hydroxide and sulfuric acid. Oilorine, normally used for water treatment, is not a supply or storage item; soditan hypochlorite, used as a blocide for the circulating water system, is generated on board as described in Section 9.3.7 of the PDR.
(1) " Wind Engineering Study of Atmospheric Disierison of Airborne Materials
.g Released from a Floating Nuclear Power Plar.t", R. N. Meroney, et. al.,
I i Colorado State University, August,1974.
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C-105 ;
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4 i
Sodium hydroxide and sulfuric acid are stored as liquid solutions in tanks i
[mV) with separated retainirg walls to contain the solutions in the unlikely event of tank leakage or rupture. %e tanks are located on the 100 ft.
level between bulkhead A & B and 5 & 6. This area is on the opposite side
. of the plant and 54 feet below the control room level. We vapor pressures of the solutions are such that exposure of the liquid in the tanks to the t atmosphere does not present an airborne hazard Iaakage from either a tank '
or filling line will neither interfere with normal operation of the plant j nor affect any safety related equipment.
4 i
Effects of offsite storage of potentially toxic chemicals are site depen- j dent. Generic data have been calculated for use by the plant owner and are i presented in Section 6.5 of the PDR. i L
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C-106
REGUIATION 10CFR50.34(e) (3) (i)
(,_)
Subject:
Experience Feedback w
To satisfy the following requirements, the apolication shall provide sufficient information to demonstrate that the requirement has been met.
21s information is of the type custorr.arily required to satisfy 10 CFR 50.34 (a) (1) or to address the applicant's technical qualifications and management structure and competence.
Provide administrative procedures for evaluating operating, design and construction experience and for ensuring that applicable important industry experiences will be provided in a timely tranner to those designing and constructing the plant. (I.C.5)
OFFSHORE POWER S'! STEMS RESPONSE Offshore Power Systems policies and procedures establish a formal system through which industry experience is continuously collected, screened and distributed to those responsible for the various plant design and tranu-facturirg functions. Significant experiences are evaluated to determine if changas are warranted to (1) the Floating Nuclear Plant design, (2) ENP Manufac ;uring techniques and/or (3) Offshore Power Systems administrative fr y procedures.
The Nuclear Engineering Division has basic responsibility for administra-tion of the experience feedback systs. Specific tasks assigned to Nuclear Engineering include collection of experience data, initial screening, distribution and record maintenance.
Offshore Powr Systems has been collecting experience data in the form of I&E Bulletins, Notices and Circulars since being placed on the Region II distribution list (at our request) in 1978. In addition, operating e:tperi-ence data have been available from Westinghouse and in the form of Licensee Event Report Summaries. %ese sources will be expanded to include reports issum by EPRI and INPO. Also available to Offshore Power Systems are the information resources of the Westinghouse Owners' Group. Information will also be available in the longer term from in-house preoperational testing experience and from customers' in-service experience.
C-107
Screening of experience data are performed to reduce the voltsne of material
[; '
for which more detailed evaluation and disposition will be required. We screening process will intercept ex traneous, insignificant or duplicate experience reprts, leaving only that information which has the potential to cause design or procedural changes to be distributed for action. In this way the experience feedback system is expected to have minimtsn adverse impact on the talance of dealgn and manufacturing functions. We screening function will be performed by persons who are qualified by experience and training to judge the potential significance to plant design and construc-tion activities.
Experience data Wich are judged to be potentially significant will be distributed to the appropriate organization for detailed evaluation and disposition. Since Of fshore Power Systems engineering is organized along functional lines (electrical, mechanical, structural, etc.) it will be a relatively simple task to establish the appropriate distribution for design-related items. Experience reports dealirg primarily with plant construct. ion will be distributed to a single designated point within the f) Operations Department for further distribution as appropriate. Quality
\#
Assurance will be included on the distribution for all ptentially sig-nificant experience reports. A lead manager (group) will be assigned to each such experience report. tis individual will be responsible for experience report evaluation. We lead manager is also required to initiate any changes Wich the experience evaluation shows to be warranted.
Changes resulting from the experience feedback systen may affect documents which come (sider the scope of the configuration control system. Configura-tion control provides a formal device to manage changes to doctanents which provide input to a more detailed level of design or which are released for procurement or manufacture.
The experience feedback system is structured in such a way as to be readily auditable.
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v C-108
REGUIATION 10CFR50.34 (e) (3) (11) i,
)
Subject:
' Quality Assurance List To satisfy the following requirements, the application shall provide sufficient information' to demonstrate that the requirement has been met.
This information is of the type customarily required to satisfy 10 CFR 50.34 (a) (1) or to address the applicant's technical qualifications and management structure and c;opetence.
Ensure that the quality assurance (QA) list required by Criterion II, App.
B., 10 CFR Part 50 includes all structures, systems, and components important to safety. (I.F.1) -
OFFSHORE POWER SYSTEMS RESPONSE Implementation of NRC requirements assure that QA measures are applied to a comprehensive set of structures systems and components important to safety.
In ado ~ won to NRC requiements, Offshore Power Systems utilizes an internal ;
classification system which adds many structures, systems and components to the Quality Assurance list @ich might not otherwise receive more than standard comercial Quality Assurance measures. We application of both NRC
[]
L.J and OPS Quality Assurance requirements in the ENP are outlined below.
The general design criteria for nuclear power plants are contained in 10 CFR 50, Appendix A. Wese criteria provide a broad definition of plant structures, Systems and components important to safety. NRC Quality Assurance regulatioris are contained in 10 CFR 50, Appendix B. %rous ' the l 1
t mechanisn of regulatory guides the NRC has imposed Appendix B Quality l
'ssurance requirements to various structures, systems and components, based on the characteristics of the particular structure, system or component concerned. Wese Regulatory Guides (which are sumarized below) deta f 1 the degree of Quality Assurance required for virtually all of the structures, systems and components which are the subject of General Design Criteria.
o Reg ilatory Guide 1.26 identifies the nuclear plant fluid systems which fall into quality classifications A, B, C and D. Offshore Power Systems complies with this Regulatory Guide; however, industry safety classi-fications (1, 2, 3 and Non-Nuclear Safety or NNS) are used in place of ,
[N quality groups A, B, C and D. Offshore Power Systems procedures require (v)
C-109
l
)
.- l appropriata 8.pperdix B Qality Assurance measures for all systems and
(,m) . components classified as Safety Class 1, Safety Class 2, safety Class 3 or NNS.
~
o Regulatory Guide 1.29 requires that *he Quality Assurance Program of 10 CFR 50, Appendi> B be applied to each of the structures, systems and components listed in Regulatory Positions 1, 2 and 3. ta Quality Assurance measures invoked' by Offshore Power Systems for Floating Nuclear Plant structures, systems and components complies with Regula-tory Guide 1.29.
t Regulatory Guide 1.120 establishes the QA requirements for the Fire Protection Sy stem. These requirements are unchanged fra those of Branch Technical Position APCSB 9.5-1 (Appendix A) which were committed to in Offshore Power Systes Report RP06A30, " Floating Nuclear Plant Fire Protection Evaluation", September, 1977.
o Regulatory Guide 1.143 supplements Regulation Guides 1.26 and 1.29 for O Radwaste Syste.ns. Regulatory Position 6 of this guide details an V
acceptable Quality Assurance Program for Radwaste Systems. Offshore Power Systems will meet or exceed the requirements of Position 6 in future design and manufacturing activities.
Offshore Power Systems engineering procedures require the responsible engineers to classify each Floating Nuclear Plant Structure ar.d System using a pre-defined set of classifications. We set of Offshore Power Systems classifications includes several classifications in addition to those defined in Regulatory Guides 1. 26 and 1.29. Offshore Power Systems Quality and Reliability procedures establish three quality levels and provide the correlation between the various engineering classifications and the three quality levels. Quality Level 1 invokes appropriate portions of the full Quality Assurance Program of 10 CFR 50, Appendix B. Quality Invel 2 invokes (as a minimum) requirements for procurement document control, control of non-conforming items and Quality Assurance records. Quality I Level 3 requires no Quality Assurance measures beyond standard comercial D)
\
v I
C-110
practice. Both engineering procedures ard quality ard reliability proce-
']
dures teceive extensive management review during preparation and are approved for use at a senior management level.
The following surrary indicates the existing Quality Assurance Level of those Floating Nuclear Plant systems which are of particular interest in light of the accident at 'Ihree Mile Island. As this sumary indicates, the application of quality assurance measures is very extensive in the Floating Nuclear Plant.
SYSTEM QUALITY LEVEL .
All structures, systems and 1 '
components listed in Regulatory Guld
- 1.26 and 1.29 Main Steam, including steam dtznp 1 (Note 1)
Main Condensers 2 - tubes t 1 - shell Circulating Water 2 Cordensate Polishing 1 (Note 1)
Condensate - Feedwater 1 (Note 1)
Instrument Air 2 ,
Dnergency Instrtrnent Air 1 '
Containment Post-Accident Sampling 1 Nuclear Plant Sampling System 1 Steam Generator Blowdown 1 (Note 1) I (1) Quality level 1 applies to main components and flowpaths. Lesser levels may be applied elsewhere in the system.
1 offshore Power Systems engineerirg procedures provide for the identifica-tion of structures, systems and components (including related consumables)
I
%J C-lli
to Wich are applied each of the three OPS quality levels. %ese struc-
[) tures, systems and components (along with the assigned quality level) appear in the Material Order List (MOL) which, *en cx>mplete, will catalog the totality of materials required to fabricate a Floating Nuclear Plant.
The Offshore Power Systems procedure Wich establishes the MOL identifies the persons responsible for its preparation and distribution. Since the MOL is e m puterized, it - can be readily sorted by quality level (or other characteristic) to provide a stenary of materials ' receiving each level of quality assurance. In the system of design, procurement and manaufacture employed at Offshore Power Systems, the MOL is not the document which controls the level of quality assurance applied to materials. Rather, the MOL functions as a master data sununary.
The gelity level required for various methods is defined initially in engineering specifications and subsequently in purchase specifications.
Theco and related doctznents (such as flow diagrams, layout drawings, etc.)
come under the scope of the formal configuration control system. Briefly, the configuration control system requires a written request for a design
[] change including identification of all affected doctrnents followed by formal review, including Quality Assurance and Management.
When a document which establishes design information (for example, safety class or quality level) is changed, appropriate revisions are made to the MOL as a part of the change process. Thus, changes to the MOL are in-directly controlled through the configuration control system. More importantly, those documents which directly affect quality are directly controlled.
Additional information is contained in the response to ,
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v C-112
REGUIATION 10CFR50.34 (e) (3) {iii) 7
Subject:
Quality Assurance Program To satisfy the following reqairements, the application shall provide sufficient information to demonstrate that the requirenent has been met.
This information is of the type customarily required to satisfy 10 CFR 50.34 (a) (1) or to address the applicant's technical qualifications and management structure and competence.
Establish a quality assurance (QA) program based on consideration of: (A) ensurirg independence of the organization performing checking functions from the organization responsible for performing the functions; (B) performing quality assurance / quality control functions at construction sites to the maximtsn feasible extent; (C) including QA personnel in the documented review of and concurrence in quality related procedures asso-ciated with design, construction and installation; (D) establishing criteria for determining QA programmatic requirements; (E) establishing qualification requirements for QA and QC personnel; (F) sizing the QA staff commensurste with its duties and responsibilities; (G) establishing procedures for maintenance of "as-built" documentation; and (H) providing a QA role in design and analysis activities. (I.F.2)
OFFSHORE POWER SYSTEMS RESPONSE (N The eight criteria contained in the rule have been developed in consider-b) able detail in a staff position paper entitled, " Proposed Quality Assurance Guidance to Satisfy NUREG-0718 and Proposed Rule." Althotsh this position paper is addressed in its entirety in this response, some of the require-ments apply to 10CFR50.34 (e) (3) (ii) and (vii). Ebr the most part the position paper consists of 'IMI-related changes to Revision 1 of the Standard Review Plan (SRP), Section 17.1.
In November 1980 Offshore Power Systems performed an internal review of the overall FNP quality assurance program to assess the degree of compliance with the acceptance criteria set forth in Section 17.1 of the SRP (Revision 1). 'Ihis study forms a convenient basis for addressing the latest staff requirements, because these requirements result, for the most part, from minor changes to SRP 17.1, Revision 1.
i V
C-113
Table C-5 provides the following information for each of the present staff positions:
v o We corresponding section in draf t Revision 2 of SRP 17.1 (in only a few cases no corresponding section is identified) o %e section in Revision 1 of SRP 17.1 which contains the same basic requirement as the present staff position o A brief description of the difference between the present staff position and SRP 17.1, Revision 1. Similar descriptions are also !
provided in those few cases tere the present staff position is not based on Revision 1, i.e., where the basic requirement has been added ir. Revision 2 or where the basic requirement is not contained in either -
Revision 1 or Revision 2.
'o %e OPS doctznent dich establish policies and procedures related to the present staff requirement o A statement as to whether the existing OPS policies / procedures are in -
compliance with SRP 17.1, Revision 1 (based on the November 1980 internal audit) .
o A statement as to whether the existing OPS policies / procedures are in t c oliance with the present staff requirements (based on the differ-ences beteen the present staff position paper and Revision 1 to SRP 17.1).
i Notes and remarks at the end of Table C-5 provide clarifications and actions which will be taken by Offshore Power Systems where necessary to establish full compliance with present staff requirements.
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TABLE C-5 (coheru/ ued) 's/
CORRESPONDl % SECTION DOCUMENT OPS POLICY / PROCEDURE STAFF 0F SRP 17.1 CONTAINI% IN COMPLIANCE 7 POSITION DIFFERENCE BET'aLEN POSITION OP5 PCLICY/ hOTES NR. IN SRP 17.1 REVISION 2 AND REVISION 1 PROCEDl RE STAFF REV. 2 REV. 1 SkP REV. 1 P051T10N 2A7 10C3 10C3 QA organization to evaluate, verify and document QRP 5.1 Yes Yes 11C1 11Cl completeness QRP 17.0 QRS 100 QRP 11.0 2A8 16.3 16.3 Follow- sp action to be taken by QA organization QRP 16.0 Yes Yes QRP 16.1 251 J'3 1C3 Person in charge of site QA Program to be iden- QRP l-1 Yes Yes 2 tified by position within the QA organization 2B2 IB6 None New in Rev. 2: Requires participation by QA -
Yes 3
organization in day-to-day planning of construc-tion activities o
- ) 2C1 2Bla 2B1 Restricts applicability to quality-affecting QA Manual Yes Yes l$ procedures EPPs Purchasing Manual Operations Manual 2C2 2 Bib None New in Rev. .: QA organization required to re- OPP 1.2 Yes Yes view and document approval of procedures af- EPP 203 fecting quality Purchasing Manual -
2C3 4A1 4Al QA personnel to review and document concurrence QRP 4.0 Yes Yes with procurement documents QRP 4.1 QRS 100 2C4 6A2 6A2 QA organization to review and decument approval QRP 6.0 Yes Yes of documents and changes EPP 206 2C5 10C1 10Cl Controlling documents to be reviewed by QA or- QRP 5.1 Yes Yes ization and technical organizations, as appro- QRP 10.0 priate QA Manual
- 0PP = Operations Policy and Procedures
s k TABLE C-5 k :.inued) U CORRESPONDING SECTION DOCUMLNT~ OPh POLICY /PROCELXJRE STAFF 0F SRP 17.1 CONTAINING IN COMPLIANCET -
POSITION DIFFERENCE BETWEEN PuSITION OPS POLICY / NOTES NR. IN SRP 17.1 REVISION 2 AND REV!510re 1 PROCEDURE
- STAFF REV. 2 REV. I SRP PEv.1 POSITION 2C6 llB1 llB1 Controlling documents to be reviewed by QA or- QRP 11.0 Yes Yes
, ganization and technical organizations, as appropriate 2C7 12.3 12.3 Review and concurrence required. Responsible QRP 12.0 Yes Yes
, organization to be identified 2c8 13.2 13.2 QA organization to review procedures and docu- QRP 13.1 Yes Yes ment approval QRP 13.2 2C9 14.1 14.1 QA organization to review procedures and docu- QRP 14.0 Yes Yes 14.4 ment approval QRP 15.0 2C10 14.2 14.2 QA organization to review procedures and docu- QRP 14.0 Yes Yes n 14.4 ment approval QC Manual b
C 2C11 14.3 14.3 QA organization to review procedures and docu- QRP 14.0 Yes Yes 14.4 ment approval 2Cl2 15.1 15.1 Requitc identification of individuals author- QRP 15.0 Yes Yes ized for independent review of non-conformances QRP 14.0 2C13 15.2 15.2 Requires documented concurrence by QA organi- QRP 15.0 Yes Yes zation QRP 15.1 2C14 16.1 16.1 QA organization to review procedures and docu- QRP 16.0
- Yes Yes ment approval 2D1 2B3 2B3 Rewritten in Rev. 2: QA participation in the QRP 2.2 Yes Yes plant definition pahse EPP 211 2D2 7B4 -
Rewritten in Rev. 2: Quality verification for -
No 4 off-the-shelf items i
~
(# ' .0 ( \
k TABLE C-5 ( b inued) O CORRESPONDl % SECTION DOCUMLh1 UP5 Put!CY/PkuCEuuRE STAFF OF SRP 17.1 CUNTAIN!?6 IN COMPLI ANCE7 POSITION DIFFERENCE BETWEEN P051TIOf4 OPS POLICY / NOTES NR. IN SRP 17.1 REVISION 2 AND REV1510N 1 PROCEDURE
- STAFF REV. 2 REV. 1 ERP REV. 1 POSITION 2D3 10A 10A Participation of QA organization in equipment QRP 10.0 Yes Yes inspection required QRP 4.0 QRS 100 QA Manual Metrology Manual 2D4 10C2 10C2 QA organization to participate in procedure es- QRP 4.1 Yes Yes tablishment QRP 5.1 2D5 11A1 llAl None QRP 11.0 Yes Yes 2D6 18B1 18B1 QA organization to analyze audit data QRP 18.1 Yes Yes 2El 2D(a)- 2D(a)- Qualification required for QA/QC personnel QRP 2-3 Yes Yes
- g 2D(c) 2D(c) QRP 2-4
- g 2D(d) 2D(d) Proficiency tests and acceptance criteria for QRP 18.0 to qualification required NDE Manual 2D(e) 2D(e) Rewritten: Qualification certificate to state specific functions and qualification criteria 2D(f) -
Retraining, etc. is required 2D(g) 2D(e) Rev. I to Reg. Guide 1,58 2E2 10B2 10B2 QA organization to direct qualification program Yes Yes QRP 2.3 NDE Manual 2F1 1A5 1A5 Basis for sizing QA organization - -
Yes 2, 5 2F2 - -
Not in SRP: QA to be involvad in long-range Yes 5
proj ec t scheduling and QA/QC staffing. Staffing periodically reviewed.
2G1 6Al 6Al(d) As-built documents versus drawings. - -
Partial 6 2G2 6C1 -
Procedures to require timely as-built documenta- - -
Partial 6 tion
s
]
N TABLE C-5 /
.tinued) v CORRESPONDING SECTION DOCUMENT UPS POLICY /PR0t.LUURE i
STAFF OF SRP 17.1 CONTAINING IN COMPtIANCE7 POSITION DIFFERENCE BETWEEN PUSITION OPS POLICY / NOTES i NR. IN SRP 17.1 REv!SION 2 AND REVISION 1 PROCEDURE STAF F l
REV. 2 REV. I SRP REV. 1 POSITION 2H1 3El 3El Rewritten in Rev. 2: Procedures to require docu- EPP 252 Yes Yes ment check for dimensional accuracy and com- QRP 3-1 pleteness. EPP 232 y 2H2 3E2 3E2 Rewritten in Rev. 2: Drawings & specifications EPP 232 Yes Yes to be reviewed by QA (or other qualified per- EPP 252 sons) to assure conformance to policies QRP 3-1 QRP 10-0 3-la - -
Role and attitude of top management towards QA PPM 6-101 Yes Yes to be described 3-lb - -
QA management and responsibilities QRP 1-0 Yes Yes 9 3-2 1A1 1A1 None QRP l-0 Yes Yes C
- 3-3 1A2 1A2 None QRP l-1 Yes Yes 7 3-4 1A3 1A3 Extent of management oversight to be described - - -
7 for delegated QA functions 3-5 1A4 1A4 None QRP 1-2 Yes Yes 3-6 1A6 1A6 None PDR Section Yes Yec 17.1.2 3-7 1B1 IBl None PDR Section Yes Yes 17.1.2 3-8 1B3 IB3 None QRP l-1 Yes Yes 3-9 1B4(a) IB4(a) None QRP 1-0 Yes Yes QRP 15-0 IB4(b) -
New in Rev. 2: Positions with stop-work auth- - -
Partial 8 ority to be identified
O TABLE C-5 tinued)
I CORRESPONDING SECTION DOCUMENT UPS POLICY / PROCEDURE STAFF OF SRP 17.1 CONTAINING IN COMPLIAXE7 POSITION DIFFERENCE BETWEEN M)SITiuN OP5 POLICY / NOTES NR. IN SRP 17.1 REVISION 2 AND REVISION 1 PROCEDURE STAFF REV. 2 REV. 1 SRP REV. 1 POSITION 3-10 1B5 -
Neu in Rev. 2: Provisions required for resolution QRP 2-0 -
Yes of QA disputes 9
3-11 1C1 lCl None QA Manual Yes Yes Policy Stmt.
3-12 1C2 1C2 None QRP 1-0 Yes Yes 3-13 2A2 2A2 None QRP 1-0 Yes Yes
!?
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TABLE C-5(continued)
NOTES -
- 1. As stated in the response to 10CFR50.34(e)(3)(ii), the Offshore Power Systems quality assurance program is applied to structures, systems and components covered by 10CFR50, Appendix A and Regulatory Guide 1.29.
Specific reference to 10CFR50, Appendix A will be added to the appro-priate policy / procedure document (s). (Reference is presently made to Regulatory Guide 1.29).
- 2. The quality assurance organization at the manufacturing facility is not a separate organization.
- 3. QA participation in the day-to-6ay planning of activities during plant manufacture will be recognized in the appropriate policy / procedure document (s) well in advance of the commencement of manufacturing ac-tivities. QA participation has always been intended, and this action is required only to provide adequate procedural documentation.
- 4. Offshore Power Systems policies / procedures will be expanded to require special verification requirements for off-the-shelf items where speci-fic QA controls appropriate for nuclear applicatior s cannot be imposed in .: practical manner. The necessary procedural revisions will be made fs in advance of affected purchasing activities.
/ T f
'/ 5. As a member of the President's Staff, the Director of Product Assurance participates in long-range planning of Floating Nuclear Plant engineering, purchasing and manufacturing activities. The QA and QC functions are and will be staffed at a level commensurate with the present and planned activities. Since Offshore Power Systems carries out all product as-surance activitics, the level of QA/QC staffing is entirely under Offshore Power Systems' control.
- 6. Since embarking on the FNP project, Offshore Power Systems has recog-nized the need to provide as-built documentation as part of the customer data package. Written procedures for production of as-built documenta-tion were (and remain) planned for future development. These procedures will be in place well in advance of their need.
- 7. Offshore Power Systems does not delegate any major portion of its QA program to other organizations.
- 8. The Director of Product Assurance is authorized in writing to stop un-satisfactory work. At an appropriate future date (in advance of com-mencing manufacturing activities) the director will identify in writing those individuals (by name or position) to whom stop-work authority is delegated.
4
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i C-121
REGUIATION 10CFR50.34 (e) (3) (iv)
,4
Subject:
Dedicated Contairunent Penetrations
.c) g To satisfy the following requirements, the application shall provide sufficient information to demonstrate that the requirement has been met.
' Itis information is of the type customarily required to satisfy 10 CFR 50.34 (a) (1) or to address the applicant's technical qualifications and management structure and competence.
Provide one or more dedicated containment penetrations, equivalent in size to a single 3-foot diameter openirg, in order not to preclude future instalation of systems to prevent containment failure, such as a filtered vented containment system. (II.B.8)
OFFSHORE POWER SYSTEMS RESPONSE The ENP containment design will reserve space for four 18 inch diameter penetrations in order not to preclude the installation of systems to prevent containment failure. These penetrations will be located at approximately 230 foot elevation on the contairunent 180 azimuth and will be capped and seal welded. 'Ihese penetrations will meet all requirements for spare penetrations.
O v
, V)
C-122
1
- REGULATION 10CFR50.34(e) (3) (v)
. y .
V
Subject:
Degraded Core-Matters
~
Ti satisfy the following requirements, the application shall provide
.sufticient information to demonstrate - that the requirement has been met.
-This information is of the type customarily required to satisfy 10 CFR ,
50.34 (a) (1). or to address the applicant's technical qualifications and management structure and competence.
Provide preliminary design information at a level of detail consistent with that normally required at the construction permit stage of review suffi-cient'to demonstrate that: (II.B.8) '
(A) Containment intep Ity will be maintained (i.e., for steel containments
~
by meeting the ie ' .rements of the ASME Boiler and Pressure Vessel Code,Section III, Jivision 1, Sub-subarticle NE-3220, Service Iavel C Limits, except that evaluation of instability is not required, i considering pressure and dead load alone. For concrete containments 4 by meeting the requirements of the ASME Boiler Pressure Vessel Code,Section III, Division 2 Subsubarticle CC-3720, Factored Ioad Category, considering pressure and dead load alone) during an accident that releases hydrogen generated from 100% fuel clad metal-water reaction accompanied by either hydrogen burning or the added pressure from post-accident inerting assuming carbon dioxide is the inerting agent, depending upon which option is chosen for control of hydrogen. As a minim m , the specific code requirements set forth above appropriate O
D for each type of containment will be met for a combination of dead load and an internal pressure of 45 psig. Modest deviations from
- these criteria will be considered by the staff, if good cause is shown by an applicant. Systems necessary to ensure containment utegrity t shall also be demonstrated to perform their function under these ,'
conditions.
(B) The containment and associated systems will provide reasonable assurance that uniformly - distributed hydrogen concentrations do not exceed 10% during and following an accident that releases an equiva-lent amount of hydrogen as would be generated from a 100% fuel clad metal-water reaction, or that the post-accident atmosphere will not support hydrogen combustion.
(C) The facility design will provide reasonable assurance that, based on a '
100% fuel clad metal-water reactirn, combustible concentrations of hydrogen will not collect in areas Wiere mintended combustion or
- detonation could cause loss of containment integrity or loss of '
appropriate mitigating features.
(D) If the option chosen for hydrogen control is post-accident inerting:
(l_) Containment structure loadings produced by an inadvertent full inerting (asstning carbon dioxide) , but not includirg seismic or design basis accident loadings will not produce stresses in steel '
' containments in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code,Section III, Division 1, Sub-subarticle NE-3220, (O) v Service Level A Limits, except that evaluation of instability is not C-123
1 required (for concrete containments the loadings specified above will n
i !
not produce strains in the containment liner in excess of the limits set forth in the ASME Boiler and Pressure Vessel Code,Section III, Division 2, Subsubarticle CC-3720, Service Ioad category) , (2) A pressure test, which is required, of the containments, at 1.10 and~
1.15 times (for steel and concrete containments, respectively) the pressure calculated to result from carbon dioxide inerting can be safely conducted, (3) Inadvertent full inerting of the containment can be safely acconinodated during plant operation.
(E) If the option chosen for hydrogen control is a distributed ignition system, equipment necessary for achieving and maintaining safe shutdown of the plant and maintaining containment integrity shall be designed to perform its function during and af ter being exposed to the environmental conditions created by activation of the distributed ignition system.
CFFSHORE POWER SYSTEMS RESPONSE (A) The following subsections discuss the pressure response of containment to an accident in which a large amount of hydrogen is produced and the capability of containment to withstand internal pressure. As will be seen, the calculated peak pressure is well within the minimtra contain-
) ment pressure capability required by the Commission. Equipment survivability in the post-accident containment is also addressed.
Containment Response To control hydrogen that could be released during a postulated degraded core accident, Offshore Power Systems will incorporate a distributed ignition system into the ENP containment design. 'Ihe Hydrogen Ignition System is described in the response to 10CFR50.34 (e)(2)(ix) and is assumed to function in the analyses described below.
Analyses were performed for the ENP using a preliminary version of the CIASIX computer program, currently mder developnent at Offshore M.er Systems. The analyses asstaned 1) hydrogen is released to the con-tainment at a constant rate between 0.5 and 5.0 pounds per second, 2) the total amount of hydrogen released to the containment is equivalent to the- amount of hydrogen generated by a 100 percent fuel clad metal-water reaction, 3) contairrnent safeguards are fully operational, b
U C-124
B
.and 4) complete combustion occurs with ignition at a hydrogen concen-tration of 10 percent by voltsne. We version of CIASIX used in these >
analyses has a conservative single node model of the ice condenser and includes convective but not radiant heat transfer to passive heat sinks.
For the cases considered, the peak calculated containment pressure during a hydrogen burn was 34 psig which is well below the minimum required containment " Service tevel C" pressure capability of 45 psig.
Conservatisms in the analyses include 1) high hydrogen production rates, 2) a high hydrogen ignition setpoint, 3) a single node repre- !
sentation of the ice condenser, and 4) no radiant heat transfer.
Recent analyses performed for the Sequoyah ice condenser containment l D
with a more detailed version of CIASIX are more realistic than the ,
analyses described above. We Sequoyah analyses use hyurogen produc-tion rates representative of a small break accident scenario with a i maximten rate of approximately one pound per second and an average rate O of approximately one-half pound per second. These analyses also consider ignition between 6 and 10 percent hydrogen by voltane as ,
supported by recent Fenwal test data (1}( } on glow pltg ignitor performance, have a two node representation of the ice condenser, and include both convective and radiant heat transfer to passive heat sinks. Peak calculated pressures in these analyses are 12 to 13 psig. -
Because of the large similarities between ice condenser containment designs, these results are representative of peak pressures expected !
to be calculated for the FNP contairrnent.
i (1) Tennessee Valley Authority, Sequoyah Nuclear Plant, "Research Program on Hydrogen Combustion and Control - Quarterly Progress Report #2",
March 16, 1981.
(2) Tennessee Valley Authority, Sequoyah Nuclear Plant, " Report on the ;
Safety Evaluation of the Interim Distributed Ignition System", '
December 15, 1980.
O v -
C-125 L
)
l Containment Functional Capability
[
\
%e Floating Nuclear Plant steel containment vessel consists of the containment shell and the contalment base plate as described in the Plant Design Report (PDR) Sections 3.8.2.6 & 3.8.2.8, respectively.
The current design of the contairunent vessel is based on the uniform internal de::iign pressure of 15 psig given in PDR Section 6.2.1.2 and the non-uniform transient pressures given in PDR Chapter 15. Analyses of the containment functional capability have been performed for the current containment design. %e results are stenarized in Figure C-11.
In the analyses, the following calculation methods and design param-eters were considered:
- 1. Shell capability ws determined as the pressure producing gross yield behavior. Yield was based on Von Mises criterion.
- 2. Actual yield stress used in the calculation was assumed to be equal to 120% of the specified minimtzn yield stress.
- 3. A hand calculation ws performed on the shell with sneared out hoop stiffeners. This approach was verified by finite element elasto-plastic analyses of panels with discrete longitudinal and hoop stiffeners. The latter analyses were derived from work done by OPS for the Sequoyah and McGuireY ice condenser containment.
- 4. Platform capability was calculated using plastic analysis methods.
- 5. Eval.uation of the shell/ platform interface was based on the area of the platform structure backing up the containment shell as shown in Figures C-12 and C-13, and Table C-6.
- 6. Buckling analyses of the torispherical dome of the containment shell and the spherical cap of the equipnent access hatch were based on realistic buckling criteria.
1/ "An Analysis of Hydrogen Control Measures at McGuire Nuclear Station,"
v Voltane 2, Section 4.2.5, Duke Power Company, November 17, 1980.
C-126
,m Modifications to the contairunent were investigated and the results l ') ~ indicated that the containment functional capability can be increased from 55 psig to a pressure of 80 psig within the existing design concept and without excessive impact on the plant design (see Table C-7.
The containment vessel will be upgraded to meet the requirements of the ASME Code Service Level C Limits, excluding evaluation of in-stability, considering pressure and dead load alone, during an accident that releases hydrogen generated from 100% fuel clad metal-water reaction accompanied by hydrogen burning. Results to date indicated that the hydrogen burning pressure load is considerably less than 45 psig. %erefore, a minimun internal pressure of 45 psig is specified for the above design consideration.
Equipment Survivability he systems necessary to maintain containment integrity will be des:gned to perform their function under the conditions calculated to OG occur during the operation of the distributed ignition system. We identification, location, evaluation, and protection (if necessary) of equipment associated with such systems will be establisbed during the FNP final design.
(B) Based on the analyses discussed in the response to (A) and the results of tests on glow plug ignitor performance, it is concluded that with the use of a distributed hydrogen ignition system there is reasonable assurance that uniformly distributed hydrogen concentrations can be controlled to 10 percent or less following an accident that releases hydrogen generated from 100 percent fuel clad metal-water reaction.
1 (C) The ice condenser containment incorporates many features and processes that enhance nearly complete mixing of the containment atmosphere and prevent hydrogen pocketing. Wese include the air return fans, the hydrogen skimer system, the upper compartment spray system, natural circulation, and diffusion. W e air return fans circulate air from the I
(] upper compartment, through the fan /accunulator rooms, and into the v
C-127
main area of the lower compartment where convection currents mix it
'^'s with the lower compartment atmosphere. We flow then proceeds through the ice condenser back into the upper compartment. Mixing is promoted both by induced turbulence within each compartment and by flow between compartments. With a recirculation flow of 80,000 cubic feet per min-ute, the equivalent of the entire containment atmosphere is circulated once every 15 minutes.
We hydrogen skimer system takes suction from the top of the upper compartment (at the containment dome) and from dead-ended regions in the lower compartment as described in Section 6.4.1 of the Plant Design Report. This flow is added to the main recirculation flow and discharged into the fan /acetsnulator rooms in the lower compartment.
We low design flow rate of the skimmer system acts to isolate the dead-ended regions from the bulk of the lower compartment (where hydrogen would be released), thereby preventing significant inflow and subsequent buildup of hydrogen in these regions.
/]
The spray system causes strong turbulence and mixing in the upper compartment due largely to momentum tr ansf e r between the spray droplets and the air. Shear forces between the sprayed region and the relatively small unsprayed region promote mixing in the latte; areas.
Existirg features and processes of the ice condenser containment promote mixing and prevent pocketing. With these features and judicious location of ignitors, there is rer,sonable assurance that combustible concentrations of hydrogen will not collect in areas where unintended combustion or detonation could cause loss of containment integrity or loss of appropriate mitigating features.
(D) Post accident inerting is not proposed for the FNP; therefore this item is not applicabic.
(E) The equipment necessary to maintain containment integrity and equip-ment considered essential to remove the heat generated by a degraded
('sj . core will tr designed to perform its function tnder the conditions C-129
. . . ..~. -. - . - - , -.- . . .. - . . - -. . . . .. -. -. . .. .._ _. . ---__- .
calculated to occur - during the operation of the distributed ignition g Lsystem. The identification, location, evaluation, ard protection (if i necessary) of - such equipnent wi]1 be established during FNP final i design.
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TABLE C-6 4 PRESSURE CA'/ ABILITY OF CONTAINMENT SHELL-PLATFORM JUNCTION O
ls ,) SUPPORT SUPPORT AREA BETWEEN SU' PORT LOCATIONS LOCATION FROM DWCS. EQUIV. SHELL EQUIV. PRESSURE (IN2 ) THICKNESS (IN) TO PRODUCE YIELD
- IN THE SHETf. (PSI)
A 256
.79 99.44 B , 126
.49 61.86 C 215 Cl
- 207 1.09 137.09 C2
- 207 1.09 137.09 D 215
.49 61.86 E 126
.79 99.44 F 256
.58 72.85 G 276
.69 87.39 H 126
.50 63.24 I 222 O'- .76 95.90 J 218
.76 95.90 K 222
.50 63.24 L 126
.69 87.39 M 276
.58 72.85 A -
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- C1 and C2 are intervals on the containment shell circumference between support locations C and D as shown in Figure C-12.
C-130 l
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. TAB E C-7
'6 COMPAINMENT MODIFICATIONS REQUIRED FOR 80 PSIG CAPABILITY
- 1. - INCREASE THICKNESS T SHELL (EEVATIW 199'4" 'ID 224'0") FRCM 5/8" 'ID 1".
- 2. INCREASE 'IHICKNESS OF SHELL (ELEVATION 162'2" TO 199'4") FROM 7/8" TO 1".
I
- 3. -INCREASE CAPABILITY T EQUIPMENT HA'IUI COVER BY ONE OF THE F0 LINING:
i
-A) INCREASE UlICKNESS PROM l-3/8" TO 1-3/4".
4 i d B) ADD STIFFENERS 'ID PREVENT BUCKLING.
C) REVERSE ORIENTATION SO DIAT PRESSURE W COVER IS INTERNAL PRESSURE.
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C-131
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l REGUIATION 10CFR50.34 (e) (3) (vi) !
(9).
Subject:
External Hydrogen Recombiners 1b satisfy the following requirements, the application shall provide sufficient information to demonstrate that the requirment has been met.
1his information is of the type customarily required to satisfy 10 CFR 50.34 (a) (1) or to address the applicant's technical qualifications and management structure and competence.
For plant designs with external hydrogen recombi.ners, provide redundant dedic *ed contairunent preparations so that, assuming a single failure, the recs ,_.,er systems can be connected to the containment atmosphere.
(II.E.4.1)
OFFSHORE POWER SYSTEMS RESPONSE This requirement does not apply to the Floating Nuclear Plant because the hydrogen recombiners are located inside containment.
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C-136 1 l
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REGUIATION 10CFR50.34 (e) (3) (vii)
/_ 8 h
Subject:
Management of Design and Construction Activities To satisfy the following requirements, the application shall provide sufficient information to denonstrate that the requirament has been met.
'Ihis information is of the type customarily required to satisfy 10 CFR 50.34 (a) (1) or to address the applicant's technical qualifications and management structure and competence.
Provide a description of the management plan for design and construction activities, to include: (A) the organizational and management structure singularly responsible for direction of design and construction of the proposed plant; (B) technical resources directed by the applRant; (C) details of the interaction of design and construction within the appli-cant's organization and the manner by which the applicant will ensure close integration of the architect engineer and the nuclear steam supply vendor; (D) proposed procedures for handling the transition to operation; (E) the degree of top level management oversight and technical control to be exercised by the applicant durirvj design and construction, including the preparation and implementation of procedures necessary to guide the effort.
(II .J . 3.1)
OFFSHORE POWER SYSTEMS RESPCNSE
) Offshore Power Systems was formed in 1972 for the sole purpose of de-signing, manufacturing and marketing Floating Nuclear Plants. For this reason, the organization and quality assurance program described in the Plant Design Report have always dealt entirely with the plant design and manufacturing processes.
The rule requires in part that applicants describe proposed procedures for handling the transition to plant operation. Plant operation is the respon-sibility of the owner and the transition to operation will be addressed in the owner's construction permit and operating license applications. There are, however, ntsnerous ways in sich Offshore Power Systems can assist the owner to prepare for plant operation. Examples include:
- 1. Customer review of plant specifications and other design documents
- 2. Customer hands-on participation in plant pre-operational testing activities
- 3. Customer participation in 0A activities related to plant manufacture o) t V
C-137
- 4. Preparation of the plant data package consisting of test results, @
(x i c records and a.s-built doc mentation G
- 5. Classroom training by OPS personnel covering Floating Nuclear Plant (
Structures and Systems. General training in nuclear technology and in l 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 familiarization in a plant !
virtually identical to their own. Although OPS cannot make comitments for !
its customers, it is likely that a limited number of personnel frm a pro spective FNP owner wuld be welcome to observe and assist in the operation of another Floatirg Nuclear Plant.
Additional information is contained in the response to 10 CFR 50.34 ,
(e) (3) (iii) .
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1 C-138
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