ML20084P671
| ML20084P671 | |
| Person / Time | |
|---|---|
| Site: | 05000601 |
| Issue date: | 11/30/1983 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML19269A120 | List: |
| References | |
| NUDOCS 8405180318 | |
| Download: ML20084P671 (86) | |
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3.0 POST-TMt REQUIREMENTS AND RECOMMENDAtl0NS l
I Shortly af ter the initial recovery phases following the March 26, 1979 incl- l dent at IM1-2, various task forces and investigating groups were set up (both inside and outside of the NHC) to make recofunendations for plant design and j operating changes to ensure that a IM1-2 type event or similar event does not i happen again. The requirements and recommendations f rom these task forces and investigating groups were consolidated and documented in NUREG-0660, "NRC Action Plan Developed as a Result of the TMI-2 Accident." 1his NUREG does not specifically address requirements for new plant designs, since at that time the NRC directed its technical review resources to assuring the safety of f j operating power reactors rather than the issuance of new licenses or permits.
In mid-1980 the NRC staf f initiated a program f or Commission approval of, a f course of action that would lead to the estabilshment of IMI 2 related re - l
! quirements for pending construction permit applications. This program led to I the issuance of NUREG-0718 Revision 2 " Licensing Requirements for Pending i/
Applications for Construction Permits and Manuf acturing License," which speci-
. fies those NRC , Action Plan (NUREG-0660) items that are required to be imple-mented or committed to by a pending applicant prior to receiving a construc-I tion permit or a license to manuf acture. In addition, the NRC has issued a revision to 10CFR 50.34, ' Contents of Applications; Technical Information,"
that essentially incorporates the post-TM! requirements of NUREG-O'718 Into ;
l their regulations. _
i i This revision to 10CFR 50.34 (which is referred to as the CP/Mt Rule) is writ-i ten such that it is applicable to construction permit and manufacturing i
license applications pending at the offective date of the rule (i.e., February i l'6, 1982). However, the " Proposed Commission Policy Statement on Severe Acci-dents and Related Views on Nuclear Reactor Regulation," (48FR16014, April 13, i 1983) indicates that the requirements of the CP/Mt. Rule are also applicable to new construction permit applications or reactivation 5. Therefore, applicable l i
post-TMI requirements of NUREG-0718/10CFR 50.34 and certain additional poten-tial requirements from the NRC Action Plan (NUREG-0660) are being addressed in l !
' the WAPWR design as indicated in the following sections.
i 3.' 0- 1 N0ygpOER, 1983 WAPWR-RC ,
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O 3.1 NUREG-0718/10CFR 50.34 (CP/ML RULE)
The following are the licensing requirements and WAPWR design responses for
,O each NUREG-0718/10CFR 50.34 item that impacts or potentially impacts the WAPWR design.
- 1. Plant / Site Specific Probabilistic Risk Assessment 10CFR 50.34(f)(1)(1)
" 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."
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! Discussiog O
Refer to Section 3.2 which has been devoted to the inter-related 'essues of l(' probabilistic risk assessment, safety goal, and severe accidents.
- 2. Auxiliary Feedwater System Evaluation ,
" Perform an evaluation of the proposed auxiliary feedwater system (AFWS),
to include: (A) a simplified AFWS reliability analysis using event-tree and f ault-tree logic techniques. (8) a design review of AFWS, and (C) an evaluation of AFWS flow design bases and criteria."
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Discussig A conventional AFWS functions, in conjunction with a seismic Category I water source, as an emergency system for the removal of heat f rom the primary system when the main feedwater system is not available. It also v plays an important role in mitigating the of fects of some design basis 3.1-1 NOVEM8;R, 1983 WAPWR-RC 0060e:1
l events (e.g., main f eedwater line breaks and some small break loss-of-coolant accidents). Existing AFWS designs hold the plant at hot standby, or cool down the primary system to temperature and pressure levels at which the low pressure residual heat removal system can operate. The AFWS can also be used during norn.al plant startup and shutdown conditions.
AFWS designs usually consist of a combination of steam turbine-driven and electric motor-driven pumps.
The WAPWR design is somewhat dif ferent than a conventional two electric motor driven and one steam turbine-driven AFWS design.
The # PWH design includes an emergency feodwater system (UWS) and a startup feedwater system (SF*,lS). The EFWS is a safety system utilizing four pumps; two electric motor-driven and two steam turbine-driven. The EFWS functions similarly to a conventional AFWS excep: that during normal plant startup/ shutdown and hot standby the SFWS is utilized. The EFWS 15 designed for such events as main steam line breaks, main feedwater line breaks, steam generator tube ruptures, loss-of coolant accidents, loss of all AC pwer, and any other event in which the main and startup feedwater systems are not available. The SIWS is a control grade system utilizing one motor driven pump and provides feedwater during normal plant startup/
shutdown and hot standby. The SFWS is also started automatically during reactor trips and other anticipated transients.
The purpose of requirement (A) above is to (1) assest the reliability of the AFWS design under various loss of feedwater transient conditions, with particular emphasis being given to determining potential f ailures that could result f rom human errors, common causes, single point vulnerabill-
' ties, and test and maintenance outages, and (2) incorporate design provi-stons and/or procedural actions 45 necessary to improve the MW'a reliabil-ity relative to the NWC generic MWS reliabilities published in NijWLG.
0611, " Generic Evaluation of f eedwater Iranslents and Small Break 1.os s '
of-Coolant Accidents in Westinghouse-Designed Operating Plants."
O WAPWR-RC 3.1 -7 NOVlM8(R, 1g8.1 0060e:1
O A quantitative f ault-tree unavailablitty analysts of the secondary side saf eguards t*l stems (5555) design f or the FM f.at been performed. The WM design was shown to have a reliability higher than the systems evelvated by the MC and de wated in WatG 04tl.
The purpose of requirement (8) above is to (1) attest the level of coe*
pliance of the AM design to the MC acceptance criterie documented in Standard Review Plan 10.4.g. 'Aunillary Peedwater Systeet.* and (2) where deviattent are identified, modify the AM design at necettery to comply with the MC acceptance criteria er justify the deviattelt.
The purpose of requirement (C) above 15 to assure that the design beset and criteria for estat11 thing AM requirements for flew to the steam generater(s) to alture adequate removal of reactor decay heat are defined and documented.
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in regard to requirement ( A), when the final details of the 5555 dettgn for the FM are estabilthed. the rellettlity reanalytit (discutted above) will be perf ormed. The final 5555 rellettlity analytts will to tuhmitted to the MC 45 part of the licensing procett for the W M detign.
In regard to regetroment (3), tiettinghevne will completely document and justify any deviattent f rom the MC Standard Review Plan 10.4.1 accep-tance criteria during the licensing procell for the FM destge, in regard to requirement (C). httinghouse will completely document an evaluation of the 5555 flow design bases and criterta during the licen-sing procent for the F M detion.
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- C 3.1 1 MVtM8(N. 1983 5
- 3. Reactor Coolant Pump Sealt 10CFR 50 aaff)(1)(iii)
' Perform an evaluation of the potential for and impact of reactor coolant pump teal damage following small-break LOCA with lots of of fitte power.
If damage cannot be precluded, provide an analysis of the limiting small-break LOCA with subsequent reactor coolant pump teal damage."
Qhtuttioti O
Within the design bases of current Westinghouse plant designs, the scenar-to postulated in this regulation does not present a problem. During nor-mal operation, seal injection f rom the chemical and volume control nyttem 11 provided to cool the reactor coolant pump tealt and the component cool-ing water system provides flow to the thermal barrier heat enchanger .to limit the heat trentfor from the reactor coolant to the reactor coolant pump internals, in the event of a lost of of f ttte power the reactor cool-ant pu.ap motor 11 de energlied, the diesel generatort are automatically started and component cooling water to the thermal barrier heat enchanger and/or teal injection flow is automatically rettore1 within seconds.
(Ither of these cooling suppliet 11 adequate tt provide seal cooling and prevent teal failure due to a lost of offitte power.
In addition to the normal seal cooling provided in conventional designs.
the HAPWR design includet upgraded teat injection capability whichi pro-videt an alternate tource of seal injection water to the reactor coolant pumpt during tituations involving the lost of both normal seal injection and thermal barrier cooling. '.uch situationt are beyond the postulated lost of offttte power of the above regulatton and involve multtple f ailuret/ operator errors or comaon mode f ailures, f or the WAPWN design, the adili t ion of the upgraded seal injection capability providet edded alturance of maintaining taal injection cooling.
O W4PWM NC 3.1 4 NOVlMHlN. lifil 0060eti
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WAPWR Response in relation to this regulation, normal reactor coolant pump seal injection for the WAPWR design is adequate and no additional evaluations will be iO performed.
- 4. Automatic PORY Isolation System
" Perform an analysis of the probability of a small-break LOCA caused by a stuck-open PORV. If this probability is a significant contributor to the probability of small-break LOCA's from all causes, provide a description and evaluation of the ef fect on small-break. LOCA probability of an auto-matic PORV isolation system that would operate when the reactor coolant system pressure falls af ter the PORY has opened.'"
ll f, Discussion -
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4 General Design Criterion -14 " Reactor Coolant Pressure Boundary," of Appendix A to._10CFR Part # 50 " requires that the reactor coolant pressure boundary be designed, fabr'icated, erected, and tested to have an extremely
' icw probability , of abnormal leakage, rapidly propagating failure, and gross rupture. Historically, the application of 'this criterion has emphas5 zed ' the integrity of passive componenSs in the reactor coolant system, such as the reactor vessel and the piping, however, this criterion t also applies to the valves that provide isolation for the system.
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- The primary purpose of pressurizer relief ;and safety . valves is that they i operate in conjunction with 'the reactivity control system't'o. limit system overpressure duEinq antickpated J operational transientbor accidents. The 3
! pressur'izer relief valves are not part of ASME Code requirements for over-pressure ' protection and,- therefore, they can be and are .isolatable with remote-opera'ted b}ock valves. The consequence of the failure of the pres-
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9 surizer. relief valves to close is the loss of coolant .and depressurization l
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NOVEMBER,.1983
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I of the reactor system. This consequence can be mitigated if the remote-operated block valves are closed either automatically or by operator action.
The purpose of this requirement is to evaluate (using probabilistic tech-niques) the benefit of incorporating an automatic pressurizer PORV isola-tion system.
Westinghouse (in support of the Westinghouse Owners Group) has performed an evaluation of the henefit of ' incorporating an automatic pressurizer PORV isolation syste;a for conventional plant designs. This evaluation (which is documented in WCAP-9804, "Probabilistic Analysia and Operational Data in Response to NUREG-0737 Item 11.K.3.2 for Westinghouse NSSS Plants") concluded that such a system should not be required. This' con-l clusion was primarily based on the reduction of the already small PORY LOCA probability due to implementation of changes to plant designs subse-quent to the TMI-2 event. These changes include both modifications which make PORV challenges less likely and changes in hardware, procedures, and training which provide assurance th~at the function of PORY isolation will -
be reliably performed by operator action. As further justification of i
this conclusion, failure to isolate stuck-open PORVs has been analyzed and the results predict no core uncovery.
The WAPWR also contains several design features which will minimize chal-lenges to the PORVs. First, the charging pumps are independent of the
, safety inje: tion system and second, the sizing of the pressurizer is such that the PORVs will not open even under a full load, rejection.
WAPWR Response In regard to the WAPWR design, Westinghouse is f urther evaluating the benefits of an automatic low pressure closing feature for the pressurizer block valves. This feature is being considered in the overall design in accordance with safety-grade cold shutdown and overpressure protection requirements. Inclusion or exclusion of this feature will be completely documented and justified during the licensing process for the WAPWR design.
WAPWR-RC 3.1 -6 NOVEMBER, 1983 0060e:1
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S. Hydrogen Control Systems Evaluation 10CFR 50.34(f)(1)(xii)
O " Perform an evaluation of alternative hydrogen control systems that would satisfy the requirements of paragraph (f)(2)(ix) of this section (50.34).
As a minimum include consideration of a hydrogen ignition and post-accident inerting system. The evaluation shall include: (A) a comparison O' of costs and benefits of the alternative systems considered. (B) for thi selected system, analyses and test data to verify compliance with the requirements of (f)(2)(ix) of this section ( ti0. 34) , and (C) for the selected system, preliminary design descriptions of equipment, function, and layout."
i Discussion Refer to item 14 of this section for a discussion of this requirement in
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conjunction with the requirements of 10CFR 50.34(f)(2)(ix).
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- 6. Simulator Capability 1
" Provide simulator capability that correctly models the control room and includes the capability to simulate small-break LOCA's."
Discussion l
Beyond the above regulation,10CFR Part 55, Appendix A, "Requalification l
i Programs for Licensed Operators of Production and _ Utilization Facilities,"
permits and encourages the use of simulators for operator training. This is due to the undesirability of imposing additional challenges to the plants protective features that would result ,1f the actual plant is used q for training operators to respond to accidents.
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3.1 -7 NOVEMBER, 1983 WAPWR-RC 0060e:1
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1 The purpose of this NRC requirement is to: (A) require simulator capabil- O!
ity, and (B) ensure that the proposed simulator capability for training of operators is performed on a simulator that correctly models the actual plant specific control room design and has the capability to accurately simulate a small-break LOCA.
In addition, the NRC has issued Regulatory Guide 1.149, " Nuclear Power Plant Simulators for Use in Operator Training," which basically endorses ANSI /ANS 3.5-1981, " Nuclear Power Plant Simulators for Use in Operator Training," and describes a method acceptable to the NRC staf f for specify-ing the functional requirements of a nuclear power plant simulator to be used for operator training.
WAPWR Response g
This requirement does not impact the WAPWR design. Simulator capability is the responsibility of each utility utilizing the WAPWR design. -
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- 7. Plant Procedures 10CFR 50.34(f)(2)(ii)
" Establish a program to begin during construction and follow into opera-tion, for integrating and expanding current ef forts to improve plar.t pro-cedures. The scope .of the program shall include emergency procedures, reliability analyses, human factors engineering, crisis management, opera-tor training, and coordination with INPO and other industry efforts."
Discussion The area of operating procedures has received great attention as a result l
of the TMI-2 event. This attention Stems f rom certain opinions that the severity of the IMI 2 event might have been significantly reduced if the O
I WAPWR-RC 3.1 -8 NOVEMBER, 1983 0060e:1
l operating procedures were better written (human engineered and supported j by appropriate analyses) and if the operators were better trained in the use of the procedures.
I Since the TM1-2 event there have been extensive industry ef forts under-I taken to improve emergency operating procedures and their use. For example. Westinghouse (in support of the Westinghouse Owners Group) nas:
(A) reviewed. and revised the generic Westinghouse Emergency Response Guidelines as a result of new snell-break LOCA analyses, inadequate core cooling analyses, transient and accident analyses, discussions with the
< NRC (and subsequent NRC reviews), and inputs from utilities, (8) estab-lished a program for additional inputs or revisions to the generic Emer-gency Response Guidelines as a result of ongoing ef forts, and (C) initiat-
! ed a human factors test of the new Emergency Response Guidelines to deter-mine any problem areas in an operating environment.
As one would expect, these ef forts to date have been' focused on current-day operating and near-term operating plants. The NRC concern that
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resulted in the above requirement is that programs for the continued improvement of plant operating procedures should be pursued and coordin-i ated with other industry ef forts (e.g., INPO) and other post-TMI related improvements (e.g. , safety parameter display systems) in relation to new
! applications.
Although the generic Westinghouse Emergency Response Guidelines have undergone extensive review and revision since the TMI event and are believed . to be a well defined and analytically supported basis for the development of plant specific operating procedures, the current generic guidelines are not expected to be totally applicable to the WAPWR design as a result of differences from conventional designs.
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3.1-9 NOVEMBER, 1983 t!APWR-RC 0060e:1
l WAPWR Response An important aspect of the WAPWR design is to allow the experienced gained in the development of the generic Emergency Response Guidelines to influ-ence the design of specific WAPWR systems.
Specihc task analyses will be performed for the WAPWR design at an early enough time in the development program to allow interaction with the design process such that any design improvements identified can be factored into the WAPWR systems.
For licensing purposes Westinghouse will outline a program for emergency response guideline development prior to receiving a preliminary design approval for the WAPWR design. Prior to issuance of a final design approval (and in a timely manner that permits verification, possible NRC review, and possible operator training) Westinghouse will develop the actual WAPWR Emergency Response Guidelines.
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- 8. Control Room Design .
" Provide, for Commission review, a control room design that reflects state-of-the-art human factor principles prior to committing to f abrica-tion or revision of fabricated control room panels and layouts."
Discussion General Design Criterion 19 " Control Room," of Appendix A to 10CFR Part 50 requires that a control room be provided f rom which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions.
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WAPWR-RC 3.1-10 NOVEMBER, 1983 0060e:1
O For current-day operating plant licensees and applicants, this item is being implemented as a detailed review of their control room designs with the purpose of correcting weaknesses to improve the ability of control room operators to prevent accidents or cope with accidents if they occur.
The NRC has issued guidance for performing control room design reviews in the form of NUREG-0700, " Guidelines for Control Room Design Reviews."
NUREG-0700 is wnitten specifically for existing control room designs and new guidance or criteria may be issued in the future for new control room O designs.
Again for existing control room designs, the NRC has issued draf t accep-tance criteria for control room design reviews which is - documented in NUREG-0801 " Evaluation Criteria for Detailed Control Room Design Review." NUREG-0801 includes guidelines for the organizational structure and personnel qualifications for performing control room design reviews as l
well as guidelines for the actual review process and results documenta-tion.
'i " This requirement', as written, simply states that the control ro'om design .
must be submitted to the NRC for review prior to fabrication. Inherent with this requirement is that it must be demonstrated to the NRC that the control room design meets applicable licensing criteria (e.g., human factors engineering, new instrumentation requirements, etc.).
Key to the overall issue of ensuring a good control room design is the fact that the're ,are numerous current-day licensing issues that impact the control room design. The NRC has indicated through NUREG-0737, Supplement 1, " Requirements for Emergency Response Capability," that these require-ments (including the safety parameter display system, Regu l'a tory Guide 1.97 instrumentation, emergency operating procedures, etc.) should be O integrated with respect to the overall enhancement of the operators abili-ty to comprehend plant conditions and cope with potential emergencies.
O 3.1-11 NOVEMBER, 1983 (dAPWR-RC 0060e:1
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WAPWR Response The WAPWR saf ety analysis report will include a section or chapter des-cribing the control room design and its conformance to applicable criteria.
The overall control room design for the WAPWR will integrate the require-ments of this regulation concerning human factors principles with the requirements of the various regulations concerning control room instrumen-tation (e.g., the instrumentation required by items 9, 10, 21, 22, and 23 below).
" 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 displaying a ful.1 range of important plant parameters and data trends on demand, and capable of indicating when process limits are being approached or exceeded."
Discussion The purpose of the plant safety parameter display console (or safety para-meter display system) is to provide a concise display of critical plant variables to control room personnel in order to assist them in rapidly and reliably determining the safety status of the plant. Although not speci-fically mentioned in the above regulation, the NRC is recommending that j the licensee consider duplication of the safety parameter display console displays in the onsite technical support center and the near-site emer-l gency operations facility to improve the exchange of inf ormation between these facilities and the control room and assist corporate and plant man-agement in the decision-making process.
O WAPWR-RC 3.1-12 NOVEMBER, 1983 0060e:1.
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O In general, this requirement is no dif ferent from that currently being implemented by operating plant licensees and applicants in response to NUREG -0737, Supplewnt 1 " Requirements for Emergency Response Capabil-ity." For certain operating plant licensees and applicants the Westing-house designed plant safety status display system, as described in WCAP-9725 (including Supplement 1), " Westinghouse Technical Support Complex,'" is being installed in the onsite technical support center and f the nearsite emergency operations facility as well as the control room to
'f satisfy this requirement.
i The NRC has also issued NUREG-0696, " Functional Criteria for Emergency Response Facilities," which provides certain guidance information for the implementation of a safety parameter display system. In addition, the NRC has issued draf t human factors acceptance criteria for safety parameter i display systems which are documented in NUREG-0835, " Human Factors Evalu-
! ation Criteria for Safety Parameter Display Systems."
1 WAPWR Response
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The human factors principles applied to the WAPWR control room design require that a task analysis be performed, and the output from this- task f
analysis wil.1 determine the nature of the safety status display. Control room instrumentation for the WAPWR design will be fully integrated with other control room requirements as discussed in item 8 above, i
During the licensing process for the @PWR design, Westinghouse . will O demonstrate the level of conformance of the WAPWR design to the NRC guidance documented in NUREG-0737 (Supplement 1), NUREG-0696 and NUREG-f 0835 and/or other applicable documents.
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- 10. Safety System Status Indication 10CFR 50.34(f)(2)(v)
" Provide for automatic indication of the bypassed and operable status of safety systems."
Discussion 10CFR 50.55a(h) requires that protection systems meet the requirements set O
forth in IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations." Section 4.13 of IEEE Standard 279-1971 requires that, if the protective action of some part of the protection system has been bypassed or deliberately rendered inoperative i for any purpose, this fact shall be continuously indicated in the control room.
The intent of this requirement is to provide the operator with an auto-matic indication of the bypassed or inoperable status of systems and components that perform a function important to saf ety in accordance with Regulatory Guide 1.47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems." It should be noted that Regulatory Guide 1.47 does permit certain limited use of manual activation of system-level indicators.
The NRC has also issued Branch Technical Position ICSB 21, " Guidance for Application of Regulatory Guide 1.47 " which provides (as its title suggests) additional NRC guidance for implemtation of Regulatory Guide 1.47.
Westinghouse has developed a bypassed and inoperable status indication system as part of the overall Westinghouse designed technical support complex. This system is described in WCAP-9725 (including Supplement 1),
" Westinghouse Technical Support Complex," and is currently being installed O
WAPWR-RC 3.1-14 NOVEMBER, 1983 0060e:1
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l by certain utilities in their onsite technical support centers as well as their control rooms. The Westinghouse bypassed and inoperable status indication system provides primary status display of the systems compris-
[ ing the engineered safety features and supporting displays of individual components within each system or subsystem.
@ PWR Response lh V A bypassed and inoperable status indication system will be included in the WAPWR control room design.
The specific nature of this system will be i determined f rom the task analysis described in item 9, above. Control
' room instrumentation for the WAPWR design will be integrated with other control room instrumentation requirements as discussed in item 8 above.
During the licensing process for the WAPWR design, Westinghouse will demonstrate the level of conformance of the FPWR design to the NRC regulatory positions and acceptance criteria documented in Regulatory Guide 1.47 and Branch Technical Position ICSB 21.
- 11. Reactor Coolant System High Point Vents t
" Provide the capability of high point venting of noncondensible gases from the reactor coolant system, and other systems that may be required to maintain adequate core cooling. Systems to achieve this capability shall O be capable of being operated f rom 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."
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3.1 -15 NOVEMBER, 1983 WAPWR-RC 0060e:1 I
l Discussion .
10CFR 50.46(b)(5) requires that af ter any calculated successful initial ;
operation of the emergency core cooling system, the calculated core temp-erature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by long-lived radioactivity remaining in the core. Additionally, General Design Criterion 35, " Emergency Core Cooling," of Appendix A to 10CFR Part 50 requires that a system to provide abundant emergency core cooling shall be provided. The system safety function shall be to transfer heat f rom the reactor core following any loss of reactor coolant at a rate such that:
(A) fuel and clad damage that could interfere with continued effective core cooling is prevented, and (B) clad metal-water reaction is limited to negligible amounts.
During ' the TMI-2 accident, a condition of low water level in the reactor vessel . and inadequate core coolirg existed and was not rectified for a long period of time. The resultant high core temperatures produced a metal-water reaction with the subsequent production of significant , amounts of hydrogen. The collection of noncondensible gases impaired natural cir-culation cooling capability. Additionally, the collection of nonconden-sible gases limited reactor coolant pump operational capability because of coolant voids in the system occupied by the gases. Even when reactor coolant pump operation was possible, the installed plant venting system was, capable of removing the noncondensible gases only through an extremely slow process.
The purpose of this requirement is to provide for the capability of reac-O tor' coolant system high point venting of noncondensib'le gases collected in the system in order to allow satisf actory long term core cooling.
The above 10CFR 50.34 regulation must be considered in conjunction with O
the recent requirements of 10CFR 50.44(c)(3)(ill). This regulation, which is part of the NRC interim requirements related to hydrogen control, also mandates the installation of high point vents.
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WAPWR -RC 3.1-16 NOVEMBER, 1983 0060e:1
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"To provide improved operational capability to maintain adequate core cooling following an accident, . . . each light-water nuclear power reac-i tor shall be provided with high point vents for the reactor coolant l,
system, for the reactor vestel head, and for other systems required to maintain adequate core cooling if the accumulation of noncondensible gases would cause the loss of function of these systems. (High point vents are for the tubes in U-tube steam generators.) The I
not required, ' oweve'r, h
Since high point vents must be remotely operated f rom the control room.
l, the vents form a part of the reactor coolant pressure boundary, the design of the vents and associated controls, instruments and power sources must conform to the requirements of Appendix A and Appendix 8 of this part j
(10CFR Part 50). In particular, the vent systen shall be designed to j
ensure a low probability that: (A) the vents will not perform their safe-l ty functions, and (8) there would be inadvertent or irreversible actuation
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Furthermore, the use of these vents during and following an-1 of a vent.
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accident must not aggravate the challenge to the containment or the course of the accident."
! \' ' Reactor coolant system high point venting for Westinghouse designs is limited to the reactor vessel and pressurizer. This requirement is no l
different than that currently implemented or being implemented by operat-l ing plant licensees' and applicants. In general, operating plant licensees l l
have installed add-on reactor vessel and pressurizer venting systems.
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Certain operating plant applicants have incorporated design modifications.
i prior to the TMI-2 event, related to safety-grade cold shutdown capability f
I that include the addition of a safety-grade reactor vessel head venting system and a safety-grade upgrade to the pressurizer venting path (i.e.,
Therefore, Westinghouse power operated relief valves and block valves). '
plants designed with safety grade cold shutdown capability were not impacted by these regulations. ,t O l 1
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WAPWR Response Safety-grade cold shutdown capability, including safety-grade reactor vessel and pressurizer venting paths to the pressurizer relief tank, are incorporated in the WAPWR design.
- 12. Plant Shielding 10CFR 50.34(f)(2)(vii)
" Perform radiation and shielding design reviews of spaces around systems that may, as a result of an accident, contain TID-14844 source term radio-active materials, and design as necessary to permit adequate access to important areas and to protect safety equipment f rom the radiation envir-onment."
Discussion 10CFR Part 20 and General Design Criteria 19, 60, and 64 of Appendix A to 10CFR Part 50 require the control of radiation exposure associated with plant operations. General Design Criterion 4, " Environmental and Missile Design Bases," requires that systems and components important to safety be designed to accomodate the environmental conditions associated with accidents.
After an accident in which significant core damage occurs, the radiation source terms may approximate those of Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential Radiological Consequences of a loss-of-Coolant Accident for Pressurized Water Reactors." In addition, systems that were not designed to contain large radiation sources may become highly radioactive. The resulting radiation fields may make it dif ficult i to ef fectively perform accident recovery operations or may impair safety equipment. Currently, Westinghouse is participating in the NRC/ nuclear I industry ef fort to more accurately define the source terms based upon the information obtained as a result of the TMI incident.
i WAPWR-RC 3.1-18 NOVEMBER, 1983 0060e:1
l~
WESTINGHOUSE PROPRIETARY CLASS 2 i
I
!O l The purpose of this requirement is to facilitate post-accident operations using systems that may contain abnormally high levels of radioactivity and to ensure that safety equipment in proximity to the resulting radiation
- fields is not unduly degraded.
Current NRC guidance for performing radiation and shielding design reviews
> is detailed in Item II.S.2 of NURES-0737, " Clarification of TMI Action Plan Requirements." Basic in this guidance is that the reviews should identify the location of vital areas and equipment (such as the control room, onsite technical support center, sampling station and sample anal-ysis area, containment isolation reset control area, security center, j radwaste control stations, emergency power supplies, motor control f centers, and instrument areas) in which personnel occupancy may be unduly limited or safety equipment may be unduly degraded by the radiation fields during post-accident operations of these systems. These design reviews l are intended to identify corrective actions (e.g., design changes) neces-
{ sary to provide for adequate access to vital areas and protection of l\ safety equipment.
4 j An important design feature of the @PWR Primary Safeguard System (PSS),is the reduction and modularization of the post-accident recirculation equip-ment located outside the primary containment building. The F PWR PSS I consists of four separate modularized subsystems which are located in four l physically separate and independent safeguard component areas. .This system configuration and related layout /HVAC ' arrangement facilitates O
pnst-accident recirculation operation / maintenance and provides a signifi -
cant improvement in the access to and the separator / shielding between .the PSS recirculation systems and other nonradioactive vital areas.
O WPWR Response Radiation and shielding evaluations will be performed (and documented with the NRC during the licensing process) for those vital areas and equipment included in the' overall PPWR design. rhe results of these evaluations WPWR-RC 3.1-19 NOVEM8ER, 1983 r,m
i l
will demonstrate adequate access to vital areas and adequate protection of safety equipment or the design will be modified. As a practical matter.
provisions for adequate shielding are being considered in the early phases l
of establishing the WAPWR plant layout.
- 13. Post-Accident Sampling l
" Provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain T10-14844 source l term radioactive materials without radiation exposures to any individual l 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, iodines and cesiums, and non-volatile isotopes), hydrogen in the containment atmos-phere, dissolved gases, chloride, and boron concentrations."
i Discussion -
Prompt sampling and analysis of reactor coolant and the containment atmosphere can provide information important to the ef forts to assess and control the course of an accident. Chemical and radiological analysis of reactor coolant liquid and gas samples can provide substantial information regarding core damage and coolant characteristics. Analysis of contain-ment atmosphere (air) samples can determine if there is any prospect of hydrogen reaction in containment, as well as provide core damage informa-tion.
Beyond the above requirement, no definitive regulations exist for obtain-ing and analyzing reactor coolant or containment samples following an accident. NRC guidance and acceptance criteria (e.g. Standard Review Plan.
9.3.2 and Regulatory Guide 1.97) have, however, been revised since the TMI-2 event to require this capability.
1 WAPWR-RC 3.1-20 NOVEMBER, 1983 0060e:1
Most recent plant designs include (or are being modified to include) post-accident sampling capability through the use of in-line monitoring systems. These systems usually have the capability to obtain samples from the reactor coolant system hot legs, the containment recirculation sump, and the containment atmosphere. The time re' quired for taking and analy-zing samples (in an onsite radiological and chemical analysis facility) must be 3-hours or less from the time a decision is made to sample, except for chloride which is 24-hours or less. ,
In addition to in-line monitoring systems, backup sampling is require to be available through grab samples. Capability of analyzing these samples must be demonstrated (established planning for analysis at offsite facili-ties is acceptable to the NRC).
l The above regulation 'also requires that radiation exposures to those individuals performing sampling and analyses be limited to acceptable l values. Facility, system, and shielding design must be such that person-
/ nel exposure is minimized.
(.
WAPWR Response The WAPWR plant design will include in-line sampling capability as well as grab sample capability. An onsite radiological and chemical analysis capability will be considered in the WAPWR design consistent with the
! scope definition for the Nuclear Power Block.
O 1During the WAPWR licensing process, Westinghouse will:
r o Demonstrate compliance with all applicable requirements of NUREG-0737 (Item II .B.3) for sampling, chemical, and radionuclide anal-ysis capability under accident conditions.
l o Demonstrate that sufficient shielding is provided to meet the requirements .of General Design Criterion 19, assuming Regulatory .
I Guide 1.4 (TID-14844) source terms.
3.1 -21 NOVEMBER, 1983 WAPWR-RC 0060e:1
l o Demonstrate compliance with the sampling and analysis requirements of Regulatory Guide 1.97, Revision 3, in accordance with the over-all WAPWR post-accident monitoring design that addresses this regulatory guide as discussed in item 23 below.
o Demonstrate that all electrically powered components associated with post-accident sampling are capable of being supplied with power and operated within 30-minutes of an accident in which there is core' degradation, assuming loss of of fsite power.
o Demonstrate that any valves associated with post-accident sampling which are not accessible for repair af ter an accident are environ-mentally qualified for the conditions in which they must operate.
o Provide a procedure for relating radionuclide gaseous and ionic species to estimated core damaged.
o Demonstrate the design or operational provisions to prevent high pressure carrier gas f rom entering the reactor coolant system f rom
,in-line ga.s analysis equipment.
o Demonstrate a method for verifying that reactor coolant dissolved oxygen is at < 0.1 ppm if reacto.* coolant chlorides are deter-mined to be > 0.15 ppm.
o Provide information on: (A) testing frequency and type of testing to ensure long-term operability of the post accident sampling system, and (B) recommended operator training requirements for post-accident sampling.
O l
l O
WAPWR -RC 3.1 -22 NOVEMBER, 1983 0060e:I l
i 6
- 14. Hydrogen Control 10CFR 50.34(f)(2)(ix) i O " Provide a system for hydrogen control that can safely accommodate hydro-gen generated by the equivalent of a 100 percent fuel-clad metal -water reaction. Preliminary design information on the tentatively preferred
~
system option of those being evaluated in pa. s aph (1)(xil) of this l permit stage. The section (50.34) is sufficient at the construc.
4 hydrogen control system and associated systems ..411 provide, with reasonable assurance, that: ,
- (A) Uniformly distributed hydrogen concentrations in the containment do f
i not exceed 10 percent dur'ing and following an accident that releases I an equivalent amount of hydrogen as would be generated f rom a 100 l percent fuel-clad metal-water reaction, or that the post-accident atmosphere will not support hydrogen combustion.
l I- (8) Combustible concentrations of. hydrogen will not collect in areas where
) unintended combustion or detonation could cause loss of containment l integrity or loss of appropriate mitigating features.
L
) (C) Equipment necessary for achieving and maintaining safe shutdown of the plant and maintaining containment integrity will perform its safety function during and af ter being exposed to the environmental condi-i tions attendant with the release of hydrogen generated by the equival--
ent of a 100 percent fuel-clad . metal-water reaction including the environmental conditions created by activation of the hydrogen control system. .
(0) If - the method ' chosen f or hydrogen control is a post-accident inerting system, inadvertent actuation of the system can be safety accommodated during plant operation."
O 9 3.1-23 NOVEMBER, 1983
! )(APWR-RC ,
0060e:1-
Discussion The accident at TM1-2 resulted in a severely damaged or degraded reactor core with a concomitant releas" of radioactive material to the primary coolant system and a large amount of fuel cladding metal-water reaction in the core with hydrogen generation well in excess of the amounts required to be. considered for design purposes by historical Commission regula-tions. The accident revealed design and operational limitations that existed relative to mitigating the consequences of the accident and deter-mining the status of the facility during and following the accident.
10CFR 50.44(c)(1) requires that it be shown that during the time period following a LOCA but prior to ef fective operation of the combustible gas control system either: ( A) an uncontrolled hydrogen-oxygen recombination would not take place in the containment, or (B) the plant could withstand the consequences of uncontrolled hydrogen-oxygen recombination without loss of safety function. If these conditions cannot be shown, the con-tainment is required to be provided with an inerted or an oxygen deficient atmosphere in order to provide protection against hydrogen burning an1 explosions.
For operating plant licensees and applicants prior to the TMI-2 event, the NRC is proposing regulations similar to those contained in 10CFR 50.34(f)
(1)(xii) and 10CFR 50.34(f)(2)(ix). The najor differences between the proposed rules for existing plants and the effective rules for new plants
( are that:
I o The uniform hydrogen concentration in the containment must not O
exceed 10 percent by volume during and following a degraded core l accident for new plants. The proposed rules for existing plants do not impose such a limit on nydrogen concentration.
o The amount of hydrogen to be considered for new plants is equival-ent to that generated f rom the reaction of 100 percent (versus 75 percent for existing plants) of the fuel cladding surrounding the actual fuel region.
l WAPWR-RC 3.1-24 NOVEMBER, 1983 0060e:1
I I
m i
For new plant designs a suitable hydrogen control system will be required
! to meet this regulation, whereas no hydrogen control system is needed for large dry containments to meet the proposed regulations of the interim I rule for existing plants.
! Among the various hydrogen control systems evaluated by the industry thus
- far, a hydrogen ignition system appears to be the best choice. A hydrogen ignition system is relatively inexpensive, easy to test, and inadvertent actuation of the system during normal plant operation will not result in any adverte effects.
It should be noted that the FPWR design will dif fer from a conventional j
pressurized water reactor in that there will be a significant increase in i
the amount of Zircaloy utilized in the design. This increase in Zircaloy I is not simply related to an increase in the amount of fuel cladding pres-ent (i.e., due to the larger core) but results from a combination of other design features.
@ PWR Response ,
Westinghouse will perform an evaluation of alternative hydrogen control f
l systems for the @PWR design. The system selected for use in the FPWR l
design will be in accordance with the above requirements and fully docu-( mented during the licensinJ process for the F PWR design. ,
Westinghouse will perform aki calculations and analyses considering the O additional Zircaloy in the @PWR design rather than restricting the cal-culations to 100 percent of the fuel cladding as required by this regula- I tion.
O l
l O
l FPWR-RC 3.1-25 \ NOVEMBER, 1983 0060e:1 L
! l
- 15. Reactor Coolant System Valve Testing 10CFR 50.34(f)(2)(x)
" Provide a test program and associated model development 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 con-ditions, transients and accidents. Consideration of anticipated tran-sients without scram (ATWS) conditions shall be included in the test pro-gram. Actual testing under ATWS conditions need not be carried out until subsequent phases of the test program are developed."
Discussion General Design Criteria 14, 15, and 30 of Appendix A to 10CFR Part 50 require that the reactor coolant pressure boundary be designed, fabri-cated, and erected to the highest quality standards and be tested to ensure an extremely low probability of abnormal leakage, rapidly propa-gating failure, and gross rupture. - These criteria also require that the design conditions of the reactor coolant pressure boundary not be exceeded during any condition of normal operatian, including anticipated operation-al occurrences.
Proper operation of the reactor coolant system relief, safety, and block valves is necessary for conformance to these design criteria. The inabil-ity of these valves to open or close could 1 ad to a violation of the integrity of the reactor coolant pressure boundary.
When the reactor coolant system relief and safety valves open, the flow through these valves is normally saturated steam. Some reactor coolant system transients and accidents as well as alternate core-cooling methods can result in solid-water or two -phase steam-water flow through these l valves. Historical qualification requirements for these valves included only flow under saturated steam conditions.
l Ol l
WAPWR-RC 3.1-26 NOVEMBER, 1983 0060e:1
i l
The purpose of this regulation is to require qualification of reactor
! coolant system relief, safety, and block valves under expected operating conditions (including solid-water and two-phase flow conditions) and AlWS
( conditions.
Generic reactor coolant system valve testing (sponsored by EPRI) has been i
conducted in support of operating plant licensees and applicants. The EPRI program included representative testing of Westinghouse reactor cool-ant system valve types at representative '.uid conditions including solid-water and two-phase flow conditions, ihe EPRI program did not, however, include specific consideration of ATWS conditions.
Operating plant licensees and applicants have submitted documentation to demonstrate applicability of the generic EPRI test results to their plant '
specific reactor coolant system valves, their plant specific expected fluid conditions, and their plant specific piping and support i
configurations.
'L I O LdAPWR Response The generic EPRI test results discussed above are expected to be directly applicable to the LdAPWR design, since the latest Westinghouse pressurizer power-operated relief valves and safety valves were included in the test program.
i Westinghouse will document the applicability of the generic EPRI test O results to the LdAPWR design (including valve designs, piping and support designs, and fluid conditions) during the licensing process for the LdAPWR design. If the generic EPRI test results do not envelope the specific.
LdAPWR design, Westinghouse will either: (A) perform additional testing, O or (B) demonstrate justification for not performing additional testing possibly through additional analyses and/or evaluations.
l lO 3.1 -27 NOVEMBER, 1983 LdAPWR-RC 0060e:1
i l
- 16. Valve Position Indication l
" Provide direct indication of relief and safety valve position (open or O closed) in the control room."
i Discussion This regulation is written in very general terms. A review of the NRC background material in relation to this regulation (i.e., NUREG-0578 NUREG-0660, NUREG -0718, and NUREG-0731) indicates that this requirement for valve position indication applies to reactor coolant system relief and safety valves.
General Design Criterion 14, " Reactor Coolant Pressure Boundriry , " of Appendix A to 10CFR Part 50 requires that the reactor coolant pressure boundary be designed, fabricated, erected, and tested to have an extremely low probability of abnormal l eaka~ge, rapidly propagating failure, and-gross rupture. Historically, the application of this criterion has emphasized the integrity of passive components in the reactor coolant system, such as the reactor vessel and the piping, however, this criterion also applies to the valves that provide isolation for the system. Failure of relief and safety valves to close can cause events that result in small -break los s -of -coolant accidents. Unambiguous indicatir. of the position of the valves can aid the operator to detect a failure and take proper corrective action.
The purpose of this requirement is to provide the control room operator a positive indication of valve position and, therefore, provide additional anurance that the integrity of the reactor coolant pressure boundary can be maintained or a loss of integrity directly diagnosed.
I l
l WAPWR -RC 3.1 -28 NOVEMBER, 1983 l 0060e:1
i i -
f Conventional Westinghouse designs include a positive control room position indication for the presssurizer power-operated relief valves (i.e., indi-cation lights which are activated by limit switches). Conventional safety
. valve designs have been upgraded to provide position indication through f stem-mounted limit switches or acoustic monitoring of flow downstream of i
the valves.
4 -
' The above regulation, as written, requires a direct valve position indica-tion and, therefore, the option of flow indication through utilization of an acoustic monitoring system does not satisfy this requirement.
WAPWR Response l
The MAPWR design will incorporate positive control room position indica-tion for the pressurizer power-operated relief valves and safety valves.
- 17. Auxiliary Feedwater System Initiation and Indication
" Provide automatic and manual auxiliary feedwater system initiation, and
! provide auxiliary feedwater system flow indication in the control room."
Discussion In Westinghouse designs the auxiliary feedwater system (AFWS) has been O treated as a safety system. It is used to remove' heat f rom the reactor system when the main feedwater system is not available.
i The need to automatically initiate the operation of the AFWS was not
, Os considered by all vendors to be essential to safety in the past, and in some plants dependence was placed on the operator to put the system in O
3.1 -29 NOVEM8ER, 1983 l MAPWR-RC l 0060e:1
service when required. Although this need was not emphasized, the initia-tion of the AFWS is automatic in conventional Westinghouse desi]ned plants l in accordance with General Design criterion 20 " Protection System Func-tions," of Appendix A to 10CFR Part 50.
General Design Criterion 13, " Instrumentation a 1 Control." sets forth the requirements for instrumentation to monitor *e . variables .and systems, over their anticipated ranges of operation, t' 4t can af fect reactor saf e-ty. Auxiliary feedwater flow indicatie- o the steam generators is considered an important adjunct to the manual regulation of auxiliary feedwater flow to maintain the required steam generator level and Westing-house has reconnended that this indication be included in plant designs.
WAPWR Response The secondary side safeguards function for the WAPWR desigri will dif fer from conventional designs as discussed in detail in item 2 above.
The secondary side safeguards function for the WAPWR design includes manual and automatic initiatirn of flow to the steam generators. Heat removal indication (i.e., flow and level) in the control room will be pro-vided by Class IE indicators in accordance with the above regulation and the guidance of Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" (refer to item 23 below).
i i 18. Pressurizer Heater Power Supplies 10CFR 50.34(f)(2)(xili)
{
" Provide pressurizer heater power supply and associated motive and control power interf aces suf ficient to establish and maintain natural circulation in hot standby conditions with only onsite powei available."
O WAPWR-RC 3.1-30 NOVEMBER, 1983 0060e:1
l .
l i
} )
1-
!O j 01scussion t
l l-j Pursuant to NRC regulations in 10CFR Part 50. Appendix A. " General Design !.
- j. Criterk for Nuclear Power Plants," the loss of offsite power is consider- [
j ed to be an anticipated operatiosial occurrence, since it is expected to occur one or more times during the life of a nuclear plant. Following a loss of of fsite power, stored and decay heat f rom the reactor would nor-f melly be removed by natural circulation using the steam gr.nerators as the 1 i
i
' heat sink. Natural circulation cooling of the primary system requires the t
use of the pressurizer to maintain a suitable overpressure on the reactor 1
coolant system. Consistent with satisfying the baste requirements in
{
, General Design Criteria 10,14,15,17, and 20 a selected number of pres-surizer heaters should be supplied from the emergency power buses. l l
P Evaluations of this item for operating plant licensees and applicants !
indicate that Westinghouse interf ace criteria in this area (i.e., minimum f t
number of pressurizer heaters necessary to support natural circulation and }
t t
the time available for connection of the emergency power source following [
lk a loss of offsite power) are conservative. f f
i
$PWR Response f
. t i
I P For the ppWR design one group of pressurizer backup heaters (manually i 1.oaded within I hour). is suf ficient to maintain natural circulation fol-lowing a loss of of fsite power. To ensure availability of at least one group of backup heaters upon loss of of f site power, emergency power will
! O be provided from separate diesel generators to two groups of backup heaters.
i j c l
)O !
. (
I f i
O f
@pWR-RC 3.1 31 NOVtM8tN, 1993 i 00We:1
- 19. Containment Isol tion System 100FR 50.34(f)(2)(xiv)
" Provide containment isolation systems that: ( A) ensure all nonessential systems are isolated automatically by the containment isolation system, (B) for each non-essential penetration (except instrument lines) have two isolation barriers in series. (C) do not result in reopening of the con-tainment isolation valves on resetting of the 1:olation signal, (0) util-tze a containment set point pressure for initiating containment isolation as low as is compatible with normal operation, and (E) include automatic closing on a high radiation signal for all systems that provide a path to the environs.'
0 i s c U11),,on, General Design Criterion 54, " Piping Systems Penetrating Containment," of Appendix A to 10CFR Part 50 requires that piping systems penetrating primary reactor containment be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and perfor-mance capability which reflect the importance to safety of isolating the piping systems. Standard Review Plan 6.2.4, " Containment Isolation System," requires that there be diversity in the parameters sensed for the initiation of containment isolation.
Some early plants (including TMt 2) provided automatic containment isola-tion demand on high containment pressure only. For small rates of loss of coolant, there would be little pressure increase in the containment, and automatic isolation could be delayed or possibly not occur. The loss of coolant at lHI-2, which produced a small pressure rise in the containment, was accompanied by substantial core damage and a large release of radio-nuclides into the containment building. Containment isolation was not achieved for some hours after the start of the event.
l lhe purpose of this requiren'ent is to ensure that of fective containment isolation is accamplished and maintained.
WAPWR-RC 3.1-32 NOVEMBER, 1983 0060e:I
.. . _ ._ _ _ - ~ _ __ ._ __- _ .-. __ _
3.c s
t-
-g
% A
.i l
y ' ~
MAPWR Response
< c h ' s ,
x In regard to requirement (A) above, careful consideration will be given to the definition and identification of essential and nonessential systems.
4
~
Westinchouse will document the basis for selection of essential systems during the licensing process for the MAPWR design.
l In reg'ard to , requirement (B) above, for post-accident situations each
$ non-essential penetration (except instrument lines) will have two isola-i tion barriers in series in accordance with the requirements of General Design Criteria 54, 55, 56, and 57, as clarified by Standard Review Plan 6.2.4. Isolat' ion will be performed automatically (i.e, no credit will be given ~ for operator action). Manual valves will be sealed closed, as defined by Stindard Review Plan 6.2.4, to qualify as an isolation bar-
)
rier. Each automatic isolation valve in a non-essential penetration will
! receive diverse isolation signals.
In regard to requirement (C) above, the design of control systems for
~
y k automatic containment isolation . valves will be such that resetting the isolation signal will-not result in the automatic reopening of containment l isolation valves. Reopening of containment isolation valves will require deliberate operator action. Administrative provisions to close all isola-f '
tion valves manually before resetting the isolation signals will not be considered an acceptable method of meeting this requirement.
Ganged reopening of containment isolation valves is not acceptable.
Reopening of isolation valves will be performed'on a valve-by-valve basis, j or on a line-by-line basis, provided that electrical independence and
! other single-failure criteria continue to be satisfied.
l In regard to requirement (0) above, the containment set point pressure that initiates containment isolation for non-essential penetrations will be the . minimum compatible with normal operating conditions. 'The' pressure set point selected will be far enough above the maximum expected pressure O .
3.1 -33 NOVEM8ER, 1983-MAPWR-RC '
0060e:1- e i
L - k
i 6 . . -
K inside containment during normal operation so that inadvertent containment isolation does not occur during normal operation from instrument drif t or fluctuations due to the accuracy of the pressure sensor.
In regard to requirement (E) above, all systems that provide a path f rom the containment to the environs (e.g., containment purge and vent systems) will close on a safety-grade high radiation signal.
- 20. Containment Purging / Venting 10CFR 50.34(f)(2)(xv)
" Provide a capability for containment purging / venting designed to minimize the purging time consistent with Al. ARA principles for occupational expos-j ure. Provide and demonstrate high assurance that the purge system will reliably isolate under accident conditions."
Discussion While the containment purge and vent systems provide plant operational flexibility, their designs must consider the importance of minimizing the release of containment atmosphere to the environs following a postulated loss-of-coolant accident. Therefore, the NRC position is that plant designs must not rely on their use on a routine basis.
The need for purging has not always been anticipated in the design of plants, and therefore, design criteria for the containment purge system have not been fully developed. The purging experience at operating plants varies considerably f rom plant to plant. Some plants do not purge during reactor operation, some purge intermittently for short periods, and some purge continuously. There is similar disparity in the need for, and use of, containment vent systems at operating plants.
O l
3.1-34 NOVEMBER, 1983 WAPWR-RC 0060e:1 1
l l
l l.
Containment purge systems have been used in a variety of ways; for example, to alleviate certain operational problems, such as excess air leakage into the containment f rom pneumatic controllers, for reducing the airborne activity within the containment to facilitate personnel access
[ during reactor power operation, and for controlling the containment pres-
!- sure, temperature, and relative humidity. Containment vent systems are typically used to relieve the initial containment pressure buildup caused by the heat load impos[ed on the containment atmosphere during reactor
O power ascension, or- to periodically relieve the pressure buildup due to the operation of pneumatic controllers.
t The sizing of the purge lines in most plants have been based on the need to control the containment atmosphere during refueling operations. This need has resulted in very large lines penetrating the containment (some on the order of 42 inches in diameter). Since these lines are normally the i only ones provided that will permit some degree of control over the con-tainment . atmosphere to facilitate personnel access, some plants have used them for containment purging during normal plant operation. The NRC 'is i O~ concerned with this situation during a postulated loss-of-coolant _ acci-dent, since the lines provide an open path f rom the containment to the
- environs and calculated accident doses could be significant.
l Therefore, the NRC is currently requiring compliance with Branch-Technical
[ Position CSB 6-4, "Containament Purging During Normal Plant Operations."
The following are included as requirements in Branch Technical Position CSB 6-4:
o The use of large containment purge lines is restricted to cold shutdown and refueling operations (the lines must be sealed closed in all other operational modes).
o Additional- smaller purge lines (about 8 inches in diameter or smaller) can be provided for continuous purging (lines larger than 8-inches in diameter must be justified to the NRC).
O 3.1-35 NOVEMBER, 1983-WAPWR-RC 0060e:1
WAPWR Response The containment purging / venting capability for the WAPWR design will be such that:
O o Reliable containment isolation will be achieved under accident conditions in accordance with item 19 above, o Purge time will be minimized consistent with ALARA principles for occupational exposure.
During the licensing process for the WAFAR design, We;tinghouse will demonstrate the level of conformance of the WAPWR design to the NRC accep-tance criteria documented in Branch Technical Position CSB 6-4.
- 21. Specific Accident Monitoring Instrument.1 tion 10CFR 50.34(f)(2)(xvii)
" Provide instrumentation to measure, record and readout in the control l room: (A) containment pressure, (B) containment water level, (C) contain'-
ment hydrogen concentration, (0) containment radiation intensity (high i level), and (E) noble gas effluents at all potential accident release points. Provide for continuous sampling of radioaftive iodines and parti-culates in gaseous effluents from all potential accident release points, and for onsite capability to analyze and measure these samples."
Discussion General Design Criterion 13 " Instrumentation and Control," of Appendix A to 10CFR Part 50 requires instrumentation to monitor variables and systems l
over their anticipated ranges for normal operation, for anticipated opera-I tional occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can effect the containment and its associated systems.
WAPWR -RC 3.1-36 NOVEMBER, 1983 0060e:1
)
In the past, General Design Criterion 13 had been implemented based on 4
design basis accidents analyzed in Chapter 15.0 of safety analysis f
reports. Based on conditions experienced at TMI-2, situations can arise which produce containment conditions beyond those postulated for the Chapter 15.0 events.
The purpose of this requirement is to ensure that capability is provided in the control room to ascertain containment conditions during the course of an accident.
i
@ PWR Response The above required instrumentation is included in Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident "
and, as such, it will be included in the overall @ PWR post-accident moni-toring design that addresses this regulatory guide. Refer to the discus-sion of item 23 below. ,
Control room instrumentation for the @ PWR design as a result of this
- regulation will be integrated with other control room instrumentation i requirements as discussed in item 8 above.
i
- 22. Inadequate Core Cooling Instrumentation 10CFR 50.34(f)(2)(xviii
" Provide instruments that' provide in the control room an unambiguous indication of inadequate core cooling, such as primary coolant saturatio'n A meters in PWR's, and a suitable combination of signals f rom indicators of b coolant level in the reactor vessel and in-core thermocouples in PWR's and BWR's."
O .
3.1-37 NOVEMBER, 1983
, @PWR-RC l 0060e:1
1 Discussion General Design Criterion 13 " Instrumentation and Control," of Appendix A to 10CFR Part 50 requires instrumentation to monitor variables for acci-dent conditions as appropriate to assure adequate safety. In the past, General Design Criterion 13 was not interpreted to require instrumentation to directly monitor water level in the reactor vessel or the adequacy of core cooling. The conventional instrumentation available that could indicate inadequate core cooling includes core exit thermocouples, cold leg and hot leg resistance temperature detectors, in-core neutron detec-tors, and ex-core neutron detectors. Generally, such instrumentation is included in the reactor design to perform functions other than monitoring of core cooling or indication of vessel water level.
During the TMI-2 accident, a condition of low water level in the reactor vessel and inadequate core cooling existed and was not recognized for a long period of time. This problem was the result of a combination of factors including an insufficient range of existing instrumentation, in-adequate emergency procedures, inadequate operator training, unfavorable instrument location (scattered information), and perhaps insufficient instrumentation.
The purpose of this requirement is to provide the reactor operator with instrumentation that, together with improved operating procedures and training, will enable him to readily recognize and implement actions to correct or avoid conditions of inadequate core cooling.
WAPWR Response The above required instrumentation (reactor vessel level instrumentation system and thermocouple / core cooling monitoring system) is included in Regulatory Guide 1.97, Revision 3. " Instrumentation f or Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and
! Following an Accident," and, as such, it will be included in the overall WAPWR post-accident monitoring design that addresses this regulatory l guide. Refer to the discussion of item 23 below.
WAPWR-RC 3.1 38 NOVEMBER, 1983 0060e:1
l O Control room instrumentation for the WAPWR design as a result of this regulation will be integrated with other control room instrumentation requirements as discussed in item 8 above.
!O
! 23. Post-Accident Monitoring Instrumentation i
10CFR 50.34(f)(2)(xix) l-. " Provide instrumentation adequate for monitoring plant conditions follow-ing an accident that includes core damage."
a Discussion i
I f
General Design Criterion 13 " Instrumentation and Control," of Appendix A
! to 10CFR Part 50 requires instrumentation to monitor variables and systems over their anticipated ranges for accident conditions as appropriate to i ensure adequate safety. General Design Criterion 19 " Control Room,"
i / requires that a control room be provided f rom which actions can be taken to maintain the nuclear power unit in a saf e condition under accident 1 conditions. In addition, General Design Criterion 64, " Monitoring Radio-activity Releases," requires means for monitoring the reactor containment i atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, ef fluent discharge paths, and the plant environs f
i for radioactivity that may be released from postulated accidents.
i j The overall subject of adequate post-accident monitoring has been a con-
! corn of the NRC and the industry for many years. As a result of this initial concern which was amplified in light of the TMI-2 accident, the l NRC has issued guidance in the form of Regulatory Guide 1.97, Revision 3 I " Instrumentation for 1.ight-Water-Cooled Nuclear Power Plants to Assess l Plant and Environs Conditions During and Following 'an Accident," which describes a method acceptable to the NRC staf f for complying with the l
l Conniission's requirements to provide instrumentation to monitor plant variables and systems during and following an accident.
3.1-39 NOVEM8ER, 1983 l - IdAPWR-RC 0060e:1
-,,.-m-- r-s, , . - - - -w.m- -- - -w,,. .,n,,..-,e.,-e. -,.vm ,
I Westinghouse has developed an interpretation of the requirements necessary to meet the intent of Regulatory Guide 1.97 for a conventional operating plant applicant. The specific variables specified in this design basis interpretation of Regulatory Guide 1.97 are not entirely applicable to the WAPWR design as a result of differences from conventional designs.
WAPWR Response Using the above mentioned Westinghouse design basis interpretation of Regulatory Guide 1.97 as a starting point, a similar document for the
- WAPWR will be developed and the results implemented in the design.
Westinghouse will completely document and justify any deviations f rom the NRC Regulatory Guide 1.97, Revision 3, positions during the licensing process for the WAPWR design.
Control room instrumentation for the WAPWR design as a result of this regulation will be integrated with other control room instrumentation requirements as discussed in item 8 above.
- 24. Power Supplies for Pressurizer Relief Valves, Block Valves, and Level
! Indicators i
" Provide power supplies for pressurizer relief valves, block valves, and level indicators such that: (A) level indicators are powered f rom vital buses. (B) motive and control power connections to the emergency power sources are through devices qualified in accordance with requirements applicable to systems important to safety, and (C) electric power is proviced from emergency power sources."
O WAPWR-RC 3.1-40 NOVEMBER, 1983 0060e:1
4 i-
- Discussion 3
l Pursuant to NRC regulations in 10CFR Part 50, Appendix A, " General Design Criteria for Nuclear Power Plants," the . loss of offsite power is con-sidered to be an anticipated operational occurrence, since it is expected to occur one or more times during the life of a nuclear plart. Following
! a loss of offsite power, stored and decay heat from the reactor would nor-mally be removed by natural circulation using the steam generators as the f
- heat sink. Alternatively, in the event that natural circulation in the reactor coolant system is interrupted, the feed and bleed mode of reactor coolant system operation can be used to remove decay heat from the reac-i tor. This method of decay heat removal requires the use of the emergency core cooling system and the pressurizer power-operated relief valves.
i Consistent with satisfying the basic requirements in General Design Criteria 10, 14, 15, 17 and 20, the pressurizer power-operated relief valves and associated block valves and level indicators must be supplied f rom emergency sower buses.
t i O More specific NRC guidance for implementation of this regulation is con-r
.' tained in Item 11.G.1 of NUREG-0137, " Clarification of IM1 Action Plan f Requirements."
l l WAPWR Response 4
I The LdAPWR design will include provisions for appropriate emergency power for pressurizer equipment. In the determination of the power supplies for O the pressurizer power-operated relief valves and associated block valves, consideration will -be given to cold shutdown and reactor coolant system overpressurization requirements in addition to the post-TM1 requirements of this regulation.
O l
lO
.3.1-41 NOVEM8ER, 1983 WAPWR-RC 0060e:1
I
- 25. Emergency Response Facilities O
" Provide an onsite technical support center, an onsite operational support center, and, for construction permit applications only, a nearsite emer-gency operations facility."
Discussion In addition to the above regulation, Article IV.E.8 of Appendix E, "Emer-gency Planning and Preparedness for Protection and Utilization Facili-ties," to 10CFR Part 50 requires that adequate provisions shall be made and described for emergency facilities and equipment, including a licensee onsite technical support center and a licensee near-site emergency opera-tions facility from which ef fective direction can be given and ef fective control can be exercised during an emergency. (Note that " effective control" must be interpreted to mean administrative control versus actual control of the plant). ,
l As one would expect, these regulations &re quite general in that they simply require emergency response facilities to be established. The NRC has, however, issued detailed guidance (e.g. functions, locations, size, structures, habitability, communications, instrumentation, etc.) for the j ,
design of emergency response facilities in the form of NUREG-0737, Supple-i ment 1, " Requirements for Emergency Response Capability," and NUREG-0696, l
" Functional Criteria for Emergency Response Facilities."
WAPWR Response The onsite technical support center will be included in the WAPWR design.
l The on-site operational supprt center and the near-site emergency opera-tions facility is the responsibility of each utility utilizing the WAPWR design.
O WAPWR RC 3.1-42 NOVEMBER, 1983 0060e:1 i
l
I-l During the licensing process for the WAPWR design, Westinghouse will ,
demonstrate the level of conf ormance of the WAPWR design to the NRC guld-ance documented in NUREG-0737 (Supplement 1) and NUREG-0696 and/or other i applicable documents.
l
- 26. Leakage Control Cutside Containment i
1COFR 50.34(f)(2)(xxvi)
" Provide for leakage control and detection in the design of systems out-side containment that contain (or might contain) T ID-14844 source term i
radioactive materials following an accident. Applicants shall submit a I leakage control program, including an initial test program, a schedule for retesting these systems, and the actions to be taken for minimizing leak-
, age from such systems. The goal is to, minimize potential exposures to 4
workers and public, and to provide reasonable assurance that excessive i leakage will not prevent the use of systems needed in an emergency."
Discussion i'
10CFR Part 20 and Part 100 specify radiation limits and guidelines that i must be met by licensed facilities to assure protection of public health and safety. In a power reactor, many systems that will or may handle liquids or gases containing large radioactive inventories af ter a serious ,
transient or accident are located outside containment. Several of the i
i i
engineered safety features and auxiliary systems located outside reactor containment will or may have to function during a serious transient or accident with large radioactive inventories in the fluids they process.
I The leakage from these systems, when operated, should be minimized or eliminated to prevent the release of significant amounts of radioactive A
l materials to the environment. Historically, these systems are checked out f
during preoperational testing and startup testing but are not usually included in any periodic leak testing program. It is beneficial if the i
l l
plant operating staf f knows the leakage rates of these systems and main-tains them at rates that are as low as practical, i
3.1-43 NOVENBER, 1983 WAPWR-RC 6060e:1 4
. , , - - - , .- - - - . _ - . , --- - . . , - - - - , - .e.,----.., . , , . , , , - - - - . , - - - . _ _ . , , , , _ - - - , , , . - - ~ , _ . . .
The purpose of this regulation is to make every ef fort to eliminate or l reduce the leakage f rom these systems, perform periodic tests to assure that the leakage from these systems is maintained as low as practical, and provide the plant staff with current knowledge of the system leakage rates.
WAPWR Response The WAPWR design includes the following features which minimize the potential for leaks and/or improve leaksge control and detection.
o The amount of equipment located outside containment has been minimized.
o The Primary Side Safeguards Equipment is segregated into four separate independent safeguard component areas.
o Capability to test the full recirculation path is provided.
The actual leakage testing procedures will be established by each utility using the WAPWR design.
- 27. Inplant Monitoring 10CFR 50.34(f)(2)(xxvii)
" Provide for monitoring of inplant radiation and airborne radioactivity as appropriate for a broad range of routine and accident conditions."
Discussion 10Cl-R Part 20 " Standards for Protection Against Radiation," provides criteria for control of exposures of individuals to radiation in restricted areas, including airborne lodir since iodine concentrates in the thyroid gland, airborne concentrati. +5 must be known in order to e
WAPWR-RC 3.1 44 NOVEMBER, 1983 0060e:1
l evaluate the potential dose to the thyroid. Historically, the concentra- l tion of iodine in atmosphere air has been determined by measuring the ;
j activity of iodine adsorbed in a carbon filter through which air has been pumped. The charcoal filter is removed f rom the air pump and allowed to f' j
ventilate to permit the noble gases to dif fuse to the atmosphere. The j
i filter is then counted for radioactivity content and the remaining activi-i
! ty is ascribed to iodine. This procedure is conservative; however, it is possible for suf ficient noble gas to be adsorbed in the charcoal 50 that the resulting todine determination may be unduly conservative (high). If i
the airborne iodine concentration is overestimated, plant personnel may be l required to perform operational functions while using respiratory equip- f l
ment, which sharply limits communication capability and may diminish [
l i personnel performance during an accident.
I l i !
j The purpose of this requirement is to improve the accuracy of measurement j
of airborne iodine concentrations as well as to ensure adequate inplant [
j monitoring of vital areas.
I(-
- LdA M Response -
l t
The FM design will include suf ficient todine samplers to sample all i I vital areas. [
t i
}
l 28. Control Room Habitability ,
I
- ]AffR 50.34(f)(2)(xxvii11 f 1
i
- Evaluate potential pathways for radioactivity and radiation that may lead l
to control room habitability problems under accident conditions resulting i
in a 110-14844 source term release, and make necessary design prowlsions !
l to preclude such problems."
i
- t
! f iO 3.1 45 movinsta, iges 1 l p m -nc l
0060 :1 !
. _ __. _. __....~ -_ -__ _. -._..__,.,_ ____.__._.~..._._._._ _ _.._.____-
l l
Discussion Control room habitability deals with assuring that control room operators will be adequately protected against the ef fects of an accidental release of toxic and radioactive gases and that the plant can be safely operated or shutdown under design basis accident conditions (in accordance with General Design Criterion 19 " Control Room," of Appendix A to 10CFR Part 50).
For plants designed over the last 5 to 8 years, this TMI item (in general) has not presentet? a significant problem (beyond sof tware documentation),
since the current NRC guidance and acceptance criteria for ensuring control room habitability was available during the design and licensing of these plants.
l WAPWR Response The WAPWR design will include appropriate provisions for control room habitability in accordance with the NRC guidance provided in Standard Review Plan ,6.4, " Control Room Habitability System." Westinghouse will completely document and justify any deviations from the NRC Standard Review Plan 6.4 acceptance criteria during the licensing process for the MAPWR design.
- 29. Industry Experiences 10CFR 50.34(f)(3)(1) l l
" Provide administrative procedures for evaluating operating, design and construction expertence and f or ensuring that appilcable important ind'as-l try experiences will be provided in a timely manner to those designing and constructing the plant."
l l
l l WAPWR RC 3.1 46 NOVEMBER, 1983 0060e:I
Discussion i
This requirement deals with administrative procedures which by themselves do not impact any design.
Westinghouse has always recognized the need to stay appraised of operating events to meet the need for feedback of operating experiences to design, construction, and operation. Currently, this is being accompitshed j
through informal methods of screening various media sources (e.g., INPO.
Westinghouse site managers daily reports, NRC Inspection and Enforcement l
Bulletins, circulars, and Information Notices) for event information.
Those events or issues identified as having potential significance are l
routed internally to appropriate cognizant personnel for their evaluation l
and follow-up action as necessary.
WAPWR Response Westinghouse is considering more formal programs for systematically fol-
]
i lowing and incorporating operating and construction experiences in the WAPWR design and ' reliability evaluations. This subject will be fully l
addressed during the licensing process for the MAPWR design.
- 30. Quality Assurance List t
" Ensure that the quality assurance (QA) list required by Criterion II.
Appendix 8, 10CFR Part 50 includes all structures, systems and components important to safety."
O Discussion Appendix 8. " Quality Assurance Criterion for Nuclear Power Plants and Fuel Reprocessing Plants," to 10CFR Part 50 establishes quality assurance O
3.1-47 NOVEMBER, 1983 WAPWR-RC 0060e:1 L t
1 requirements for all activities af fecting the design, construction, and operation of those safety-related structures, systems, and components that prevent or mitigate the conseglences of postulated accidents or could cause undue risk to the health and safety of the public. Criterion II of Appendix B futher requires the identification of the structures, systems, and components to be covered by the quality assurance program.
Historically, the requirements of Appendix B have been mostly applied only to safety-related structures, systems, and components (e.g., for a conven-tional design this encompasses Safety Class 1, 2, and 3 structures, sys-tems, and components). This approach has been (in general) accepted by the NRC in the past even though Appendix A. " General Design Criteria for Nuclear Power Plants," of 10CFR Part 50 requires the establishment of principle design criteria for structures, systems, and components impor-tant to safety; that is, structures, systems, and components that provide reasonable assurance that the facility can be operated without undue risk I
to the health and safety of the public. Structures, systems, and compon-ents important to safety can and do include equipment that has been his-torically classified as Non-Nuclear Safety. A Non-Nuclear Safety classi-fication has been translated into a quality assurance program less strin-gent than that required by Appendix B of 10CFR Part 50.
The purpose of this regulation is to ensure that appropriate quality assurance is applied to all structures, systems, and components important to safety versus only those that are safety-related.
WAPWR Response l
l The WAPWR design, quality assurance program, and associated listings of structures, systems, and components will comply with this requirement.
Westinghouse will apply applicable 10CFR Part 50, Appendix B, quality
! assurance criteria to structures, systems, and components important to l safety.
O
\
WAPWR-RC 3.1-48 NOVEMBER, 1983 j
0060e:1
0
'O 31. Quality Assurance Program t
- Establish a quality assurance (QA) program based on consideration of:
(A) ensaring independence of the organization performing checking func-tions from the organization responsible for performing the functions; (8) l
! performing quality assurance / quality control functions at construction sites to the maximum feasible extent; (C) including QA personnel in the documented review of and concurrence in quality related procedures associ-ated with design, construction and installat,lon; (D) establishing criteria for determining QA progransnatic requirements; (E) establishing qualifica-tion requirements for QA and QC personnel; (F) sizing the QA staf f commen-2 surate 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."
l 1
- Discussion
! (1
+
j Various TMI-2 accident investigations and inquiries identified problems relating to the quality assurance organization, authority, reporting, and l
inspection. This regulation is a result of NRC actions taken to improve j the quality assurance program for design, construction, and operations to i provide greater assurance that these activities are conducted in a manner i coemeensurate with their importance to safety. ,
Westinghouse has established anj is implementing a quality assurance pro-I gram, approved by the NRC, that complies with 10CFR Part 50, Appendix 8 "Qaulity Assurance Criterion for Nuclear Power Plants and Fuel Reproces-sing Plants," and the considerations listed in the above regulation. This i program currently addresses Westinghouse design and construction activi-ties and may be revised in the future to include onsite construction activities and operations.
!O WAPWR-RC 3.1-49 NOVEM8f.R . ' 1983 0060e:1
_ , , _ , - , . . - , _ . _ . , , - , - - , . , - , , , , . . . . . _,m._--_.,~.__,,___--____e-
(
l WAPWR Response Westinghouse will document the quality assurance program applicable to the WAPWR program during the licensing process for the WAPWR design.
O
- 32. Dedicated Containment Penetrations 10CFR 90.34(f)(3)(iv)
' Provide one or more dedicated containment penetrations, equivalent l'n O
size to a single 3-foot diameter opening, in order not to preclude future installation of systems to prevent containment failure, such as a filtered vented containment system."
Discussion As discussed in more detail in Section 3.2, there are rulemaking ef forts currently underway to establish policy, goals, and requirements related to accidents involving core damage greater than the present design basis.
One of the design requirements being considered in these efforts is the need for a new structure for controlled filtered venti; ' of the reactor containment structure.
The purpose of this requirement is to ensure the capability of installing such a system should it be determined necessary.
Westinghouse studies' (as well as other industry studies) of conventional plant designs have indicated that filtered vented containment systems may not be cost-effective for large dry containments similar to the WAPWR containment design.
WAPWR Response As part of the WAPWR design risk assessment (discussed in more detail in Section 3.2), Westinghouse will evaluate the potential benefits of WAPWR-RC 3.1-50 NOVEMBER, 1983 0060e:I l
i
]
filtered vented containment systems. 9ased on these evaluations'and ;
! associated cost-benefit considerations Westinghouse will either:
5 o Include one or more dedicated penetrations in the W Pk.A design for potential future installation of a filtered vented containment system in accordance with this regulation, or o Request exemption from this regulation for the WAPWR design.
i
- 33. Containment Design I
i l 10CFR 50.34ff)(3)(v) L b " Provide preliminary design information at a level of detall consistent with that normally required at the construction permit stage of review t l sufficient to demonstrate that:
( A)(1) Containment integrity will be maintained (i.e., for steel contaln-( ments by meeting the requirements of the ASME Boiler and Pressure Vessel (
Code, Section !!!, Division 1 Subsubarticle NE-3220, Service Level C l
Limits, except that evaluation of instability is not required, considering l
! pressure and dead load alone. For concrete containneats by meeting the :
f requirements of the ASME Soiler and Pressure Vessel Code. Section !!!,
Division 2 Subsubarticle CC-3720 Factored Load Category, considering i
pressure and dead load alone) during an accident that releases hydrogen .
! generated from 100 percent fuel clad instal-water reaction accompanied by either hydrogen burning or the added pressure f rom post-accident inerting j assuming carbon dioxide is the inerting agent. As a minimum, the specific code requirements ' set forth above, appropriate for each type of contain-f ment, will be met for a combination of dead load and an internal pressure of 45 psig. Modest deviations f rom these criteria will be considered by f
the staf f, if good cause is shown by an applicant. Systems necessary to ensure containment integrity shall also be demonstrated to perform their I
l function under these conditions.
lO WAPWR-RC 3.1-$1 ; NOVEM8EN 1983
- 0060e
- 1 ;
l (2) Subarticle NE-3220, Division 1, and subarticle CC-3720, Division 2, of Section !!! of the July 1, 1980 ASME Boller and Pressure Vessel Code, which are referenced in paragraph (f)(3)(v)( A)(1) and (f)(3) (v)(B)(1) of this section, were approved for incorporation by reference by the Director of the Office of the Federal Register. A notice of any changes made to the material incorporated by reference will be published in the Federal Register. . . .
(B)(1) Containment structure loadings produced by an inadvertent full actuation of a post-accident inerting hydrogen control system (assuming carbon dioxide), but not including seismic or design basis accident load-ings 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, Subsubarticle NE -3220, Service Level A Limits, except that evaluation of instability is not required (for concrete containments the loadings specified above will 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-3120, Service Load Cate-gory), (2) The containment has the. capability to safely withstand pressure tests at 1.10 and 1.15 times (for steel and concrete containments, respec-tively) the pressure calculated to result from carbon dioxide inerting."
Discussion The accident at THI-2 resulted in a severely damaged or degraded reactor core with a concomitant release of radioactive material to the primary coolant system and a large amount of fuel cladding metal-water reaction in the core with hydrogen generation well in excess of the amounts required l
l to be considered for design purposes by historical Commission regula-tions. The accident revealed design and operational limitations that existed relative to mitigating the consequences of the accident and deter-mining the status of the facility during and following the accident.
l l
9 WAPWR-RC 3.1 52 NOVEMBf.R. 1983 0060e:I
r This regulation, in conjunction with the hydrogen control regulations of items 5 and 14, is intended to assure that containment structural integ-i rity is maintained during severe accident conditions,
- MAPWR Response The WAPWR containment design will be in accordance with this regulation.
- J
' 34. Hydrogen Recombiners I
"For plant designs with external hydrogen recombiners, provide redundant f
I dedicated containment penetrations 50 that, assuming a single failure, the recombiner systems can be connected to the containment atmosphere."
4 l Discussion
- V I
!' In accordance with 10CFR 50.44; " Standards for Combustible Gas Control l
System in t.ight Water Cooled Power Reactors," plant designs since about ,
late 1910 must include a combustible gas control system (such as recom- ,
i
)' biners) as the primary means for controlling combustible gases following a loss-of-coolant accident.
Certain plant designs satisfied this requirement with provisions for
! post-accident installation and operation of an external hydrogen recombin-er for combustible gas control. For example, TMI-2 had this external recombiner capability. The design of the external recombiner hookup at l
TMI-2 used the 36-inch containment penetrations for the normal containment purge system by tapping 4-inch lines of f the purge lines outside the con-tainment building between the building and the outer containment isolation I valves, 10 place the hydrogen recombiner into service required the opening of the inboard 36-inch containment isolation valve in both a i
3.1 53 NOVIN#lR 1983 WAPWR-RC 0040e:1
l n :
I containment purge system inlet and outlet line. With this design, once the hydrogen recombiner is put into operation, containment integrity is vulnerable to a single active failure. That is, a spurious or inadvertent opening of one of the 36-inch outboard containment isolation valves would have resulted in the venting of the containment to the environment. In addition, the design of the system to include use of the large (36-inch) containment purge penetrations resulted in the operation of the recombiner beyond the design capacity of the unit.
This requirement does not apply to Westinghouse designed plants that incorporate internal hydrogen recombiners.
Since the TMI-2 event, the NRC has revised 10CFR 50.44 to also require dedicated containment penetrations for external recombiners.
WAPWR Response The WAPWR design will include a manually actuated recombiner system which is redundant, qualified, and Installed inside containment. However, the total hydrogen control system will be in accordance with item 14 above.
- 35. Management Plan 10CfR 50.34(f1(3)(vi1).
" 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; (0) technical resources director by the applicant; (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; (0) proposed procedures for handling the transition to operation; I (E) the degree of top level management oversight and technical control to O
l 3.1454 NOVLMULR, 1983 WAPWH RC 0060e: 1
!O i
be exercised by the applicant during design and construction, including i
i the preparation and implementation of procedures necessary to guide the
! effort."
i
!O i
1 Discussion 2
I 4 One of the major findings as a result of the 1MI-2 accident was the need f to improve staf fing to oversee design and construction activities. This regulation is intended to address this finding.
t l
The NRC has issued draft NUREG 0731, " Guidelines for Utility Management l
) Structure and Technical Resources," which is expected to be used by utili-j ties as guidance in meeting this regulation.
l l ljAPWR Response I +
This regulation is directed at utility management organizational and administrative capabilities and is not applicable to Westinghouse in i( "
I relation to the tjAPWR design.
l O
O O
3.1 $$ NOVEMBER, 1983 WAPWR-RC 0060e:I
4
, l LO 3.2 SEVERE ACCIDENT RULEMAKING AND RELATED CONSIDERATIONS i Discussion The 1MI'-2 accident and the results of subsequent reviews and investigations prompted the Conunission to r,econsider certain aspects of its licensing policy. One of the conclusions from the post-TM1 investigations was that j, attention should be given to the probability and consequences of severe acci-dents (beyond the normal design basis accidents) and that a policy statement on the acceptance level of risk to the public health and safety was needed.
Since the completion of WASH-1400, " Reactor Safety Study - An Assessment of Accident Risks in U.S. Commercial Nuclear Fower Plants," development in prob-3 abilistic risk assessment techniques has led to th? recognition that it is i feasible to use quantitative guidelines in evaluat Ing reactor saf ety. These techniques are a vlable means of assessing plant risks and comparing them with the prorosed safety goal discussed below.
k The Commission is requiring (through the regulations of 10CFR 50.34(f)(1)(i)
- and their proposed rule on severe accidents) performance of a prob- abilistic j risk assessment and associated reliability engineering programs for standard plant designs to be referenced by new construction permit applica- tions.
l The Commission initiated ef forts to develop a safety goal policy in its,1981 Federal Register Notice entitled " Development of a Safety Goal - Preliminary Policy Consideration" and subsequently published a " Proposed Policy Statement l
4 on Safety Goals for Nuclear Power Plants" (47FR7023 dated February 17, 1982).
(' This proposed statement served as an initial step toward the ultimate goal of explicitly defining the acceptable level of risk to the publ*c health and l Safety from nuclear power reactors and included several qualitative safety I goals and quantitative probabilistic guidelines for severe accidents.
l O
3.2 1 NOVLMBLH, 1983
@PWR-RC 0060e:1
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Af ter several draf ts and revisions as a result of comments and recommendations received by the NRC, on March 14, 1983, the Commission published a " Policy Statement on Safety Goals for the Operation of Nuclear Power Plants" (48FR10774) for a two year trial use and evaluation period. The policy state-ment includei the following preliminary safety goals and preliminary numerical design objectives, o Safety Goals -
" Individual members of the public should be provided a level of pro-tection f rom the consequencies of nuclear power plant operation such that Individuals bear no significant additional risk to life and l health."
i A
" Societal risks to life and health from nuclear power plant operation should be comparable to or less than the risks of generating electric-ity by viable competing technologies and should not be a significant addition to other societal risks."
i o Quantitative Design Objectives ,
f j 1. Individual and Societal Mortality Rhks "The risk to an average individual in the vicinity cf a nuclear j
power plant of prompt f atalities that might result f rom reactor I accidents should not exceed one-tenth of one percent (0.1%) of the sum of prompt fatality risks resulting f rom other accidents to which members of the U.S. population are generally exposed."
l l The risk to the population in the use of a nuclear power plant of cancer fatalities that might result from nuclear power plant operation shcJ1d not exceed one-tenth of one percent (0.1%) of the sum of cancer fatality risks resulting from all other causes."
O WAPWR-RC 3. 2 -2 NOVEMBER, 1983 0060e:1
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r i
> 2. Benefit-Cost Guideline '
j "The benefit of an 1.1cremental reduction of societal mortality risks should be compared with the associated costs on the basis of 4
$1000 per person-rem averted."
l I
- 3. Plant Performance Design Objective LO
! "The likelihood of a nuclear reactor accident that results in a !
. large-scale core melt should normally be less than one in 10.000 per year of the reactor operation."
i i
In reprd to the individual and societal mortality risk guidelines the Policy Statement defines "the vicinity" to be considered in the individual risk of f prompt fatality as the area within one mile of the site boundary, and in the f design objective for cancer fatality the population within 50 miles of the t plant. site is to be considered. The Policy Statement also notes that: the j application of the benefit cost guideline should be focused principally on
! (" situations where one of the quantified safety goals is not met; and the design {
. objective for large scale core melt is subordinate to the principal design l j' objectives limiting individual and societal risks. . l D
The Commission's intention is that the design objectives and benefit-cost ,
I i guideline would be used by the NRC staf f in conjunction with probabilistic
}
l risk assessments and would not substitute f or reactor regulations in 10CFR
, {
Chapter I. Rather, individual licensing decisions would continue at present l f
l to be based principally on compliance Vith the Commission's regulations.
I r
During the next two years the safety goal policy will be evaluated as to their adequacy and usefulness in the regulatory process. In this process there is 1
1 i to be trial application to a number of generic issues to gain hands on exper- !
! ience, however the safety goal is not to be a f actor in their resolution. The NRC has indicated that the implementation plan will require the following for j
new construction permit applications ard standard plant applications: ;
3.2 3 NOVI.M5f.R, 1983
! WAPWR-RC 0060e:1 l
[ .
L- L , . . . _ _
O A plant / site specific probabilistic risk assessment o Achievement of the Design Objectives o further safety improvements in accordance with the Benefit-Cost Guideline in addition, the Conentsdon had previously published an advanced notice of proposeil r'alemaking concerning consideration of degraded or melted cores in safety regulations (456 R654 74 datc<'. October 2, 1980), in that notice, the Conen ssion indicated that a long-term rulemaking ef fort was being initiated that would establish policy, quals, and requirements relating to core-melt accidents greater than the prcsen* design basis. The current NRC direction in this area replares the loeg-term generic rulemaking ef fort with severe acci-dent rulemakings designed to certify specific standard plant design appitca-tions and with regulatory decisions based on generic evaluations and decisions regarding all classes of exilting plar.L. The " Proposed Commission Policy Statement or, Severe Accidents and pelated Views on Nuclear Reactor Regulation" (40FR16014, April 11. 1983) ccMains the following nine inter related compon-ents: ,
o Policy Statement on Safety Goals o use of Probabilistic Risk Assessment in Severe Accident Decision Making o lessons Learned from Three Mile Island o Standard Review Plan 9
o Standardization Policy further Research on Severe Accidents O
o i
l O
3.2 4 NOV[H8tH, 1983 WAPWR-RC 0060e:I
.f)
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o Treatment of Severe Accidents in Ongoing Licensing Proceedings o Present views on Other Safety issues and Efforts in Progress 73
'l q ,'
)
o implementation Guidelines for Severe Accident Policy These inter-related components are sumarized below and cross-referenced to
) discussions provided in various other sections of this document as appro-priate. In addition, their impact on the MAPWR design is evaluated.
- 1. Policy Statement on Safety Goals Discussign lhe NRC ' Policy Statement on Safety Goals for the Operation of Nuclear Power Plants * (49f R10il2), as sumartzed earlier in Section 3.2, has the ,
ultimate goal of explicity defining the acceptable level of risk to the public health and safety from nuclear power reactors; and include $'several qualitative safety goals and quantitative probabilistic guidelines for severe accidents.
The impact of the Comission's safety goal policy on the WAPWR design is encompassed by the 'WAPWR Response
- given under item 2, below
- 2. Use of Probabilistic Risk Asseesment in Severe Accident Cecision Making (n)
'd Discussion l
! The NRC has concluded that the use of probabilistic risk assessment (PRA) techniques improves the ' understanding of the severe accident sequences to f')
f
\ which plants are most vulnerable and therefore of the dominant constit-
' uents of the risk posed by specific plants.' lhe performance of 1 plant / site specific PRA is a regulatory requirement (see Section 3.1, iten I; 10CfR50.34(f)(1)(1)).
(3 Lj 3.2 $ NOVLNHLR, 19f3 WAPWR-RC 0060e I
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l The NRC has further stated that "the utility of PRA can be improved if it is integrated with the design process." Therefore, the severe accident policy statement, once promulgated, will likely require "the performance of a PRA that is as complete as practical for any standardized design to be referenced in future construction permit applications."
WAPWR Response The. WAPWR development program is dedicated 'to using probabilistic risk assessment both in selecting component and system designs from among alternatives, and in system reliability evaluations. Westinghouse will completely document an integrated design and siting PRA for the WAPWR.
For the PDA, the PRA will be "as complete as practical." The WAPWR risk assessment will:
l l
l o include internal and external hazards as well as other accident I- indicators.
limitations and requirements, thereby O
o Reflect any operational establishing appropriate technical specifications as an integral part of the design.
o Demonstrate that the WAPWR design has a suf ficiently low risk that further consideration of the benefit-cost guideline (ALARA policy) is not necessary and, in that regard, establish that the 'WAPWR design is inunune f rom consideration of new regulatory requirements.
in addition, Westinghouse agrees with the NRC staf f that the Commission O
safety goal should not consider risks from sabotage. However, as l
discussed in Section 5.1, item 29, Westinghouse will perfonn a sabotage assessment for the WAPWR design using risk models which is intended to be I
provided to utilities utilizing the design for consideration in their physical protection plans.
O WAPWR-RC 3.2 -6 NOVEMBER, 1983 0060e:1
O 3. Lessons Learned From Three Mile Island (TMI)
Discussion
,O -
The lessons learned f rom TMI, as summarized in NUREG-0737, " Clarification of TMI Action Plan Requirerents," have been applied by the NRC to operating plants, plants in operating license review, and plants now undergoing construction permit review. These requirements, which apply to ,
future construction permit applications as well, have been codified in 10CFR50.34(f). A complete discussion of 10CFR50.34(f), including its impact on the WAPWR design, is given in Section 3.1.
- 4. Standard Review Plan l Discussion
. On March 19, 1982, the NRC incorporated a paragraph (g) to 10CFR50.34 that
- requires future applicants for operating licenses, construction permits, s' manufacturing licenses, and for preliminary or final design approvals for standard plants to identify and evaluate dif ferences from the acceptance criteria of the applicable revision of the Standard Review Plan (SRP) as part of the technical information to be submitted as 'part of an i.
application.
The SRP rule and its impact on the LdAPWR design is fully discussed - in Section 6.3.
- 5. Standardization Policy Discussion The NRC has reiterated its support for . standardization of nuclear power plants. This proposed rule requires applicants for PDAs and FDAs to l
address the guidance contained therein. Once a standard design has been f
taken through rulemaking, the approval would be binding .on both the NRC '
and the applicants "for a period of ten years unless significant new 3.2 7 NOVEMBER, 1983 LdAPWR-RC
'0060e:I
safety information becomes available." Applicants who intend to take i their standard design through a rulemaking procedure, per 10CFR Part 50 !
Section 7, will be given review priority by the NRC.
WAPWR Response This " Regulatory Conformance" document, in conjunction with the supporting RESAR SP/90 PDA submittals, demonstrates the level of conformance with the requirements of the proposed Commission policy statement on severe accidents. Westinghouse fully intends to pursue the completion of a rulemaking for the WAPWR design upon receipt of an FDA, consistent with the NRC objectives for licensing standard plant designs as embodied by this proposed rule.
- 6. Further Research on Severe Accidents Discussion The NRC is conducting research on severe accidents. The NRC does not expect its fundamental views on severe accident considerations to change substantially due to the on -going NRC research or industry sponsored research. The intent is to obtain suf ficient information in about 2 years to complete decision making. The NRC research program includes the following.
o Probabilistic risk assessment methods, including those treating external events; i
o Common-cause accident contributors:
O System interactions, including analysis of systems transients involving core damage; l
WAPWR-RC 3.2 -8 NOVEMBER, 1983 0060e:1
l mO o Accident management, including guidelines for recovery from a .
core-damaging event;
. o Phenomenological research on fuel and fission product behavior of i damaged cores and containment response to severe loadings; o Human factors; o Applications research on behavior of existing systems and ,
components in the severe accident environment; o Fission product release and transport; and I o Safety-cost tradeof f analysis of changes in hardware.
o An improved methodology for probabilistic risk assessment plus a significant extension of the data base for severe accident assessment; ,
( .
o- Data for a better estimate of the radiological source term used to
! assess accident consequences; WAPWR Response Westinghouse is taking an active role in industry research programs related to severe accidents including participation in the Industry Degraded Core Rulemaking (10COR) program. The results of . such programs will be continually, f actored into the design process for the WAPWR. In addition, Westinghouse will follow and give careful consideration to the results of the NRC study and research programs summarized above, and factor such results into the WAPNR design process as appropriate. As discussed throughout this document, the WAPWR design and licensing program is dedicated to conscientiously addressing the above listed severe accident considerations for which the NRC study and research program will GN investigate. For example. (1) PRA techniques are - used throughout the
- 3. 2 -9 NOVEMBER, 1983 MAPWR-RC' 0060e:1
design process, including the selection of design features from among alternatives; (2) a systems interaction study will be performed and documented during the licensing process for the WAPWR; (3) emergency response guidelines will be developed and documented f or the WAPWR; (4) human factors principles will be integrated into the overall control room design; and (5) Westinghouse intends to utilize a more realistic radiological source term in the assessment of accident consequences for the WAPWR.
- 7. Treatment of Severe Accidents in Ongoing Licensing Proceedings
( Discussion With respect to the impact of this severe accident policy statement on operating plants and plants under construction, the NRC has concluded:
(1) individual licensing proceedings, including hearings, are not appropriate forums for examination of the Commission's regulatory requirements on accidents more severe than the design basis, (2) the requirements of 10CFR50.34(f) are suf ficient for Class 9 accident review and hearings, and (3) review of non-standardized plants can proceed and be found acceptable for severe accident concerns if they meet 1) the TMI requirements, 2) Standard Review Plans and 3) achieve resolution of the Unresolved Safety Issues.
The Industry Degraded Core Rulemaking (IDCOR) program is directed at information that can be gained from operating plants, including non-nuclear plants, relative to systems, components and functions that relate to the potential for a degraded core accident, and the NRC is interested in assuring that the 10COR program and the NRC program are coordinated and complimentary.
WAPWR Response This section of the severe accident policy statement is primarily concerned with the impact it has on currently operating plants and plants WAPWR -RC 3.2-10 NOVEMBER, 1983 0060e:1
l l
under construction. As stated in the TAPWR Response" to item 6 above, Westinghouse will closely follow. the- NRC severe accident study and research program, and f actor the results into the WAPWR design process as O appropriate.
l Present Views on Other Safety issues and Efforts In Progress l
8.
Accident Prevention and Consequence A. Striking a Balance Between Mitigation Discussion I As a result of the TMI accident, the NRC developed an objective (for themselves as well as the nuclear industry) to give further consideration to severe accidents beyond the design basis, and to explore means to decrease the probability as well as mitigate the consequences of such accidents. .For example, there has been increased
- ( recognition that one of the most important systems in providing core-melt prevention is a reliable decay heat removal system.
WAPWR Response A fundamental design objective of the WAPWR is to significar.tly decrease the probability of a core-melt compared to current plant designs. -There are numerous design features and systems in'the LiAPWR, as discussed throughout this document, which substantially reduce the probability of and/or mitigate the consequences'of a severe accident.
- 8. Containment Strength Discussion The NRC has identified the need - to gain a better understanding of containment building failure characteristics and design features or 3.2-11 NOVEMBER, 1983 WAPWR-RC j.
0060e:1
l emergency actions that decrease the likelihood of containment building failure. Following the outcome of the NRC severe accident research program, the NRC will decide whether to establish performance criteria for containment systems.
O In addition, the NRC is studying the need for the following additional containment features, each of which are discussed in the following '
paragraphs:
o Filtered venting of containment; O
o Core-retention devices; and o Hydrogen control features.
WAPWR Response t
10CFR50.34(f)(3)(v) encompasses the severe accident policy statement concern of the adequacy of the containment design for future plants.
See Section 3.1, item 33 for a discussion of this issue and its impact on the WAPWR design.
C. Filtered-Vented Containment Systems Discussion The NRC has stated that for future construction permit applicants,
" filtered--vented containment systems, or a variation of such systems, should be provided if these yield a cost-effective reduction in risk."
( WAPWR Response The subject of the need for f iltered -vented containment systems is encompassed by 10CFR50.34(f)(iv). See Section 3.1, item 32.
O l WAPWR-RC 3.2-12 NOVEMBER, 1983 1 0060e:1 l
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- 0. Core-Retention Devices Discussion lhe NRC has stated that " studies (such as NUREG-0850) of large, dry containment buildings indicate that classical core-retention devices are probably not cost-ef fective in reducing the release of radioactive materials to the atmosphere. However, unique basemat designs and unique or undesirable liquid-pathway characteristics should be carefully weighed in future construction permit applications before deciding that this concept can be dismissed."
LdAPWR Response As part of the LdAPWR design risk assessment, Westinghouse will evaluate the potential benefits of a core-retention device. At present, the design is proceeding assuming such a feature is not
( cost-benefit effective.
E. Hydrogen Control Systems Discussion The NRC intends to require hydrogen control systems to deal with degraded -core accidents for all dry containments, ice condensor containments, and the Mark I, II, and III containments. In addition, they stated the cost-ef fectiveness of . combustible gas contr'o1 systems for accidents proceeding with core melt and vessel melt-through and large combustible gas releases should be examined for future construction permit applications.
l l J 3.2 13 NOVEMBER, 1983 WPWR-RC 0060e:1
I \
i WAPWR' Response 9i The subject of hydrogen control systems and their impact on the WAPWR design can be found in the following portions of this document:
Section 3.1, items 5,14, 33 and 34 l
Section 3.3, item 5 Section 4.0, item 26 Section 6.1.1, item 3 Section 6.1.2.1, item 4 F. Reliable Containment Heat Removal Discussion The NRC is studying the need for more reliable subsystems for containment heat removal as possible alternatives to filtered venting for prevention of gradual over-pressurization failure of the containment building. In addition, the NRC again emphasizes the need to assure high reliability of decay heat removal systems. Both of these items are to be addressed by applicants for standard design approvals.
WAPWR Response See Section 3.1, item 32 for a discussion of the WAPWR position relative to filtered-vented containments. WAPWR design features aimed at improving the reliability of decay heat removal systems are presented in Section 4.0, item 23.
G. Other Consequence Mitigation Measures The NRC recognizes that core-melt consequ,ence mitigation design features and procedures should be evaluated on as realistic a basis as O
WAPWR-RC 3.2-14 NOVEMBER, 1983 0060e:1
l l
l possible, i.e., such features are there to mitigate the consequences of extremely low probability events. Thus, the acceptance criteria for such features will be established accordingly. In addition, the f
l Aj NRC recognizes that such design and operational improvements for core-melt mitigation will have certain attendant risks; and that these must not be ignored.
WAPWR Response For the MAPWR, the attendant risks associated with any such core-melt mitigation design features will be factored into the probabilistic risk assessment.
H. External Events, Human Errors, and Sabotage Discussion D
The NRC expects that applicants for standard design approvals will
( address in the Safety Analysis Report the relation to severe accident considerations of sabotage and external events as well as other accident initiators such as multiple human errors and design errors.
MAPWR Response As discussed in the "MAFAR Response" to item 2 above and in Section
!O 5.1, item 29,~ the WAPWR .PRA will include internal and external hazards as ' well as other accident initiators. In addition, Westinghouse will perform a sabotage assessment for the MAPWR.
O O
3.2-15 NOVEMBER, 1983 WAPWR-RC 0060e:1
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- 1. Siting Policy Ol I i Discussion l
\
A modified siting rum applicable to future plants which incorporate new radioactive source term information for severe accidents is '
expected to be issued for comment in the near fut'ure.
WAPWR Response See Section 6.1.2.2, item 2, for a discussion of this issue and its impact on the WAPWR design.
- 9. Implementation Guidelines for Severe Accident Policy Discussion i
The NRC has established the following conditions for standard designs for reference in future construction permit applications or in reactivations of previously docketed construction permit applications:
o Demonstration of compliance with current Commission regulations o Completion of a PRA before a standard design can be taken through rulemaking o Completion of a Staff review of safety acceptablity; the review will be based upon the Standard Review Plan (NUREG-0800).
e o Consideration of all apL!icable Unresolved Safety issues o Adherence to the post-TMl requirements as set forth in the CP rule.
O WAPWR-RC 3.2-16 NOVEMBER,_1983 0060e:1
- . - - . . - - -_. - -. - -_ .. - . . - - . . _ = . _ _ _ . .. - . - - --- -- - .- - . - --_. --
d l WAPWR Response l
l The purpose of this " Regulatory Conformance" document is to establish the l
Westinghouse position for the WPWR design relative to NRC regulations, regulatory guidance and policy, and generic safety issues; including the i
five items listed above.
f l
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@PWR-RC 0060e:1
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O 3.3 OTHER POST-TM1 ISSUES 3.3.1 NUREG-0737 NUREG-0737, " Clarification of TMI Action Plan Requirements," contains those post-TM1 requirements that have been approved for impl,ementation by the Com-mission for operating plant licensees and applicants. In many cases, the specific requirements of NUREG-0737 are identical to those of NUREG-0718/10CFR 50.34 discussed in Section 3.1. However, the NRC has determined that certain, of the items contained in NUREG-0737 are not applicable at the construction permit stage and are, therefore, not included as requir'ments e in NUREG-0718/10CFR 50.34. Westinghouse believes that the NRC does not intend to imply that certain requirements imposed on operating or near-term operating plants i
are not applicable to a later vintage plant, but simply that certain require-ments can be more appropriately addressed at the operating license stage.
Therefore, the @ PWR design, will include appropriate consideration of the i
i additional requirements of NUREG-0737.
- 1. Pressurizer Water t.evel (NUREG-0737. Item II.K.l .17)
Discussion
! This item is really applicable to certain older generation Westinghouse operating plants that utilized a low pressurizer level coincident with low l
pressurizer pressure logic to provide a safety injection signal. This design feature is not utilized in current-day Westinghouse designs.
l L WPWR Response This design feature has not been included in the @ PWR design.
- 2. Thermal Mechanical Report - Ef fect of High Pressure Injection on Vessel Integrity f or small-Break EOCA with no Auxiliary Feedwater (NUREG-0737, item !!.K.2.13) l 3.3 1 NOVEM8ER, 1963
- FPWR-RC 0060e
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Discussion Refer to the discussion of Unresolved Safety Issue A-49, " Pressurized Thermal Shock," in Section 4.0 (item 27).
Installation and Testing of Automatic Power-Operated Relief Valve Isola-O 3.
tion System (NUREG-0737, item !!.K.3.1) l Discussion Refer to the discussion of item 4 in Section 3.1.
- 4. Automatic Trip of Reactor Coolant Pumps During LOCA (NUREG-0737, Item !
II.K.3.5)
Discussion For this item the NRC considered a requirement for plant designs to incor-porate automatic tripping of the reactor coolant pumps in the case of a small-break LOCA. 1his item and its impact on the WAPWR design is fully discussed in Section 6.4, Items 92 and 93.
- 5. Emergency Preparedness (NUREG-0737, Item III.A.2)
Discussign Refer to the discussion of 10CFR Part 50, Appendix E " Emergency Planning and Preparedness for Production and Utilization Facilities," in Section 6.1.1 (item 1).
3.3.2 NUREG-0660 Certain of the post IMI ssues identified in NUREG 0660, "NRC Action Plan Developed as a Result of the IMI -2 Accident," are related to on-going NRC I activities and are not included in the set of post-TM1 issues currently WAPWR-RC 3.3-2 NOVEMBER, 1983 0060e:1 .
O approved by the Commission for implementation through NUREG-0718/10CFR 50.34 or NUREG-0737. Since the NRC has made it clear that future requirements may result f rom the activities related to these issues, some level of considera-tion of the potential impact of these issues will be given in the design of the WAPWR.
In certain instances, the NRC program and activities are not currently well defined and appropriate consideration might be simply to be aware of the issue and possible courses of NRC action.
!O i
- 1. Control Room Design Stardard (NUREG-0660, item 1.D.4)
! Discussion f The NRC plans to develop guidance for the designs of future control f ,
rooms. This NRC guidance is anticipated to be in the form of a regulatory guide that endorses IEEE Standards 566 and 567 concerning the design of display and control functions and the design of the control room complex, respectively. The IEEE schedule for issuance of the post-TM1 versions of ,
l
! / these standards is uncertain.
\, .
n_dAPWR Response i
l Westinghouse will give appropriate consideration to any available draf ts
- of these standards in the design of the FPWR control room.
l l
- 2. Improved Control Room Instrumentation Research (NUREG-0660, Item 1.0.5)
O Discussion
' This item deals with NRC Of fice of Nuclear Regulatory Research initiated-studies aimed at developing new (longer term) instrumentation to enhance the performarce of the control room operator.
l O 3.3-3 NOVEM8ER,11983 WPWR-RC 0060e:1 l
1
The specific studies are related to:
o Alarms and displays for improving the man-machine interface in reactor control rooms.
o Plant status monitoring to improve the ability of reactor opera-tors to prevent, diagnose, and properly respond to accidents.
o On-line reactor surveillance utilizing noise diagnostic and pattern recognition techniques.
o Disturbance analysis system feasibility and development.
WAPWR Response These studies are in various degrees of completion; however, Westinghouse will give appropriate consideration to any available results in the design of the WAPWR control room.
- 3. Siting Policy Reformulation (NUREG-0660 Item II.A.1)
Discussion Refer to the discussion of the NRC advanced notice of rulemaking concern-ing reactor siting criteria in Section 6.1.2.2 (item 2).
- 4. Research on Phenomena Associated with Core Degradation and Fuel Melting (NUREG-0660 Item II.B.5) l Discussion The NRC Of fice of Nuclear Regulatory Research is conducting major research O
programs associated with core degradation and fuel melting. These pro-
, grams are intended to support the basis for rulemaking and confirm certain 1
O l WAPWR-RC 3.3-4 NOVEMBER, 1983
. 0060e:1
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O licensing decisions related to core degradation and fuel melting. Refer to the discussion of the NRC activities related to degraded or melted cores in Section 3.2.
O 5. Analysis of Hydrogen Control (NUREG-0660, item II.B.7)
Discussion I
'- Specific NUREG-0718/10CFR 50.34 requirements related to hydrogen control for the WAPWR design are discussed in Section 3.1 (item 14).
- 6. Continuation of Interim Reliability Evaluation Program (NUREG-0660, item II.C.2)
Discussion This item deals with possible future NRC requirements for operating plant
. / licensees to perform probabilistic reliability / risk studies. Specific requirements related to probabilistic risk assessment for the WAPWR design l are discussed in Section 3.2.
i
- 7. Systems Interaction (NUREG-0660, item 11.C.3)
Discussion This item is actually a subpart of the overall issue of " Systems Interac-l tions in Nuclear Power Plants" (Unresolved Safety Issue A-17). Refer to l
the discussion of Unresolved Safety Issue A-17 in Section 4.0 (item 13).
- 8. Update Standard Review Plan and Develop Regulatory Guide (NUREG-0660, item O 11.E.1.3)
Discussion l This item deals with NRC activities related to:
WAPWR-RC 3.3-5 NOVEMBER, 1983 0060e:1 l
l l- - - - - . - _ _ _ _ _ . . _ _ . _ . _ _ _ , _ _ _ . _ .__. _ _ _ _ _ _ ,_. _
l o Updating Standard Review Plan 10.4.9, " Auxiliary Feedwater g
System," to include IMI-2 lessons learned recommendations /
requirements.
l o issuing a new regulatory guide on auxiliary feedwater system designs that will possibly endorse ANSI /ANS 51 .1 0 " Auxiliary Feedwater System for Pressurized Water Reactors."
The NRC Standard Review Plan is discussed in Section 6.3. In relation to Standard Review Plan 10.4.9, the 1MI-2 lessons learned reconsnendations/
requirements discussed in Section 3.1 (items 2 and 17) have been included in the latest version (i.e., Revision 2: July 1981).
WAPWR Response Westinghouse has given appropriate consideration to the criteria of ANSI /ANS 51.10 in the design of the WAPWR secondary side safeguards capa-bility.
- 9. Reliance on Emergency Core Cooling System (NUREG-0660, item II.E.2.1)
Discussion This issue involves a potential deficiency in the reliability of emergency core cooling systems (ECCS). The concern results from a higher than anti-cipated frequency of ECCS challenges in operating reactors, in part because of the reliance on ECCS for other than loss-of-coolant accidents.
The reliability of ECCS is believed to be high, but it is not clear that it is sufficiently high to accomplish its safety function with high assur-ance, considering the increase in expected challenges. Further study has been recommended by the NRC to determine if this issue should be reported as an Unresolved Safety Issue. The further study would be in the form of scoping calculations related to ECCS challenges and reliability.
O 3.3-6 NOVEMBER, 1983 WAPWR-RC 0060e:1 l
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~WAPWR Response
' lhis issue is not applicable to the. WAPWR Emergency Core Cooling System (ECCS). lhe WAPWR ECCS reliability has been significantly improved over the current operating reactor by utilizing four high head safety injection pun 9s, four core reflood tanks, and four accumulators exclusively to per-form the required ECCS function. Consequently, the anticipated f requ'ency of ECCS challenges has been essentially eliminated as compared to current operating reactors. .
- 10. Research on Small-Break LOCAs and Anomalous Transients (NUREG-0660, Item II.E.2.2)
Discussion i
i The NRC is conducting research that focuses on small-breaks and transi-ents, including experimental research at the LOF1 facility, systems engin-4
/O
\.
eering, and materials effects programs.
Westinghouse . typically follows NRC-sponsored research programs to ensure the applicability and acceptability of Westinghouse codes to accurately predict the consequences of postulated LOCAs. When appropriate, Westing-house also provides test predictions related to specific test programs to the NRC for review and comparison purposes.
I WAPWR Response
- O Beyond following these research programs which puts Westinghouse in a pos-ition to possibly identify any potential analytical or hardware-related problem areas, consideration of this issue in the dcvelopment'of the WAPWR O design is not appropriate.
O 3.3-7 NOVEMBER, 1983 IIAPWR-RC L
0060e:1
- 11. Decay Heat Removal Systems Reliability and Coordinated Study of Shutdown Heat Removal Requirements (NUREG-0660, items II.E.3.2 and 11.E.3.3)
Discussion O
These items are actually subparts of the overall issue of " Shutdown Decay Heat Removal Requirements" (Unresolved Safety Issue A-45). Refer to the
discussion of Unresolved Safety issue A-45 in Section 4.0 (item 23).
- 12. Decay Heat Removal Alternate Concepts Research (NUREG-0660, Item II.E.3.4)
O Discussion The NRC plans to determine the technical feasibility of passive contain-ment cooling including add-on decay heat removal systems for new plants and possible backfitting to existing plants.
WAPWR Response Since this issue and its ultimate resolution are not sufficiently defined to permit appropriate design consideration, Westinghouse plans to follow NRC activities in relation to this issue in lieu of arbitrarily specifying requirements for the WAPWR design. Containment cooling for the WAPWR is performed by the containment spray and containment fan coolers. Passive features are not included in the design.
- 13. Decay Heat Removal Regulatory Guide (NUREG-0660, item II.E.3.5)
Discussion i
The NRC plans to provide improved guidance on the reliability and capabil-ity of nuclear power plant systems for removing decay heat and achieving l
safe shutdown conditions following transients and under post-accident )
I O
l WAPWR-RC 3.3-8 NOVEMBER, 1983 0060e:1
O conditions. This guidance will be in t!.e form of Revision 1 to Regulatory Guide 1.139, " Guidance for Residual Heat Removal."
WAPWR Response Westinghouse will address Revision 1 to Regulatory Guide 1.139 in relation to the WAPWR design when it is issued by the NRC.
- 14. Study of Control and Protective Action Design Requirements (NUREG-0660 Item II.F.4)
Discussion This issue involves a potential deficiency related to: (A) basing protec-tive actions on derived variables rather than direct reading of process variables; (B) protective actions relying on coircidence of independent process variables rather than relying on either variable; and (C) lack of g
testing of control circuit components at expected degraded power supply conditions. The NRC believes that existing requirements ,already preclude
.these deficiencies.
' ~
WAPWR Response l
Westinghouse agrees with the NRC that this issue does not present a sig -
, nificant safety problem. However, the WAPWR protection system will be designed to meet all applicable safety requirements.
- 15. Classification of Instrumentation, Control, and Electrical Equipment (NUREG-0660. Item II.F.5)
O Discussion The NRC planned to prepare a standard (in conjunction with IEEE) and a regulatory guide that endorses the standard that provides a classification 1
3.3 -9 NOVEMBER, 1983 WAPWR-RC (
0060e:1
approach for determining the applicability of design criteria and design requirements for plant Instrumentation, control, and electrical systems and equipment based on the level of their importance to safety. This standard, IEEE P-827, was drafted, but subsequently withdrawn by the IEEE. The industry cooperative ef fort planned via IEEE P-827 has been replaced by the ANS 51.1 effort.
WAPWR Response See Section 6.1.2.1, item 5 for a discussion of Westinghouse activities relative to ANS 51.1.
- 16. Nuclear Data Link (NUREG-0660. Item Ill.A.3.4)
Discussion This item will (when finalized) require each utility to provide equipment and interface with the NRC data acquisition system to remotely access facility data and transmit the data and display information in the NRC Operations Center.
Although not finalized (in terms of issuance for implementation), the NRC criteria for the nuclear data link are provided in NUREG-0696, " Functional Criteria for Emergency Response Facilities."
WAPWR Resp 5nse The WAPWR design for instrumentation to be incorporated as part of the onsite technical support center includes an output interf ace for of f site data comunication. Westinghouse will give appropriate consideration to of fsite data comunication equipment for the WAPWR design upon finaliza-tion of NRC requirements in this area.
O 3.3-10 NOVEMBER, 1983 WAPWR-RC 0060e:1
5O 17. Radioactive Gas Management (NUREG-0660, Item III.D.I.2)
Discussion LO
! The NRC plans to sponsor a future study to determine the applicability and desirability of use of available technology to minimize the release of
) radioactive noble gases during and following various postulated accident conditions.
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) At this time it is not clear if this item will have any impact on Westing-house designs including the WAPWR design, i
MAPWR Response I
Since this issue and its ultimate resolution are not suf ficiently defined i to permit appropriate design consideration, Westinghouse plans to follow NRC activities in relation to this issue in lieu of arbitrarily specifying requirements for the MAPWR design. Presently, the design of WAPWR gaseous j (~
t waste management systems is based on normal operation only.
j 18. Ventilation System and Radiciodine Adsorber Criteria (NUREG-0660 Item
- 111.0.1.3)
! Oiscussion 1he NRC plans to develop future requirements for ensuring that there is f adequate filtration of radioactivity in ventilation exhausts and that l
acceptable collection ef ficiencies of radiotodine adsorbers are maintained l.
during accident conditions.
The NRC has indicated that their new requirements / recommendations will be issued as revisions to Regulatory Guides 1.52 and 1.140.
d lO 3.3 11 NOVEMBER, 1983 WAPWR-RC i 0060e:1
I WAPWR Response O
Westinghouse will address any future revisions to Regulatory Guides 1.52 and 1.140 in relation to the WAPWR design when they are issued by the NRC.
- 19. Radwaste System Design Features to Aid in Accident Recovery and Decontam-ination (NUREG-0660, Item 111.0.1.4)
Discussion the NRC plans to sponsor a future evaluation of radwaste system design features that will provide the capability to process accident-related liquids and gases and to conduct decontamination effectively.
WAPWR Response Since this issue and its ultimate resolution are not sufficiently defined to p.ermit appropriate design consideration, Westinghouse plans to follow NRC activities in relation to this issue in lieu of arbitrarily specifying requirements for the WAPWR design. Presently, the design of WAPWR rad-waste systems is based on normal operation only.
- 20. Radiological Monitoring of Effluents (NUREG-0660. Item III.O.2.1)
Discussion ,
The NRC plans to develop future requirements for revised systems for radiological monitoring of effluents (e.g., development of atmospheric steam dump monitoring of both noble gas and radiotodine af ter an accident).
The NRC has indicated that their new requirements / recommendations will be issued as revisions to Regulatory Guides 1.21 and 1.97.
O NOVEMBER, 1983 WAPWR-RC 3.3-12 0060e:1
O WAPWR Response Westinghouse will address any future revisions to Regulatory Guides 1.21 l and 1.97 in relation to the WAPWR design when they are issued by the NRC.
i 21. Offsite 00se Measurements (NUREG-0660, Item 111.0.2.4)
. Discussion The NRC plans to sponsor a future study of the feasibility of environment-al monitors capable of measuring real-time rates of exposures to noble gases and radioiodines.
4 WAPWR Response This issue is not applicable to Westinghouse in relation to the LdAPWR design. Environmental monitors a. e the responsibility of each utility
, O utilizing the @ PWR design.
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- O 3.3-13 NOVEMBER, 1983 WAPWR-RC 0060e
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