ML20044G240
| ML20044G240 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 05/28/1993 |
| From: | Liparulo N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Borchardt R NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| ET-NRC-93-3895, NUDOCS 9306020220 | |
| Download: ML20044G240 (133) | |
Text
{{#Wiki_filter:. -. -- - - - -. Westinghouse Energy Systems Box 355 -: P ttsburgh Pennsylvania 15230 0355 l Electric Corporation
' ET-NRC-93-3895 '
NSRA-APSL-93-0192 I l Docket No.: STN-52-003 i May 28,1993 i Document Control Desk l U.S. NucIcar Regulatory Commission Washington, D.C. 20555 ATTENTION: R.W.BORCHARDT , l
SUBJECT:
WESTINGHOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL' l INFORMATION ON THE AP600 - ; l- '
Dear Mr. Borchardt:
a 6 Enclosed are three copies of the Westinghouse responses to NRC requests for additional information . on the AP600 from your letters of January 26,1993, March 12,~ 1993 and April 13,1993. This ; transmittal completes the responses to the January 26,1993 letter. A listing of the NRC requests for '! additional information responded to in this letter is contained in Attachment A. Attachment B is a complete listing of the questions associated with the January 26,-1993 letter and the corresponding ., Westinghouse letters that provided our response. , q If you have any questions on this material, please contact Mr. Brian A. McIntyre at 412-374-4334. : A [ - j Nicholas J. Uparulo, Monager l Nuclear Safety & Regulatory Activities l
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i ET-NRC-93-3895 A'ITACHMENT A AP600 RAI RESPONSES SUBMITTED MAY 28,1993 ; l RAINo. Issue 100.008 l USIs and GSis 210.027 l May 7,1992 RAI on Valves i 210.028 l Valve Design and Reliability Assumptions i 251.011R01l Containment of RCP flywheel rupture 280.004 i HWRF Test Data 410.098 l HVAC conformance to GDCs (WCAP-13053) 420.009 i Conformance to software standards ) 420 010 l Conformance to EMI/RFI standards i 420.019 i LLNL report on IEEE Std. 796-1983 i 420.036 i Testing of protection system actuated equipment l 420.039 l IPS/ICS interface 420.044 l Transmission of post accident monitoring info 420.052 i Extreme environmental & energy supply conditions < 420.053 i FMEA, protection cabinet power supply arrangement t 420.086 i Time response during upset condition 450.002R01l Control room habitability 471.007 i Steam generator manway ease of entry 471.008 i Overflow lines into waste collection system 471.012 i Limitations on cobalt impurity content 471.016 i Containment monitoring activities l 471.017 l Radiation compartment wall & floor coatings 630.007 i Tech specs versus standard tech specs 630.009 l Tech spec completion times & surveillance interval 1
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Printed: 05/27/93 1 ATTACHMENT B CROSS REFERENCE OF WESTINGHOUSE RAI RESPONSE TRANSMITTALS TO NRC LETTER OF JANUARY 26,1993 Question issue NRC Westinghouse No. Letter Transmittat Date 100.008 USis and GSis 01/ & 93 05/2&S3 210.027 May 7,1992 RAI on Valves 01/2693 05/28/93 210.028 Valve Design and Reliability Assumptons 01/26/93 0528/93 410.093 Classificaten of Structures, Systems, Components 01/2S 93 03/1&93 410.094 HVAC system drawings 01/2693 03/1 &93 410.095 HVAC conformance to SRP 01/26/93 03/30/93 410.096 Control room ventilation system (WCAP-13053) 01/26/93 03/3G93 410.097 VBS support for control room design basis 01/26/93 01'30/93 410.098 HVAC conformance to GDCs (WCAP-13053) 01/2693 05/28/93 410.099 SFP/ Aux /Radweste area ventilation (WCAP-13053) 01/26/93 05/1493 410.100 VCS & VFS conformance to SRP (WCAP-13053) 01/26/93 0422/93 410.101 VRS cordormance to SRP (WCAP-13053) 01/26/93 05/1493 410.102 Turbine tAdg ventilaten (WCAP.13053) 01/2693 04/29/93 410.103 Diesel generator Building HVAC (WCAP-13053) 01/2G93 03/30/93 410.104 HP & hot rnachina shop HVAC (WCAP-13053) 01/2693 05/1493 410.105 Turtune building closed cooling water system 01/26/93 03/30/93 440.032 Emergency response guidelines 01/2693 05/1493 620.050 Display /controllalarm matrix 01/2G93 04/29/93 630.006 LCOs for passive systems 01/26,93 03/30/93 630.007 Tech specs versus standard tech specs 01/26/93 052&93 630.008 Tech spec topical reports 01/2693 03/30/93 630 009 Tech spec completon times & surveillance interval 01/26/93 0528/93 Records printed: 22 l
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NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 100.8 Section 1.9.4 of the AP600 Standard Safety Analysis Report (SSAR) addresses generic safety issues. 10 CFR 52.47(a)(1)(iv) requires the applicant to propose technical resolutions to those Unresolved Safety Issues (USIs) and medium- and high-priority Generic Safety Issues (GSIs) that are identified in the version of NUREG-0933 current on the date six months prior to application and which are technically relevant to the design. The version of NUREG-0933 that is applicable to AP600 is dated July 1991, as correctly stated in the SSAR. However, Section 1.9.4 of the SS AR provides proposed resolutions only to those issues categorized in NUREG-CA10, "NRC Program for the Resolution of Generic Issues Related to Nuclear Power Plants," dated January 1978. As a result, it does l not appear that all applicable generic issues have been addressed. For example, Generic Issue 15, " Radiation Effects on Reactor Vessel Supports," which is high-priority as stated in NUREG-0933, is not addressed in the SSAR. Provide the proposed resolutions of all USIs and GSis in the SSAR in accordance with 10 CFR 52.47(a)(1)(iv). Response: I SSAR Subsection 1.9.4 will be revised to include proposed resolutions of all USIs and GSis in accordance with 10 CFR 52.47(a)(1)(iv). SSAR Revision: j The proposed revision to SSAR Subsection 1.9.4 is attached. This revision to Subsection 1.9.4 will replace the existing SSAR Subsection 1.9.4. A proposed revision to SSAR Subsection 3.9.1.1.2.11 follows. . 3.9.1.1.2.11 Loss of Power with Natural Circulation Cooldown This event is the same as a loss of power transient, except that the reactor coolant system temperature is reduced to cold conditions by natural circulation through the operation of either the startup feedwater pumps and l steam dump through the power-operated relief valves or the passive residual heat removal system transferring heat to the in-containment refueling water storage tank. For design purposes 30 natural cooldown occurrences-For ; hy+pepw4htma! =! ' aw- are assumed. [ WOStingh00SB
I l l NRC REQUEST FOR ADDITIONAL INFORMATION
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1.9 COMPLIANCE WITH REGULATORY CRITERIA 1.9.4 Unresolved Safety issues and reviewed to determine which issues are technically Generic Safety issues relevant to the AP600 design. In this review process, the following screening criteria were applied: Proposed technical resolmions of Unresolved Safety Issues and medium- and high-priority Generic Safety a. Issue has been priontized as Low, Drop, or has not Issues, as identified in NUREG-0933, Reference 3 are been prioritized. required for new plants as part of the NRC policy on b. Issue is not an AP600 design issue. Issue is appli-severe accidents and are required for design certification cable to GE, B&W, or CE designs only. in accordance with 10 CFR 52.47(a)(1)(iv). c. Issue resolved with no new requirements. The current program for identifying and establishing d. Issue is not a design issue (Environmental Issue, the priority of open safety issues is summarized in Licensing Issue, Regulatory Impact issue, or cov-NUREG-0933. This program provides for the priori. ered in an existing NRC program). tization and tracking of previously categorized Unre e. Issue superseded by one or more issues. solved Safety Issues and Generic Safety Issues, New f. Issue is not an AP600 design certification issue. Generic Issues, TMI Action Plan items Under Devel- Issue is applicable to NTOL plants only, responsi-opment, and Human Factors Program Plan Issues. bility of combined license applicant, or issue is ne following subsection reviews each of the limited to current generation operating plants. NUREG-0933 safety issues and identifies the safety issues that are applicable to the AP600. For each of Issues meeting one or more of the preceding screening these issues guidance is provided on how the issue is ad. criteria were screened out of the review process as dressed for the AP600. issues that are not applicable to the AP600 design. The remaining issues fall into one of the following two 1.9.4.1 Review of NRC List of Unresolved categories: Safety issues and Generic Safety
- g. Issue is res lved by establishment of new regulatory issues requirements and/or guidance.
Applicants for design certification are required by 10
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CFR 52.47(a)(1)(iv) to identify: prioritized as Ifigh, Medium, or possible resolution identified).
*/pfroposed technical resolutions of those Table 1.9-2 identifies the results of the screening Unresohed Safety Issues and medium- and review. For those issues identified as relevant to the high-priority Generic Safety Issues which are AP600 design (i.e., issues screened as g or h), Table idenii/ icd in the ersion of NUREG-0933 cur- 1.9-2 identifies the SSAR subsection that addresses the rent on the date six months prior to apphcatuon ; s,u,,
and which are technically relevant to the de-1.9.4.2 AP600 Resolution of Unresolved NUREG-0933, "A Prioritization of Generic Safety Safety issues and Generic Safety issues," through Supplement 12, dated July 1991, Issues identifies a total of 767 issues. These issues were I.D.5(2) Plant Status and Post-Accident Monitoring 100.8-2 W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Discussion: Discussion: TM1 action plant item 1.D.5(2) addresses the meed TMI action plan item I.D.5(3) addresses the benefit to improve the operators' ability to prevent, diagnose to plant safety and operations of continuous on-line and properly respond to accidents. The emphasis is on automated surveillance systems. Continuous on-line the information needs (i.e., indication of plant status) of surveillance systems that automatically monitor reactors the operator. This issue was resolved with the issuance can benefit plant operations and safety by providing of Revision 2 to Regulatory Guide 1.97, *lnstrumenta- diagnostic information which can predict anomalous tion for Light Water Cooled Nuclear Power Plants to behavior and thus be used to maintain safe conditions. Assess Plant Environs Conditions During and Following Various methods of on-line reactor surveillance an Accident." have been used, including neutron noise monitoring in boiling water reactors (BWRs) to detect internals AP600 Response: vibration, and pressure noise surveillance at TMI-2 to Re AP600 conforms to and meets the intent of monitor primary loop degasification. Regulatory Guide 1.97. Regulatory Guide 1.97 pmvides the requirements for post-accident monitoring of nuclear AP600 Response: reactor safety parameters, including plant process The AP600 reactor coolant pressure boundary is parameters important to safety and the monitoring of monitored for leaks from the reactor coolant and associ-effluent paths and plant environs for radioactivity. ated systems by a variety of components located in These guidelines include definition and categorization of multiple systems. The leak detection system provides plant variables that are available te the main control information permitting the plant operators to take room operators for monitoring the plant safety status corrective action if any detected leakage exceeds techni-following a design basis event. cal specifications. The leak detection system is designed For the AP600, an analysis is conducted to identify according to the requirements of 10 CFR 50, Appendix the appropriate variables and to establish the appropriate A, General Design Cdterion 30. The syste.m provides design basis and qualification criteria for instrumentation a means to detect and, to the extent practical, to identify used by the operator for monitoring conditions in the the source of the reactor coolant pressure boundary reactor coolant system, the secondary heat removal leakage. The systems conform with the recommenda-system, the containment, and the systems used for tions in Regulatory Guide 1.45, except that no credit is attaining a safe shutdown condition, as discussed in taken for the airborne particulate monitor to quantify a Section 7.5. leak rate. SSAR Subsection 5.2.5 provides further ne instrumentation is used by the operator to discussion of leak detection. monitor and maintain the safety of the plant during A digital metal impact monitoring system (DMIMS) operating conditions, including anticipated operational monitors the reactor coolant system for the presence of occurrences and accident and post-accident conditions. loose metallic parts. This system conforms with the The plant parameters identified to satisfy the Regulatory guidance provided in Regulatory Guide 1.133, Rev.1, Guide 1.97 guidelines are processed and displayed by May 1981. An advanced microprocessor-based system, the qualified data processing system (QDPS), which is employing digital technology, automatically actuates ; discussed in Subsection 18.9.5. The verification and audible and visual alarms if a signal exceeds the preset ; validation (V&V) of the QDPS complies with the same alarm level. V&V process described in Section 18.8. II.K.3(9) Proportional Integral Derivative Controller j 1.D.5(3) On-Line Reactor Sur eillance System Modification i W-Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION
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Discussion: system, service water system, feedwater system, and TMI action plan item II.K.3(9) required all West- steam lines. inghouse plants to raise the interlock bistable trip setting Water hammer issues are considered in the design to preclude derivative action from opening the PORVs. of the AP600 passive core woling system. Thepassive core cooling system design includes a number of design AP600 Response: features speci6cally to prevent or mitigate water ham-This isrue is not applicable to the AP600. The mer. AP600 does not include power-operated relief valves. The automatic depressurization system operation uses multiple, sequenced valve stages to provide a A-1 Water llammer relatively slow, controlled depressurization of the reactor coolant system, which helps to reduce the potential for Discussion: water hammer. Generic Safety Issue A-1 was raised after the Once the depressurization is complete, gravity occurrence of various incidents of water hammer that injection from the in-containment refueling water storage involved steam generator feedrings and piping, emer- tank is initiated by opening check valves, which reposi-gency core cooling systems, residual heat removal tion slowly. Gravity injection flow actuates slowly, systems, containment spray, service water, feedwater, without water hammer, as the pressure differential and steam lines. The incidents have been attributed to across the gravity injection check valves equalizes, and such causes as rapid condensation of steam pockets, the valves open and initiate flow. steam-driven slugs of water, pump startup with partially The passive residual heat removal heat exchangers empty lines, and rapid valve motion. Most of the are normally aligned with open inlet valves and closed damage has been relatively minor and involved pipe discharge valves. This alignment keeps the system hangers and restraints. However, several incidents have piping at reactor coolant system pressure, preventing l resulted in piping and valve damage. This item was water hammer upon initiation of flow through the heat originally identified in NUREG-0371, (Reference 4) and exchangers. was later determined to be an Unresolved Safety Issue. The core makeup tanks are normally aligned with an i open pressurizer pressure balance line to keep the tanks AP600 Response: at reactor coolant system pressure. In addition, the Specific sections of the Standard Review Plan pressurizer pressure balance line is normally kept filled , (NUREG-0800) address criteria for mitigation of water with steam to prevent water hammer upon core makeup hammer concerns. The applicable Standard Review tank actuation. Section 6.3 of the SSAR provides j Plan sections as well as information provided in additional information on the passive core cooling ' NUREG-0927 (Reference 5) were reviewed. The system. AP600 meets the water hammer provisions as specified. The potential for water hammer in the feedwater j The discussion that follows provides a brief description line is minimized by the improved design and operation ; of selected systems identified as being subject to water of the feedwater delivery system. The steam generator i hammer occurrences and special design features that features include introducing feedwater into the steam mitigate or prevent water hammer damage. generator at an elevation above the top of the tube Design features are incorporated as appropriate to bundles and below the normal water level by a top prevent water hammer damage in applicable systems discharge J-tube feedring. The layout of the feedwater I l including steam generator feedrings and piping, passive line reduces the magnitude of a steam generator water core cooling system, passive residual heat removal hammer. Additionally, operational limitations on flow to recover steam generator levels and on early feedwater 100.8-4 W Westinghouse l
NRC REQUEST FOR ADDITIONAL INFORMATION n um J !!! flow into the steam generator to maintain the feedring pressurization including reactor cavity asymmetric full of water minimize the potential for water hammer. pressurization transients, and traveling pressure waves The startup feedwater system is a non-safety-related from the depressurization of the system. system that provides heated feedwater during normal The AP600 reactor coolant loop and pressurizer plant startup, shutdown, and hot standby. The heated surge line are designed in accordance with mechanistic 1 feedwater reduces the potential of water hammer in the pipe break criteria. In addition, other high energy l feedwater piping or steam generator feedrings. ASME Code, Section III, Class I and 2 piping of 4 The main steam line drains are designed to remove inches and greater nominal diameter is evaluated against ; accumulated condensate from the main steam lines and leak-before-break criteria. The evaluation methodology ! to maintain the turbine bypass header at operating is described in Subsection 3.6.3 and Appendix 3B. ) temperature during plant operation. The system is ! designed to accommodate drain flows during startup, A-3 Steam Generator Tube Integrity shutdown, transient, and normal operation to protect the j turbine and the turbine bypass valves from water slug Discussion: damage. Pressurized water reactor steam generator tube , integrity is subject to various degradation mechanisms, ! A-2 Asymmetric Blowdown Loads on Reactor including corrosion-ind uced wastage, cracking, reduction Primary Coolant Systems in tube diameter, denting, (which leads to primary side stress corrosion cracking), vibration-induced fatigue Discussion: cracks, and wear or fretting due to loose parts in the Generic Safety Issue A-2 pertains to asymmetric secondary system. The primary concern is the capabili-loadings that could act on a pressurized water reactor's ty of degraded tubes to maintain their integrity during primary system as the result of a postulated normal operation and under accident conditions (LOCA l double-ended rupture of the piping in the primary or a main steam line break) with adequate safety mar- l coolant system. The magnitude of these loads is poten- gins. i tially large enough to damage the supports of the reactor Steam generator tube integrity concerns for the i vessel, the reactor internals, and other primary compo- three steam generator suppliers, Westinghouse, Combus- ) nents of the system. Therefore, the NRC initiated a tion Engineering, and Babcock and Wilcox, are ad- ! generic study to develop criteria for an evaluation of the dressed by an integrated NRC program for Generic l response of the primary systems in pressurized water Safety Issues A3, A,4 and A5. This program addresses reactors to these loads. the areas of steam generator integrity, plant systems response, human factors, radiological consequences, and AP600 Response: the response of various organizations to a steam genera-The use of mechanistic pipe break criteria permits tor tube rupture. elimination of the evaluation of dynamic effects of sudden circumferential and longitudinal pipe breaks in AP600 Response: 1 the structural analysis of structures, systems, and The AP600 steam generators are designed in components. General Design Criterion 4 allows the use accordance with the recommendations of Generic Letter of analyses to eliminate from the design basis the 85-02 and NUREG-0844 (References 6 and 7). He dynamic effects of pipe ruptures postulated at locations AP600 steam generator is equipped with a number of defined in Subsection 3.6.2. Dynamic effects includejet features to enhance steam generator tube performance impingement, pipe whip, jet reaction forces on other and reliability. These features are described in Subsec-portions of the piping and components, subcompartment tion 5.4.2. W-Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION
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A-9 Anticipated Transients Without Scram AP600 Response: He steam generator and reactor coolant pump Discussion: supports are described in Subsection 5.4.10. The Generic Safety Issue A-9 was resolved with the supports are designed in accordance with subsection NF publication of 10 CFR 50.62. This regulation sets forth of Section Ill of the ASME Code. Design and fabrica-the requirements for reduction of risks from anticipated tion of these supports in accordance with Subsection NF transients without scram. requirements provide acceptable fracture toughness of materials, and conform with NUREG-0577. .AP600 Response: The AP600 design complies with the requirements of 10 CFR 50.62. A discussion of the AP600 design A-13 Snubber Operability Assurance features used to address the probability of an ATWS is presented in Subsection 1.9.5 and Section 7.7. Discussion: Generic issue A-13 addresses snubber operability A-11 Reactor Vessel Materials Toughness concerns. Snubbers are utilized primarily as seismic and pipe whip restraints at nuclear power plants. Heir Discussion: safety function is to operate as rigid supports for Generie Issue A-11 addresses a concern with the restraining the motion of attached systems or compo-reduction of reactor vessel fracture toughness as plants nents under rapidly applied load conditions such as accumulate more and more service time. 10 CFR 50, earthquakes, pipe breaks, and severe hydraulic tran-Appendix G provides requirements for reactor vessel sients. rraterial toughness. Operating experience reports show that a substantial number of snubbers have leaked hydraulic fluid and that AP600 Response: the rejection rate from functional testing and inspection The AP600 reactor vessel design complies with the is high. This has led to an NRC and ACRS concern requirements of 10 CFR 50, Appendix G and includes regarding the effect of snubber malfunctions on plant numerous features to reduce neutron fluence, enhance safety. material toughness at low temperature and eliminate weld seams in critical areas. Material requirements are AP600 Response: provided in Subsection 5.3.2. Pressure and temperature The use of snubbers is minimized in the AP600. limits are provided in Subsection 5.3.3. Gaped support devices, leak-be fore-break considerations, and state-of-the-art piping analysis methods are used to A-12 Fracture Toughness of Steam Generator and minimize the use of snubbers. Snubbers applied in Reactor Coolant Pump Supports safety-related applications are constructed to ASME Code, Section 111, Subsection NF as discussed in SSAR Discussion: Subsection 3.9.3.4.3. Generic Safety Issue A-12 addresses a concern with the potential for lamellar tearing of steam generator and A-24 Qualification of Class IE Safety-Related RCP support material. NUREG-0577 (Reference 8) Equipment categorizes operating plants relative to the adequacy of the plant's steam generator and reactor coolant pump Discussion: supports with respect to fracture toughness. Generic Issue A-24 was resolved with the publica-tion of 10 CFR 50.49, prescribing aging and testing for 100.8-6 W - Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Q 5 e synergistic effects. He NRC has also issued Revision I monitoring system. The non-Class 1E de and UPS to Regulatory Guide 1.89 for comment. The proposed power system supplies non-safety-related loads that revision desenbes a method acceptable to the NRC staff protect certain equipment and support plant control and to demonstrate compliance with the requirements of 10 monitoring systems. Subsection 8.3.2.1.2 provides a CFR 50.49. discussion on the Class IE power source. AP600 Response: A-26 Reactor Vessel Pressure Transient Protection he AP600environmentalqualificationmethodology described in Section 3.11 is based on the generic Discusdon: Westinghouse qualification program approved by the Generic issue A-26 addresses the need to provide NRC. The Westinghouse methodology addresses the reactor vessel overpressure protection whenever plants requirements of General Design Criteria 4 and 10 CFR are in a cold shutdown condition. Branch Technical 50.49, as well as the guidance of Regulatory Guide 1.89 Position RSB 5-2 establishes the current NRC criteria and IEEE Standard 323-1983. See Appendix 1 A and for a low-temperature overpressurization protection Reference 9. system. A-25 Non-Safety leads on Class 1E Power Source AP600 Response: The AP600 conforms with the criteria established in Discussion: Branch Technical Position RSB 5-2. He AP600 Generic Issue A-25 addresses whether non-safety pressurizer is sized to accommodate most pressure loads should be allowed to share Class IE power transients. Overp--ssure protection for the reactor sources with safety-related plant systems. Past regulato- coolant system is provnied by either the pressurircr ry practice has allowed the connection of non-safety safety valves or the normal residual heat removal relief loads in addition to the required safety loads to Class 1 E valves, as described in Subsection 5.2.2. power sources by imposing some restrictions. The purpose of this issue is for the NRC to determine A-18 Increase in Spent Fuel Pool Storage Capacity whether the reliability of the Class IE power sources is significantly affected by the sharing of safety and non- Discussion: safety loads. Generir issue A-28 addresses the safety significance The NRC considers this issue as technically re- of damage to spent fuel, primarily from a lack of solved with the issuance of Revision 2 to Regulatory adequate cooling, that could result in the release of Guide 1.75. This regulatory guide includes special radioactivity. requirements for connection of non-safety loads to a Class 1E source. AP600 Resp (mse: The AP600 design incorporates the NRC criteria, AP600 Response: and the heat load is evaluated for the spent fuel storage ne AP600 conforms with the criteria of Regulatory capacity. Guide 1.75 (see Appendix 1 A and Reference 17). The AP600 safety-related power source is the Class 1E de and UPS system, which supplies power to the ac inver-ters for the plant instrumentation and control systems. He system also provides power to de loads associated with the four protection channels and the accident W westinghouse l
NRC REQUEST FOR ADDITIONALINFORMATION wt m!i
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_ t A-31 Residual IIcat Removal Requirements sources. See Section 6.3 for a description of passive core cooling. Discussion: l Generic Issue A-31 addresses the desire for plants A-33 NEPA Reviews of Accident Risks l Lo be able to go from hot-standby to cold-shutdown conditions (when this is determined to be the safest Discussion: course of action) under any accident condition. He safe Generic Issue A-33 addresses the need to assess the shutdown of a nuclear power plant following an accident possible impact of accidents on the environment. This not related to a loss-of-coolant accident has been typical- issue was addressed by SECY-80-131. The policy i ly interpreted as achieving a hot standby condition (the statement required investigating in-plant accident se-reactor is shut down, but system temperature and pres- quences that can lead to a spectrum of releases. The l sure are at or near normal operating values). There are extent to which events arising from causes external to events that require eventual cooldown and long-term the plant that are considered possible contributors to risk cooling to perform mspection and repairs. shall also be discussed. 1 I ) I AP600 Response: The AP600 employs safety-related core decay heat AP600 Response-The consideration of possible impacts from acci- l' removal systems that establish and maintain the plant in dents on the environment is included in the AP600 a safe shutdown condition following design basis events. Probabilistic Risk Assessment Report. This assessment it is not necessary that these passive systems achieve includes the AP600 design on a generic site. A site- J cold shutdown as defined by Regulatory Guide 1.139. specific assessment is performed by the combined ; The AP600 complies with General Design Criteria license applicant. l 34 by using a more reliable and simplified system design. The passive core cooling system is employed A-35 Adequacy of Offsite Power Systems ! for both hot-standby and long-term cooling modes. Hot-standby conditions are achieved immediately and a Discussmn temperature of 420'F is reached within 36 hours. Generic issue A-35 addresses the susceptibility of ; l Reactor pressure is controlled and can be reduced to safety-related electric equipment to offsite power source I about 250 psig. The passive residual heat removal degradation. The NRC considers this issue as technical-system provides a closed cooling system to maintain ly resolved with the issuance of the Standard Review long-term core cooling. His capability reduces the Plan, Section 8.3.1 criteria specified in Appendix A, dependency on open loop-cooling systems, which have Branch Technical Position BTP PSB 1,
- Adequacy of limited the ability to remain in hot standby for long-term Station Electric Distribution System Voltages."
core cooling. Since the passive core cooling system maintains safe AP600 Response: i conditions indefmitely, cold shutdown is necessary only ne AP600 ac power system is discussed in Subsec- l to gain access to the reactor coolant system for inspec- tions 8.2.2.1 through 8.2.2.6 and Table 8.1-1. The l tion or repair. On the AP600, cold shutdown is accom- AP600 does not require any ac power source to achieve plished by using non-safety-related systems. These and maintain safe shutdown. systems are highly reliable. They have similar redun- , dancy as current generation safety-related systems and A-36 Control of Ileavy Loads Near Spent Fuel 4 are supplied with ac power from either onsite or offsite Discussion: 100.8-8 W
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NRC REQUEST FOR ADDITIONALINFORMATION I Generic issue A-36 addresses the need to review A-43 Containment Emergency Sump Performance requirements, facility designs, and Technical Specifi-cations regarding the movement of heavy loads near Discussion: spent fuel. The NRC has documented its technical Generic issue A-43 addresses technical concems as position on this issue in NUREG-0612 (Reference 10) follows: and that issued Standard Review Plan, Section 9.,1.5 which includes NUREG-0612 as a part of the review
- Pressurized water reactor sump (or boiling water plan, reactor residual heat removal system suction intake) hydraulic performance under post-loss-of-coolant AP600 Response: accident adverse conditions resulting from potential The AP600 design conforms to NUREG-0612 ed vedex formation, air ingestion, and subsequent Standard Review Plan, Section 9.1.5. Light load nan- pump failure dling systems are described in Subsection 9.1.4, and overhead heavy-load handling systems are descnbed in
- The possible transport of large quantities of insu-Subsection 9.1.5. lation debris generated by a loss-of-coolant accident resulting from a pipe break to the sump debris A-40 Seismic Design Criteria - Short Term Pro- screen (s), and the potential for sump screen (or gram suction strainer) blockage to reduce net positive suction head (NPSH) margin below that required Disculon: for the recirculation pumps to maintain long-term Generic issue A-40 addresses a desire to identify cooling and quantify conservatism in the seismic design process.
The Standard Review Plan, Section 3.7 provides clarifi-
- The capability of residual heat removal and con-cation of development of site-specific spectra, justifi- tainment spray system pumps to continue pumping cation for use of single synthetic time-history by power when subjected to possible air, debris, or other
! spectral density function, location and reductions of effects, such as particulate ingestion on pump seal input ground motion for soil-structure interaction, and and bearing systems design of above-ground vertical tanks. The revised provisions are used for margin studies and re-evaluations AP600 Response: l or individual plant examination for extemal events. Air ingestion, vortexing, and debris blockage are not significant concems for the AP600. The contain-AP600 Response: ment recirculation sumps include sump screens that The AP600 conforms to the criteria outlined in the conform to the criteria specified in Regulatory Guide Standard Review Plan, Section 3.7. The seismic design 1.82. The sump screens have a large cross-sectional criteria and seismic evaluation methodology are de- area to reduce the fluid flow velocity through the screen scribed in Section 3.7. and to provide a large screening area to accommodate l The AP600 employs generic, enveloping seismic accumulated debris. design criteria and applies established seismic evaluation Since the AP600 design does not use pumps to methodology that complies with current regulations and provide safety injection flow, the passive core cooling regulatory guidance. For sites having specific character- system injection flow rates are substantially lower than istics outside the range of the selected parameters, the those for plants with pumped injection flow. This AP600 is evaluated to demonstrate acceptability to the results in lower fluid flow velocities through the screens, site-specific characteristics. 100.8-9 W-Westinghouse
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i NRC REQUEST FOR ADDITIONALINFORMATION
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reducing the potential to draw debris into the sump required to place and maintain the plant in safe shut- ., l screens. down. ; The containment recirculation sump piping inlet is . i located slightly above the compartment floor, which is AP600 Response: (; substantially below the expected flood-up water level. AC electrical power is not needed to establish or This reduces the potential for air ingestion in the piping maintain a plant safe shutdown condition for the AP600. _3 since recirculation does not initiate until the flood-up The ac power system is discussed in Chapter 8. water level is well above the piping inlet. ; The elimination of pumps also eliminates concerns A-46 Seismic Qualification of Equipment in Oper- ' about the effects on -safety injection capability for . ating Plants ! vortexing, air ingestion, and blockage effects on pump net positive suction head. Discussion: j The AP600 includes the capability to use non- Generic Issue A-46 addresses the variability among .; safety-related normal residual heat removal pumps to operating plants in the margins of safety provided in . take a suction from the containment recirculation sump equipment to resist seismically induced loads and , to provide reactor coolant system injection. The sump perform the intended safety functions. The NRC i screen design addresses concerns with screen debris, believes that the seismic qualification of equipment in j vortexing, and air ingestion. . operating plants must, therefore, be reassessed to - l Section 6.3 provides additional information on the confirm the ability to bring the plant to a safe shutdown . -l operation of the passive core cooling system. Appendix condition when it is subject to a seismic event. I A describes conformance with Regulatory Guide 1.82. ! Section 6.2 provides additional information on the AP600 Response: ; coatainment recirculation sump. This issue applies to operating plants and, as such, , does not specifically apply to the AP600, which_ is l' designed in accordance with current seismic require-A-44 Station Blackout ments. The seismic Category I mechanical and electri . -! cal equipment utilized for the AP600 is qualified in Discussion: accordance with the AP600 qualification methodology is - l i Generic Issue A-44 was resolved with the publica- discussed in Section 3.10. The methodology is based on tion of 10 CFR 50.63, which provides requirements that ' the generic Westinghouse qualification program previ-light-water-cooled nuclear power plants be able to ously approved by the NRC. This methodology ad-- withstand for a specified duration and recover from a dresses IEEE Standard 344-1987 ( Reference 13) and . station blackout. It specifies that an alternate ac power Regulatory Guide 1.100. See Subsection 1.9.1 (Appen-- i source constitutes acceptable capability to withstand dix 1 A), station blackout provided an analysis is performed that 1 demonstrates that the plant has this capability from the A-47 Safety implications of Control Systans onset of the station blackout until the alternate ac .! source (s) and required shutdown equipment are started . Discussion: .
-l and lined up to operate. .
Generic Issue A-47 addresses the safety impact of 4 10 CFR .50.2 for the alternate ac source ' notes that non-safety-related control systems on plant dynamics. the alternate ac power source must have sufficient - Instrumentation and control systems used by nuclear _ i capability and reliability for operation of all systems- plants comprise safety-related protection systems and required for coping with station blackout for the time non-safety-related control systems. Safety-related - l l l 100.8-10 W-Westingh0use- - i
NRC REQUEST FOR ADDITIONAL INFORMATION NY $i.i
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systems are used to trip the reactor when specified Discussion: parameters exceed allowable limits and to protect the Generic Issue A-48 addresses postulated light water core from overheating by initiating emergency core reactor accidents resulting in a degraded or melted core cooling systems. Non-safety-related control systems are that could result in the generation and release to the used to maintain the plant within prescribed parameters containment of large quantities of hydrogen. One during shutdown, startup, normal h>ad, and varying source of hydrogen is from the reaction of the zirconium power operation. Non-safety-related systems are not fuel cladding with the steam at high temperatures. The relied on to perform any safety functions during or NRC requires design provisions for handling hydrogen following postulated accidents, but are used to control releases associated with rapid reaction of a large portion plant processes. of fuel cladding (10 CFR 50.44 and 10 CFR 50.34). AP600 Response: AP600 Response: For the AP600, control system failures are con- The AP600 design complies with the requirements sidered as potential initiating events. The analyses of of 10 CFR 50.44 and 10 CFR 50.34 (f). The mecha-these transients demonstrate that the consequences of nisms used to monitor and control hydrogen inside such failures are bounded by ANS Condition 11 criteria. containment are discussed in Subsection 6.2.4. No design basis failure of a control system is expected to violate Condition 11 criteria. A-49 Pressurized Thermal Shock The integrated control system for the AP600 obtains certain of its control input signals from signals used in Discussion: the integrated protection system. With the integrated Generic Issue A-49 addresses transients and acci-control and protection system, functional independence dents postulated to occur in pressurized water reactors of the control and protection systems is maintained by that can result in severe overcooling (thermal shock) of providing a signal selection device in the control system the reactor vessel, concurrent with high pressure. In for those signals used in the protection system. The these pressurized thermal shock events, rapid cooling of purpose of the signal selection device is to prevent a the reactor vessel internal surface causes a temperature failed signal caused by the failure of a protection distribution across the reactor vessel wall that produces channel from resulting in a control action that could lead a thermal stress with maximum tensile stress at the to a plant condition requiring that protective action. The inside surface of the vessel. The magnitude of the signal selection device provides this capability by thermal stress varies with the rate of change of tempera-comparing the redundant signals and automatically ture and is compounded by coincident pressure stresses. eliminating an aberrant signal from use in the control As long as the fracture resistance of the reactor system. This capability exists for bypassed sensors or vessel material is relatively high, these events are not for sensors whose signals diverge from the expected expected to cause vessel failure. The fracture resistance error tolerance. of the reactor vessel material decreases with the integrat-The plant control system incorporates design ed exposure to fast neutrons. The rate of decrease is features such as redundancy, automatic testing, and dependent en the chemical composition of the vessel self-diagnostics to prevent challenges to the protection wall and weld materials. and safety monitoring system. Chapter 7 provides a discussion of the AP600 instrumentation and controls. AP600 Response: The AP600 complies with the requirements of A-43 Ilydrogen Control Measures and Effects of 10 CFR 50.61. Material requirements and pressure-Ilydrogen Burns on Safety Equipment W-Westinghouse
NRC REQUEST FOR ADDITIONALINFORMATION E temperature limits are discussed in Subsections 5.3.2 For events where operator actions are taken, the and 5.3.3. AP600 design is based on previous experience and the guidance of ANSI 58.8-1984 (Reference 21). At least B-17 Criteria for Safety-Related Operator Actions 30 minutes is available following design basis events for the operator to initiate planned actions. Discussion: Generic Issue B-17 addresses the development of a B-22 LWR Fuel time criterion for safety-related operator actions includ-ing a determination of whether or not automatic actua- Discussion: tion is required. De evaluation of this issue includes Generic Issue B-22 addresses the reliability of fuel Issue 27, Manual versus Automated Actions, behavior predictions during normal operation and postulated accidents. Standard Review Plan, Section 4.2 AP600 Response: provides detailed NRC criteria for the design of fuel and The AP600 automatically initiates the safety-related core components, actions required to protect the plant during design basis events. The plant systems are designed to provide the AP600 Response: required information to the operator to monitor plant The AP600 reactor core design complies with the conditions and to evaluate the performance of the safety- Standard Review Plan, Section 4.2. See Section 4.2 for related passive systems, as well as the non-safety-related a discussion of the fuel system design. active systems. The active systems are designed to automatically actuate and provide defense-in-depth for B-29 Effectiveness of Ultimate IIcat Sinks various plant events, to preclude unnecessary actuation of the safety-related passive systems. The plant design Discussion: also provides the capability for a backup manual initia- Generic Issue B-29 addresses NRC confirmation of tion of both the passive and the active systems. currently used mathematical models for prediction of As described in Chapter 15, the AP600 safety ultimate heat sink performance by comparing model systems maintain the plant in a safe condition following performance with field data and deve'opment of better design basis events. For most of the design basis guidance regarding the criteria for weather record events, this is accomplished without the need for selection to define ultimate heat sink design basis operator action for up to 72 hours. Operator action is meteorology. planned and expected during plant events to achieve the The NRC considers this issue to be technically i I most effective plant response consistent with event resolved with the publication of three reports: conditions and equipment availability. For events where NUREG-0693, NUREG-0733, and NUREG-0858 operator action is taken, the plant design maximizes the (References 23,24 and 25). time available to complete actions for events. For example, during a steam generator tube rupture, no AP600 Response: ] operator action is required in order to establish safe The AP600 passive containment cooling system ; shutdown conditions or prevent steam generator overfill. complies with Standard Review Plan, Section 9.2.5 by j It is expected that the main control room operators take providing passive decay heat removal that transfers heat l actions similar to those taken in current plants to identify to the atmosphere, which is the ultimate heat sink for ' and isolate the faulted steam generator and to stabilize accident conditions. The passive containment cooling plant c<mditions. system is described in Subsection 6.2.2. I l l 100.8-12 l T Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION B-32 Ice Effects on Safety-Related Water Supplies stated in General Design Criterion 17, " Electric Power Systems,* must prototype-test the generator load circuit Discussion: breaker to demonstrate functional capability. Generic Issue B-32 addresses the potential effects of extreme cold weather and ice buildup on the reliability AP600 Response: of various plant water supplies. Current NRC criteria ne AP600 design incorporates a generator load are provided in Standard Review Plan, Section 2.4.7, circuit breaker to provide a reliable source of ac power " Ice Effects." to the electrical systems. Exceptions to General Design Criteria 17, as discussed in Section 3.1, are due to the AP600 Response: AP600 design not requiring ac power sources for a Subsection 6.2.2 describes the ultimate heat sink design basis accident. Subsection 8.2.2.5 provides design and discusses the features that prevent freezing of further discussion, the passive containment cooling waters. B-56 Diesel Reliability B-36 Develop Design, Testing, and Maintenance Criteria for Atmosphere Cleanup System Air Discussion: Filtration and Adsorption Units for Engi- Generic Safety Issue B-56 addresses the reliability neered Safety Features Systems and for Nor- of emergency onsite diesel-generators. Diesel reliability mal Ventilation Systems is a factor in the criteria associated with the resolution of Unresolved Safety Issue A-44. The resolution of Discussion: issue B-56 is the development of guidelines for an Generic Issue B-36 addresses the development of acceptable emergency diesel-generator reliability pro-revisions to current guidance and technical positions gram to ensure conformance with the emergency diesel-regarding engineered safety features and normal ventila- generator target reliability (0.95 to 0.975) identified in tion system air filtration and adsorption units. The N RC the proposed resolution of Unresolved Safety issue considers this issue technically resolved with the issu- A-44. ance of Revision 2 to Regulatory Guide 1.52 and Revision I to Regulatory Guide 1.140. AP600 Response: ,
%e AP600 diesel-generators are not safety related.
AP600 Response: He AP600 diesel-generator reliability is based on There are no safety-related air filtration systems in diesel-generator industry standards and practices. The l the AP600. Thu specific functions are outlined in diesel generator is discussed in Subsection 8.3.1. The Sectior 6.4 and Subsection 9.4.1. Conformance with diesel generator reliability is modeled in the PRA. The Repla'ory Guide 1.140 is discussed in Appendix 1 A. reliability program is discussed in Section 16.2. B-53 Load Break Switch B-61 Allowahle ECCS Equipment Outage Periods 1 Discussion: Discussion: Generic Issue B-53 addresses the use of the gen- Generic Safety Issue B-61 addresses surveillance erator load break switch for isolating the generator from test intervals and allowable equipment outage periods in the step-up transformer following turbine trip. Plant the technical specifications for safety-related systems. designs that utilize generator load circuit breakers to This task involves the NRC development of analytically satisfy the requirement for an immediate access circuit based criteria for use in confirming or modifying these
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i NRC REQUEST FOR ADDITIONAL INFORMATION
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surveillance intervals and allowable equipment outage releases of toxic and radioactive gases. The NRC periods. considers this issue as being technically resolved, and criteria have been incorporated in Standard Review AP600 Response: Plan, Section 6.4. The AP600 surveillance test intervals and allowable outage times help to meet plant safety goals while AP600 Response: maximizing plant availability and operability. In The AP600 main control room is essentially leak-determining these limits for the AP600 technical specifi- tight. A description of the control room habitability cations, a combination of deterministic, analytical, and systems is contained in Section 6.4. PRA-based evaluations is incorporated. Verification of design infiltration rates is as speci-fied in Standard Review Plan, Section 6.4. The AP600 B-63 Isolation of Low-Pressure Systems Connected minimizes unfiltered in-leakage by maintaining the main to the Reactor Coolant Pressure Boundary control room at a slightly positive pressure. Discussion: C-1 Assurance of Continuous Long-Term Capabil-Generic Issue B-63 addresses the adequacy of the ity of IIennetic Seals on Instrumentation and isolation of low-pressure systems that are connected to Electrical Equipment the reactor coolant pressure boundary. The NRC staff requires that valves formi ; he interface between high- Discussion: and low-pressure system. usociated with the reactor Generic issue C-l addresses the long-term capability coolant boundary have sufficient redundancy to prevent of hermetically sealed instruments and equipment that the low-pressure systems from being subjected to must function in post-accident environments. The NRC pressures that exceed their design limits. considers this issue as being technically resolved with the issuance of current criteria for qualification of AP600 Response: safety-related electrical equipment. The AP600 includes interconnections between high-and low-pressure systems. Each of these systems AP600 Response: interfaces contains appropriate isolation provisions. The AP600 environmental qualification program I Valves at the interface between high- and low-pressure described in response to Unresolved Safety issue A-24 systems have redundancy to prevent low-pressure addresses qualification of safety-related instrumentation I systems from being subjected to pressures that exceed and electrical equipment that must function under their design limits. The AP600 design meets the accident conditions. This program confirms the integri-provisions of the Standard Review Plan, Section 3.9.6. ty of seals employed in the design of Class IE equip- ! The normal residual heat removal system interface ment. See item A-24 of this subsection and Sec- ) is addressed in Subsection 5.4.7. tion 3.11 for AP600 qualification methodology. B-66 Control Room Infiltration Measurernents C-4 Statistical Methods for ECCS Analysis Discussion: Discussion: i Generic Safety issue B-66 addresses the adequacy Generic issue C-4 addresses NRC development of I I of control room area ventilation systems and control a statistical assessment of the certairty level of the peak building layout to enwre that plant operators are ade- clad temperature limit. Appendix K, *ECCS Evaluation quately protected against the effects of accidental Models," to 10 CFR 50 specifies the requirements for 100.8m W-Westinghouse P
NRC REQUEST FOR ADDITIONAL INFORMATION y ui( 1 ECCS analysis. These requirements call for conser- bined. An evaluation was made of the combined effect vatisms to be applied to cenain models and assumptions of power density, decay heat, stored energy, fission used in the analysis to account for data uncertainties at power decay, and their associated uncertainties with the time Appendix K was written. The resulting conser- regard to calculations of LOCA heat sources. vatism in the calculated peaic clad temperature (PCT) has not been thoroughly compared against the uncertain- AP600 Response: ty in peak clad temperature obtained from a realistically See Subsection 15.6.5 for a discussion of LOCA j calculated (best-estimate) LOCA. The staff allows heat sources. voluntary use of statistical uncertainty analysis tojustify relaxation of all but the required conservatisms con- C-10 Effective Operation of Containment Sprays in tained in current ECCS evaluation models. a LOCA l l AP600 Response: Discussion: l Chapter 15 discusses the methodology applied for Generic Issue C-10 addresses the effectiveness of LOCA analysis for the AP600. containment sprays to remove airborne radioactive materials that could be present within the containment l C-5 Decay IIcat Update following a LOCA. The NRC considers this issue as I being technically resolved with the issuance of ANSI Discussion: 56.5-1979 (Reference 28), which is referenced in Stan-Generic issue C-5 involves following the work . .f dard Review Plan, Section 6.5.2. research groups in determining best-estimate decay heat ! l data and associated uncertainties for use in LOCA AP600 Response: j calculations. The AP600 design does not employ a containment i 1 The staff has determined that the 1979 ANSI 5.1 is spray system for removal of airborne radioactive technically acceptable and has allowed the use of this materials in containment. Subsection 15.6.5.3 provides data tojustify relaxation of non-required conservatisms details of source term and mitigation techniques. in current ECCS evaluation models. The ECCS rule j change allows the use of this new data. His icsue was C-17 Interim Acceptance Critcria for Solidification i determined to be resolved. Agents for Radioactive Solid Wastes l AP600 Response: Discussion: The large-break LOCA analyses for the AP600, Generic Issue C-17 addresses the development of which employ the best-estimate W COBRA / TRAC criteria for acceptability of radwaste solidification analysis methodology (Subsection 15.6.5), use the decay agents. He NRC considers this issue as technically re-heat model identified in the 1979 ANSI 5.1 (Reference solved with the issuance of 10 CFR 61.56. 26). AP600 Response: C-6 LOCA Ileat Sources The AP600 solid radwaste system transfers, stores, and prepares spent ion exchange resins for disposal. It Dircussion: also provides for disposal of filter elements; soning, Generic Issue C-6 addresses the impact on LOCA shredding, and compaction of compressible dry active calculations of LOCA heat sources, their associated wastes; cleaning of protective clothing; and decontami-uncertainties, and the manner in which they are com- nation of small tools and components. The solid W-Westinghouse
i NRC REQUEST FOR ADDITIONAL INFORMATION A radwaste system does not provide fer liquid waste the impeller and rotor is contained within the pressure concentration or solidification. These functions, if used, boundary; therefore, seals are not required in order to are provided using mobile systems. Solidification of restrict leakage out of the pump into containment. AP600 wastes complies with the requirements in 10 Subsection 5.4.1 provides additionalinformation on the CFR 61.56. canned motor pump design for the AP600 reactor coolant pumps. Since the reactor coolant pumps do not 15 Radiation EfTects on Reactor Vessel Supports rely on seals as a reactor coolant pressure boundary, this issue is not applicable to the AP600. Discussion: Generic Safety issue 15 addresses the potential 29 Bolting Degradation or Failure in Nuclear problem of radiation embrittlement of reactor vessel Power Plants support structures. There is a potential for radiation embrittlement of the reactor vessel support stmeture Discussion: from long-term exposure to neutrons with an energy of Generic Safety Issue 29 addresses a concern about 1 MeV or greater. Embrittlement due to neutron pressure boundary integrity and component support darre may increase the potential for propagation of reliability associated with bolt failures. exstmg flaws. As documented in Generic Letter 91-17, the NRC has provided resolution of this issue. The resolution is AP600 Response: documented in NUREG-1339, " Resolution of Generic The supports for the AP600 reactor pressure vessel Safety Issue 29: Bolting Degradation or Failure in are designed for loading conditions and environmental Nuclear Power Plants," and NUREG-1445, " Regulatory factors including consideration of neutron fluence levels. Analysis for the Resolution of Generic Safety Issue 29: The material requirements include fracture toughness Bolting Degradation or Failure in Nuclear Power requirements and impact testing requirements in compli- Plants." The resolution was based on a number of ance with ASME Code, Section III, Subsection NF industry initiatives and NRC staff actions. NRC staff requirements. The reactor pressure vessel supports are actions include issuing a number of bolting-related not in the region of high neutron fluence where neutron bulletins, generic letters and information notices. embrittlement of the supports would be a significant Industry initiatives include the publishing of EPRI concern. Reports NP-5769,
- Degradation and Failure of Bolting in Nuclear Power Plants," and NP-5067, " Good Bolting 13 Reactor Coolant Pump Seal Failures Practices, A Reference Manual for Nuclear Power Plant Maintenance Personnel."
Discussion: EPRI . Report NP-5769 establishes the characteristic Generie Safety Issue 23 addresses reactor coolant that bolted connections exhibit leakage prior to failure pump seal failures that challenge the makeup capacity in resulting from bolt degradation. The NRC has endorsed PWRs. Such seal failures represent small-break loss-of- the recommendation in NP-5769 that plant-specific coolant accidents. bolting integrity programs be established that encompass safety-related boltmg. NUREG-1339 includes recom-AP600 Response: mendations and guidelines for the content of a compre-The AP600 reactor coolant pumps are canned motor hensive bolting integrity program. pumps. A canned motor pump contains the motor and all rotating components inside a pressure vessel designed AP600 Response: for full reactor coolant system pressure. The shaft for 100.8-16 W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION HF u 4 E i The elements of resolution pertain mainly to opera- and the turbine building closed cooling water system tional and maintenance practices instead of hardware heat exchangers and has no safety-related functions. requirements. These practices are addressed by the None of the safety-related equipment requires cooling maintenance program of the combined license holder. water to effect a safe shutdown or mitigate the effects of Conformance to ASME Code, Section III requirements design basis events. Heat transfer to the ultimate heat for pressure boundary components and related supports sink is accomplished by heat transfer through the provides safe operation in the event of bolting degrada- containment shell to air and water flowing on the outside tion. Because of the emphasis in the AP600 design on of the shell. axess for maintenance and inspection, the recommended The design of the service water system and the maintenance practices can be readily implemented. provisions for minimizing long-tern corrosion and organic fouling are described in Subsection 9.2.1. 45 Inoperability of Instrumentation Due to 57 Effects of Fire Protection System Actuation Extreme Cold Weather Discussion: Discussion: Generic Safety Issue 57 addresses the potential for Generic Safety Issue 45 addresses the inoperability adverse interactions from actuation of the fire protection of instrumentation due to extreme cold weather. This system with safety-related equipment. Operating issue was resolved with the issuance of changes to experience has shown that safety-related equipment Standard Review Plan, Section 7.1, Appendix A to subject to fire protection system water spray and other Section 7.1, Section 7.5, and Section 7.7. suppressant chemicals can be rendered inoperable. AP600 Response: AP600 Response: The AP600 complies with Standard Review Plan The fire protection system and fire protection Section 7.1 Appendix A to Section 7.1, Section 7.5, program in the AP600 minimize the potential for and Section 7.7. adverse interactions of safety-related equipment with the fire protection system. He means used to achieve this 51 Proposed Requiranents for Improving the result include: isolating combustible material and Reliability of Open Cycle Service Water limiting the spread of fire by subdividing the plant into Systems fire areas separated by fire barriers, providing separate and redundant safe shut down components and associat-Discussion: ed electrical divisions to preserve the ability to safely Generic Safety Issue 51 addresses the susceptibility shutdown the plant following a fire, and providing floor of open cycle service water systems to fouling including drains sized to remove expected firefighting water the buildup of aquatic bivalves and corrosion products without flooding safety-related equipment. The design that can significantly degrade the performance of the of the fire protection system is described in Subsection system. In operating plants, the service water system is 9.5.1. typically used to cool safety-related equipment and to transfer decay heat to the ultimate heat sink. 79 Unanalynd Reactor Vessel Thennal Stress During Natural Convection Cooldown AP600 Response The service water system in the AP600 provides Discussion: cooling water to the component cooling water system W Westinght se
NRC REQUEST FOR ADDITIONAL INFORMATION Generic Safety Issue 79 addresses the thermal Safety-related valves must meet the requirements of stresses that occur in the reactor vessel head flange ASME Code, Section Ill to provide pressure boundary during a natural circulation cooldown. High stresses in integrity. Valves and valve operators are sized to the flange or studs during a natural circulation cooldown provide operation under a full range of design basis flow in PWRs could violate ASME code allowables. Cycling and pressure drop conditions. For the AP600, motor-of the stresses could reduce the fatigue margin. operated valve designs are subject to qualification testing to demonstrate the capability of the valve to open, close, AP600 Response: and seat against maximum pressure differential and The natural circulation cooldown transient is flow. The requirements for this testing are based on evaluated as part of ASME Code vessel evaluations and ANSI B16.41,
- Functional Qualification Requirements is discussed in Subsection 3.9.1.1.2.11. for Power Operated Active Valve Assemblies for Nuclear Power Plants." See Subsection 5.4.8 for an 87 Failure of IIPCI Steam Line Without Isola- outline of AP600 valve requirements.
tion The inservice testing program for safety-related valves is the responsibility of the combined license Discuwion: applicant. The design of the systems in the AP600 Generic Safety issue 87 addresses the uncertainty permits testing in accordance with the guidance in regarding the operability of the motor-operated isolation Generic Letter 89-10 and the position on inservice valve valves for the steam supply lines of the high-pressure testing in SECY-93-087. Where practical, motor-coolant injection (HPCI) system in boiling water reac- operated and check valves are to be tested under full tors following a postulated break in the supply line. A flow under actual plant conditions. The valves built to break in the line could lead to high flow or high ASME Code, Section III may be tested in compliance differential pressure that may inhibit closure of the with the requirements found in the ASME standard, isolation valve. These valves typically cannot be tested
- Operation and Maintenance of Nuclear Power Plants."
in-situ for the design flow rates and pressures. Al- For additional information on inservice testing of safety-though the AP600 does not have a high-pressure coolant related valves, see Subsection 3.9.6. injection system, it does have isolation valves designed to close against high flow or high pressure differential 93 Steam Binding of Auxiliary Feedwater Pumps in the event of a postulated pipe break. The issue of the operability of motor-operated Discussion valves has received considerable attention since Generic Generic Safety Issue 93 addresses the potential for Safety Issue 87 was initiated. The NRC provided a common mode failure of the pumps in an auxiliary or guidance for inservice testing of motor-operated, safety- emergency feedwater system. Ilot water leaking related valves in Generic Letter 89-10. SECY-93-087 through one or more isolation valves can flash to steam identifies the proposed position on inservice testing of at the auxiliary feedwater pump potentially resulting in safety-related valves for advance light water reactors. the failure of the pump to operate if required because of The guidance in these documents recommends that steam binding. The NRC addressed this issue in safety-related valves be tested under full flow under Bulletin 85-01,and reinforced it in Generic letter 88-03, actual plant c(mditions where practical. EPRI has a by requesting that the fluid conditions in the auxiliary program to demonstrate operation of motor-operated feedwater system be monitored and procedures be valves. developed to recognize steam binding and restore the auxiliary feedwater system to operable status if steam AP600 Response: binding should occur. 100.8-18 W-Westinghotise
NRC REQUEST FOR ADDITIONAL INFORMATION : A His IF
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AP600 Response: tance to embrittlement. See Subsection 5.3.2 for The AP600 does not have a safety-related auxiliary additional information on the requirements to address j feedwater system. The passive core cooling system fracture toughness of the reactor vessel. ! provides the safety-related function of cooling the The normal residual heat rernoval system is de-reactor coolant system in the event ofloss of feedwater. signed to provide the safety-related function of low l l The startup feedwater system provides the steam genera- temperature overpressure protection for the reactor tors with feedwater during plant conditions of startup, coolant system during refueling, startup, and shutdown hot standby, and cooldown and when the main feedwater operations. The system is designed to limit the reactor pumps are unavailable. The startup feedwater system coolant system pressure within the limits specified in 10 has no safety-related function other than containment CFR 50, Appendix G. The relief valve in the normal isolation. residual head removal system is used to provide the The startup feedwater system does include tem- overpressure protection. See Subsection 5.4.7 for , perature instrumentation in the pump discharge that additional information on the design of the normal I I would permit monitoring of the temperature of the residual heat removal system and the overpressure startup feedwater system. protection function. 94 Additional Low-Temperature Overpressure 103 Design for Probable Maximum Precipitation Protection for Light Water Reactors l Discussion: Discussion: Generic Safety Issue 103 addresses the methodology Generic Safety Issue 94 addresses the establishment used for determining the design flood level for a particu. of additional guidance for reactor coolant system low- lar reactor site. This issue was resolved by incorporat-temperature overpressure protection to ensure reactor ing the methodology into the Standard Review Plan. vessel and reactor coolant system integrity beyond that identified in the resolution to Generic Safety Issue (GSI) AP600 Response: A-26. Low-pressure overpressurization events that This is a site-related parameter. The AP600 is occurred subsequent to the implementation of the designed for air temperatures, humidity, precipitation, guidelines for resolution of GSI A-26 indicated a need snow, wind, and tornado conditions as specified in for additional low-temperature overpressure protection. Table 2.0-1. The combined license applicant will To resolve this issue, the NRC issued Generic Letter 90- demonstrate that the site parameters are within the limits 016 which required a revision to plant technical specifi- specified for the standard design. cations for operability of the low-temperature over- The site is acceptable if the site characteristics fall pressure protection system. Other possible solutions within the AP600 plant site design parameters in Ta-identified in GL 90-016 included hardware modifications ble 2.0-1. For cases where a site characteristic exceeds including use of residual heat removal system relief the envelope parameter, it will be necessary for the valves and requiring the low temperature overpressure combined license applicant referencing the AP600 to protection system to be fully safety related. demonstrate that the site characteristic does not exceed the capability of the design. For additionalinformation AP600 Response: on the site interface parameters, see Chapter 2. The reactor vessel for the AP600 is designed to be less susceptible to brittle fracture during low tempera- 105 Interfacing System LOCA at BWRs ture overpressure events. The material requirements and welding processes are developed to enhance resis- Discussion: W Westinghouse
NRC REQUEST FOR ADDITIONALINFORMATION Generic Safety Issue 105 addresses concerns over areas of the plant that are removed from the Nuclear l the adequacy of isolation valves between the RCS and Island (see Subsection 9.3.2 for a description of the low-pressure interfacing systems in BWRs. This issue, plant gas system). The exception to this is the hydrogen which is limited to pressure isolation valves in BWRs, supply line to the chemical and volume control system is related to Generic Safety Issue 96. which considers (CVS). the failure of the pressure isolation valves between the ne CVS is the only system on the 1,uclear island RCS and the RHR system in PWRs. Overpressurization that uses hydrogen gas. Hydrogen is supplied to the of low-pressure piping systems due to reactor coolant AP600 CVS inside containment from a single 550-scf system boundary isolation failure could result in rupture hydrogen bottle located in the plant gas storage area. of the low-pressure piping outside containment. This The release of the contents of an entire bottle of hydro-may result in a core melt accident with an energetic gen in the most limiting building volumes (both inside release outside the containment building that cou'd cause containment and in the auxiliary building) would not a significant offsite radiation release. Designing inter- result in a volume ps cent of hydrogen large enough to facing systems to withstand full reactor pressure is an reach a detonable leve. i acceptable means of resolving this issue. He CVS hydrogen supply piping is routed from the l remote plant gases storage tank area through the turbine AP600 Response: building and into the auxiliary building and then into For information on this issue, see Subsection containment. The H2supply line is routed through the 1.9.1.5, SECY-90-016 Issues. See Subsection 5.4.7 for piping / valve room on elevation 100'-O' of the auxiliary additional information on the normal residual heat building. The piping / valve penetration room in the removal system design. auxiliary building on elevation 100'-0" is designed as a 3-hour fire zone. A fire in this area would not inhibit 106 Piping and Use of Ilighly Combustible Gases the safe shutdown of the plant. More information is in Vital Areas contained in Appendix 9A, Subsections 9.A.3.1.1.5 and 9A.3.1.2.9.7. Discussion: The turbine building does not house any safety-Generic Issue 106 addresses the normal process related systems or equip. ment. The release of hydrogen system use of relatively small amounts of combustible from the CVS hydrogen supply line into an area of the ! I gases on site and also addresses leaks or breaks in the turbine building does not represent a threat to the safety hydrogen piping and supply system that could result in of the plant. the accumulation of a combustible or an explosive The AP600 containment has hydrogen sensors that l mixture of air and hydrogen within the auxiliary systems would detect hydrogen leaks. The containment hydro-building. The accumulation of combustible or explosive gen concentration monitoring subsystem is designed as ! mixtures of gas in the auxiliary systems building could Class lE and seismic Category I. Here are two represent a threat to safety-related equipment if the independent trains, each consisting of eight hydrogen combustible gases are inadvertently ignited. sensors in various locations throughout the containment free volume. Subsection 6.2.4.1 describes this subsys-Response: tem. The AP600 uses small amounts of combustible gases for normal plant operation. Most of these Fases 113 Dynamic Qualification Testing of Large-Bore are used in limited quantities and are associated with Ilydraulic Snubben plant functions or activities that do not jeopardize any safety-related equipment. These Fases are found in Discussion: 100.8-20 W westinghouse
NRC REQUEST FOR ADDITIONALINFORMATION Generic Safety Issue 113 addresses the requirements The concerns are prediction of conditions in realistic for qualification and periodic operability testing of large configurations, and containment and equipment surviv-bore hydraulic snubber for operating plants. Large-bore ability. hydraulic snubbers are used to a limited extent on the AP600 to provide support, particularly for seismic Response: events, of piping systems and components while allow- The AP600 includes provisions for hydrogen control ing for movement due to thermal expansion. The NRC, for the unlikely severe accident cases in which large in a draft regulatory guide (SC-708-4, " Qualification and amounts of hydrogen could be generated because of Acceptance Test for Snubbers Used in Systems Impor- degraded core events. Analyses were performed to tant to Safety *), has established recommendations for examine the consequences of hydrogen burn and to testing of hydraulic snubbers on a forward-fit basis, that evaluate the likelihood of deflagration to detonable is, units without a license at the time the recommenda- transitions. tions were established. For severe accident cases, the containment hydrogen control system prevents hydrogen burn initiation at high AP600 Position: hydrogen concentration levels. Ilydrogen igniters Since this issue applies to operating plants, it does promote burning when the lower flammability limit is not directly apply to the AP600. The AP600 plant uses reached and limits the containment hydrogen concentra-significantly fewer hydraulic snubbers than do currently tion to less than 10 volume percent during and following operating plants. In addition to the recommendations in a degraded core or core melt. the draft regulatory guide, testing requirements have Thus, for severe accident cases, the AP600 is been established in ASME OM Code - 1990, ' Code for designed to prevent the occurrence of hydrogen deto-Operation and Maintenance of Nuclear Power Plants.* nation, thereby preventing the possibility of the resultant Qualification and periodic operability testing require- large pressure spikes in containment, which is the source ments for the few snubbers in the AP600 will be of concern for containment integrity and equipment established based on the requirements in ASME OM survival. Details of the hydrogen ignition subsystem are Code, Subsection ISTD and the recommendations in the provided in Subsection 6.2.4.2.3. Placement of the draft regulatory guide. The design of the hydraulic hydrogen igniters is discussed in Chapter 16 of the PRA snubbers permits required preoperational and inservice evaluation report. testing. When preservice operabihty testing is per- A hydrogen burn analysis shows that the AP600 formed in the numufacturer's facility, the testing is done hydrogen igniter system is effective in maintaining the in compliance with the requirements of ASME OM hydrogen concentration throughout the containment close Code, Subsection ISTD. to the lower flammability limit, and that the peak pressure in the containment during and following l 121 Ilydrogen Control for Lnrge, Dry PWR Con- hydrogen burn remains well below ASME service level tainments C stress intensity limits. The hydrogen concentration is similar in all compartments analyzed, indicating that the Discussion: hydrogen released mixes wellin the AP600 containment. Generic Safety Issue 121 concerns ongoing NRC The analyses predict conditions in realistic configura-experimental and analytical programs addressing the tions. Peak gas temperatures and pressures in each likelihood of safe shutdown equipment surviving a compartment for each case analyzed are provided, thus l hydrogen burn. The staff also intends to explore the providing the hydrogen burn thermal environment that ; possibility and probable consequences of the formation containment equipment will experience. Details are of local detonable concentrations in large, dry PWRs. provided in Chapter 14 of the PRA report. 100.8-21 W-Westinghouse I
NRC REQUEST FOR ADDITIONAL INFORMATION Ha Hii E 7 The challenge to the AP600 containment integrity safety-related function of cooling of the reactor coolant , from hydrogen deflagrations and detonations during core system in the evert of loss of feedwater. The startup l damage events is examined in the hydrogen deflagration feedwater system provides the steam generators with and detonation analyses. This bounding evaluation feedwater during plant conditions of startup, hot stand-assumes that an amount of hydrogen equivalent to 100- by, and cooldown and when the main feedwater pumps percent active cladding oxidation burns all at once in the are unavailable. The startup feedwater system has no AP600 containment, with no credit taken for the hydro- safety-related function. gen igniters. The evaluation concludes that a hydrogen deflagration is unlikely to cause containment failure. 128 Electrical Power Reliability Other analyses show that the likelihood of a deflagration to detonation transition in any part of the AP600 con- Discussion: tainment is unlikely. Containment failure from a Generic Safety issue 128 addresses the reliability of detonation is not considered a credible event for the onsite electrical systems and encompasses GSI 48, GSI AP600 because of the lack of conditions supporting a 49, and GSI A-30. deflagration to detonation transition, the provision and placement of hydrogen igniters, and the containment AP600 Position: design features resulting in a well-mixed atmosphere. The design basis and design criteria for the Class Details are provided in Subsection 10.2.5 of the PRA lE de and UPS system is provided in Subsections evaluation report. 8.1.4.2.1 and 8.1.4.3. The class IE de and UPS system The hydrogen igniters and the containment electrical design is described in Subsection 8.3.2.1.1. Specifical-and mechanical penetrations are designed to operate in ly, this design addresses IEEE Standards 603 and 308. the most limiting severe accident environment, including This includes the following generic issues: a hydrogen burn. (See Subsection 10.2.5 of the PRA evaluation report.) The approach of using controlled
- Generic Safety Issue 48, LCO for Class IE vital burning to prevent accidental hydrogen burn initiation instrument buses in operating reactors. Chapter 16 provides confidence that safety-related equipment will provides the AP600 technical specifications.
continue to operate during and after hydrogen burns. Subsections 16.1.3.8.3 and 16.1.3.8.4 provide the (See Subsection 6.2.4) limiting conditions for operation in the event of a loss of one or more Class IE 120-vac vital instru-124 Auxiliary Feedwater System Reliability ment buses and the associated inverters. Discussion:
- Generic Safety issue 49, interlocks and LCOs for Generic Safety issue 124 addresses the use of Class IE tie breakers. Based on the historical probabilistic risk assessment to evaluate the reliability of background, this issue is not applicable to the the auxiliary fwdwater system. The issue was resolved AP600 design. There is no 4160V Class IE tie by the NRC's issuing plant-specific requirements for a breaker or de tie breakers between the four class 1E few plants that did not initially have a reliability higher divisions.
than a minimum criteria.
- Generic Safety issue A-30, adequacy of safety-AP600 Response: related de power supplies. The AP600 incorporates This issue is not applicable to the AP600. The the following recommended enhancements:
AP600 does not have a safety-related auxiliary feedwater system. The passive core cooling system provides the 100.8-22 WBStingh0USS
l NRC REQUEST FOR ADDITIONAL INFORMATION li!"
.It:
Nil
*41 e
- The Class IE de distribution system design is in 2. Reviewing SGTR results and conclusions to develop accordance with the guidelines ofIEEE Standard regulatory analysis supporting Standard Review 384 and Regulatory Guide 1.75. Plan changes.
- 3. Reassessing SGTR associated issues including radio-
- Four Class 1E de power supplies into four logical, design basis, tube integrity, procedures, and separate electrical divisions. RCS pressure control.
- 4. Reviewing the effects of water hammer, overfill and The AP600 design provides additional testing capability water carryover.
through the installed spare battery bank with one in- The results of the tasks will provide the staff with a stalled battery charger. The spare battery bank permits basis to develop a position on offsite dose, operator frequent full-component testing without compromising action, tube integrity, water hammer, and valve opera-plant availability. Battery equalization can be performed bility, off-line. The battery and battery charger can be tested and maintained separately. AP600 Response: The AP600 design features are discussed below. 130 Essential Service Water Pump Failure at Multiple Plant Sites TASK 1: Appendix 1A identifies the level of confor-mance with Regulatory Guide 1.83, " Inservice Inspec-Discussion: tion of Pressurized Water Reactor Steam Generator Generic Safety issue 130 addresses the use of Tubes." As detailed in Appendix 1 A, the AP600 design shared or cross-connected essential service water essentially conforms with the regulatory guidance except systems at sites with two or more reactor plants. where state-of-the-art advances have enhanced inservice During some situations the crosstied pumps may not be inspection techniques. Further, as specified in Subsec-available for accident mitigation operations. tion 5.4.2.5, the steam generators permit access to tubes for inspection and/or repair or plugging, if necessary, AP600 Response: per the guidelines described in Regulatory Guide 1.83. The AP600 is a single independent plant that does The AP600 steam generator includes features to enhance not share or cross-tie systems or components with robotics inspection of steam generator tubes without another plant. This issue a not applicable to the manned entry of the channel head. AP600. TASYs 2: Subsection 15.6.3.1.4 discusses anticipated 135 Integrated Steam Generator issues operator recovery actions and the effects of those actions in the mitigation of a steam generator tube rupture Discussion: (SGTR). As discussed in Subsection 15.6.3.2, the Generic Safety Issue 135 was initiated in order to AP600 incorporates automatic steam generator overfill provide an integrated wo k plan for the resolution of protection. The details of the design are provided in steam generator issues including steam generator overfill Subsection 15,6.3.2, with the control logic provided in consequences, water hammer, and eddy current testing. proprietary Section 7.2. The issue was divided into the following four tasks: TASK 3: A compilation of a number of generic con-
- 1. Assessing current capabilities of eddy current cerns is most appropriately addressed by providing a testing and developing recommendations. reference to which sections of the SSAR provide perti-nent details.
W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION i
- Reassessment of radiological consequences: Sub- program, replace failed or unacceptable isolators, and section 15.6.3 provides details of the scenario, implement an annuri program to inspect and test all analysis assumptions, and results. electronic isolators between Class IE and non-Class IE
= Re-evaluation of design basis SGTR: The design systems, basis SGTR evaluated on the AP600 design is discussed in Subsection 15.6.3, providing details of AP600 Response: the scenario, analysis assumptions and results. The use of isolation devices in the AP600 Instru-
- Supplemental Tube Inspections: See Subsection mentation and Control Architecture is described in 5.4.2.5, Appendix 1 A, Regulatory Guide 1.83. Subsections 7.1.2.11, " Isolation Devices," 7.1.4.2.7,
- Denting criteria: Subsection 5.4.2.4.3 provides a "Conformance to the Requirements Concerning Control discussion of steam generator design and tubing and Protection System Interaction (Paragraph 4.7 of compatibility with secondary coolants. IEEE 279-1971, GDC-24)," and 7.7.1.11, " Diverse
- Improved accident monitoring and reactor vessel Actuation System." As stated in Subsection 7.1.4.2.,7 inventory measurement: Section 7.5 discusses the the isolation devices are tested to conform to require-safety related display information. ments. This testing meets the requirement for an
- Reactor coolant pump trip: Subsection 7.3.1.1.3.3 inspection and test program and identifies those devices discusses reactor coolant pump trip. that are potentially susceptible to electrical leakage.
- Control room design: Sections 7.5 and 18.9 discus Implementation of an armual program to inspect and test the control room design and design process. all electronic isolators between Class IE and non-Class
- Emergency operating procedures: Subsection 18.9.8 IE systems is the responsibility of the combined license addresses the development of emergency operating holder, procedures.
- Organizational responses: Section 18.12 identifies IIF4.1 Inspection Procedures for Upgraded Emer-that the organizational responses are a part of the gency Operating Procedures combined license application.
- Reactor coolant pressure control: Subsection 7.7.1.6 Discussion:
addresses primary system pressure control. Iluman Factor Issue 4.1 involves the development of criteria to provide assurance that plant procedures are TASK 4: Steam generator overfill, water carryover and adequate and can be used effectively. Criteria to water hammer are addressed as previously identified and evaluate and inspect EOPs by the regions have been are discussed in Subsection 15.6.3.2, with the control prepared by NRR and OIE and were published as an logic provided in proprietary Section 7.2. OIE Temporary Instmetion. Similar criteria and { inspection modules will be developed when the guide- ] 142 Leakage Through Electrical Isolators in lines for upgrading of other procedures are completed. ; Instrumentation Circuits AP600 Response: Discussion: The design of the EOPs is consistent with NUREG-Generic Issue 142 addresses the susceptibility to 1358 and its supplement, as well as with other current i leakage of isolation devices between safety- and non- regulatory guidance and standards. See Subsection , I safety-related electrical systems. The NRC requires that 18.9.8 for additional information. licensees identify isolation devices in instrumentation circuits that are potentially susceptible to electrical leakage, define and perform an inspection and test IlFS.1 Local Control Stations I 100.8-24 W westinghouse 1 l
NRC REQUEST FOR ADDITIONALINFORMATION Discussion: Human Factors issue 5.1 addresses the need to develop additional guidance for the design of local control stations. AP600 Response: Westinghouse employs the techniques and experi-ence gained in the design of the main control room and remote shutdown panel to the local control station designs. The methodology used to analyze the job / tasks of the control room crew is applied to the job / tasks of auxiliary personnel to identify and describe communica-tion and control links between the control room and the auxiliary control stations. IIF5.2 Review Criteria for lluman Factors Aspects of Advanced Controls and Instrumentation Discussion: Human Factors issue 5.2 addresses review criteria for human factors aspects of advanced controls and instrumentation. AP600 Response: The AP600 design is in conformance with current guidance and requirements relative to integrated human factors design. A description of the Westinghouse advanced alarm system is described in Subsection 18,9.2. A description of the computerized procedures can be found in Subsection 18.9.8.6. A detailed de-scription of the qualified display processing system can be found in Subsection 18.9.5. The plan for verification and validation of the AP600 M-MIS is described in detail in Subsection 18.P.2.3. 00.8-25 W Westinghouse l i
NRC REQUEST FOR ADDITIONAL INFORMATION
!!!!! E Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria TMI Action Plan Items I.A.1.1 Shift Technical Advisor f I.A.l.2 Shift Supervisor Administrative Duties f I.A.I.3 Shift Manning f 1.A.I.4 long-Term Upgrading c
- 1. A.2.1(1) Qualifications - Experience f I. A.2.1(2) Training f I. A.2.1(3) Facility Certification of Competence and Fitness of Appli- f cants for Operator and Senior Operator Licenses 1.A.2.2 Training and Qualifications of Operations Personnel e I.A.2.3 Administration of Training Programs f 1.A.2.4 NRR Participation in Inspector Training d I. A.2.5 Plant Drills c 1,A.2.6(1) Revise Regulatory Guide 1.8 f I. A.2.6(2) Staff Review of NRR 80-117 c I.A.2.6(3) Revise 10 CFR 55 e I.A.2.6(4) Operator Workshops c
!. A.2.6(5) Develop Inspection Procedures for Training Programs c I
- 1. A.2.6(6) Nuclear Power Fundamentals a 1 1
1.A.2.7 Accreditation of Training Institutions e I.A.3.1 Revise Scope of Cnteria for Licensing Examinations f 100.8-26
NRC REQUEST FOR ADDITIONAL INFORMATION Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable Itern/ Issue Title Screening Notes No. Criteda I. A.3.2 Operator Licensing Program Changes c 1.A.3.3 Requirements for Operator Fitness c 1.A.3.4 Licensing of Additional Operations Personnel c 1.A.3.5 Establish Statement of Understanding with INPO and DOE d I.A.4.l(l) Short-Term Study of Training Simulators c I. A.4. l(2) Interim Changes in Training Simulators f I.A.4.2(1) Research on Training Simulators f I.A.4.2(2) Upgrade Training Simulator Standards f 1.A.4.2(3) Regulatory Guide on Training Simulators f I.A.4.2(4) Review Simulators for Conformance to Criteria c 1.A.4.3 Feasibility Study of Procurement of NRC Training Simulator d I.A.4.4 Feasibility Study of NRC Engineering Computer d I.B.I.l(l) Prepare Draft Criteria c I.B.I.1(2) Prepare Commission Paper c 1.B.I.l(3) Issue Requirements for the Upgrading of Management and c Technical Resources ! I.B.I.l(4) Review Responses to Determine Acceptability c ; I.B.I.l(5) Review Implementation of the Upgrading Activities c I I.B.I.l(6) Prepare Revisions to Regulatory Guides 1.33 and 1.8 e 1.B.I.l(7) Issue Regulatory Guides 1.33 and 1.8 e I.B.I.2(1) Prepare Draft Criteria c 10 .8-27 l W Westinghouse
m - - __ . ._ _ . _ _ . . . t NRC REQUEST FOR ADDITIONAL INFORMATION -_ t
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Table 1.9-2 ! Listing of Unresolved Safety Issues and Generic Safety issues , Action Plan Applicable Itan/ Issue Title Screening Notes No. Criteria I.B.I.2(2) Review Near-Term Operating License Facilities c I.B.I.2(3) Include Findings in the SER for Each Near-Term Operating c License Facility ., 1.B.I.3(1) Require Licensees to Place Plant in Safest Shutdown Cooling d i Following a Loss of Safety Function Due to Personnel Error .j 1.B.I.3(2) Use Existing Enforcement Options to Accomplish Safest d 6 Shutdown Cooling . 1.B.I.3(3) Use Non-Fiscal Approaches to Accomplish Safest Shutdown d l Cooling ! I.B.2.l(l) Verify the Adequacy of Management and Procedural Con- d l trols and Staff Discipline f i I.B.2.1(2) Verify that Systems Required to Be Operable Are Properly d j Aligned ;. I.B.2.l(3) Follow-up on Completed Maintenance Work Orders to En- d ! sure Proper Testing and Return to Service I.B.2.1(4) Observe Surveillance Tests to Determine Whether Test In- d ; struments Are Properly Calibrated j I.B.2.l(5) Verify that Licensees Are Complying with Technical Specifi- d cations I.B.2.l(6) Observe Routine Maintenance d I.B.2.l(7) Inspect Terminal Boards, Panels, and Instrument Racks for d Unauthorized Jumpers and Bypasses ; I.B.2.2 Resident Inspector at Operating Reactors d I.B.2.3 Regional Evaluations d l.B.2.4 Overview of Licensee Performance d Y 100.8-28
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NRC REQUEST FOR ADDITIONAL INFORMATION j.11 Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable item / Issue Title Screening Notes No. Criteria I.C.l(l) Small Break LOCAs f I.C.l(2) Inadequate Core Cooling f I.C.l(3) Transients and Accidents f I.C.l(4) Confirmatory Analyses of Selected Transients e I.C.2 Shift and Relief Turnover Pmcedures f 1.C.3 Shift Supervisor Responsibilities f I.C.4 Control Room Access f 1.C.5 Procedures for Feedback of Operating Experience to Plant g See SSAR Subsection Staff 1.9.3, item (3)(i). 1.C.6 Procedures for Verification of Correct Performance of Oper- f ating Activities 1.C.7 NSSS Vendor Review of Procedures f 1.C.8 Pilot Monitoring of Selected Emergency Procedures for f l Near-Term Operating License Applicants j 1.C.9 Long-Term Program Plan for Upgrading of Procedures c
- 1. D.1 Control Room Design Reviews g See SSAR Subsection 1.9.3, item (2)(iii).
1.D.2 Plant Safety Parameter Display Console g See SSAR Subsection 1.9.3, item (2)(iv).
- 1. D.3 Safety System Status Monitoring h (Medium) See SSAR Subsection 1.9.3, item (2)(v).
1.D.4 Control Room Design Standard c ) 1.D.5(1) Operator-Process Communication c
' ~ ~*
W wesungnouse
NRC REQUEST FOR ADDITIONALINFORMATION mi tiH I!! .l! Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria 1.D.5(2) Plant Status and Post-Accident Monitoring g See SSAR Subsection 1.9.4, item 1.D.5(2). I.D.5(3) On-Line Reactor Surveillance System h See SSAR Subsection 1.9.4, item 1.D.5(3). 1.D.5(4) Process Monitoring Instrumentation c I.D.5(5) Disturbance Analysis Systems d I.D.6 Technology Transfer Conference d I.E.1 Office for Analysis and Evaluation of Operational Data d I.E.2 Program Office Operational Data Evaluation d I.E.3 Operational Safety Data Analysis d 1.E.4 Coordination of Liceasee. Industry, and Regulatory Programs d I.E.5 Nuclear Plant Reliability Data Systena d 1.E.6 Reporting Requirements d 1.E.7 Foreign Sources d 1.E.8 iluman Error Rate Analysis d I.F.1 Expand QA List c 1.F.2(1) Assure the Independence of the Organization Performing the a Checking Function 1.F.2(2) Include QA Personnel in Review and Approval of Plant g See SSAR Subsection Procedures 1.9.3, item (3)(iii). 1.F.2(3) Include QA Personnel in All Design, Construction, Installa- g See SSAR Subsection tion, Testing, and Operation Activities 1.9.3, item (3)(iii). 100.8-30 [ WB5tirigh0l!$8
i i l NRC REQUEST FOR ADDITIONAL.INFORMATION I
- =
15 : l Table 1.9-2 3 Listing of Unresolved Safety Issues and Generic Safety Issues : Action Plan Applicable f Item / Issue Title Screening Notes ) No. Criteria ; l 1.F.2(4) Establish Criteria for Determining QA Requirements for a , Specific Classes of Equipment ! 1.F.2(5) Establish Qualification Requirements for QA and QC Person- a nel I.F.2(6) Increase the Size of Licensees' QA Staff f t I.F.2(7) Clarify that the QA Program Is a Condition of the Construc- a tion Permit and Operating License I.F.2(8) Compare NRC QA Requirements with Those of Other Agen- a caes I.F.2(9) Clarify Organizational Reporting Levels for the QA Organi- f . zation .j I.F.2(10) Clarify Requirements for Maintenance of *As-Built" Docu- a mentation I.F.2(ll) Define Role of QA in Design and Analysis Activities a 1.G.1 Training Requirements f q 1.G.2 Scope of Test Program f . II. A.1 Siting Policy Reformulation c , II.A.2 Site Evaluation of Existing Facihties e j ll.B.1 Reactor Coolant System Vents g See SSAR Subsection 1.9.3, item (2)(vi). II.B.2 Plant Shielding to Provide Access to Vital Areas and Protect g See SSAR Subsection - ; Safety Equipment for Post-Accident Operation 1.9.3, item (2)(vii). l 11.B.3 Post-Accident Sampling g See SSAR Subsection , 1.9.3, item (2)(viii). ll.B.4 Training for Mitigating Core Damage f j 100.8-31 l l
l l 1 NRC REQUEST FOR ADDITIONAL INFORMATION Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable item / Issue Title Screening Notes No. Criteria II.B.5(1) Behavior of Severely Damages Fuel d II.B.5(2) Behavior of Core Melt d II.B.5(3) Effect ofIlydrogen Burning and Explosions on Containment d Structures ll.B.6 Risk Reduction for Operating Reactors at Sites with liigh f Population Densities II.B.7 Analysis of liydrogen Control e II.B.8 Rulemaking Proceedings on Degraded Core Accidents g See SSAR Subsection 1.9.3, items (1)(i), (1)(xii), (2)(ix), (3)(iv), and (3)(v). II.C.1 Interim Reliability Evaluation Program c II.C.2 Continuation of Interim Reliability Evaluation Program c II.C.3 Systems Interaction e ll.C.4 Reliability Engineering c II.D.1 Testing Requirements g See SSAR Subsection 1.9.3, item (2)(x). II.D.2 Research on Relief and Safety Valve Test Requirements a II.D.3 Relief and Safety Valve Position Indication g See SSAR Subsection 1.9.3, item (2)(xi). II.E.1.1 Auxiliary Feedwater System Evaluation g See SSAR Subsection 1.9.3, item (1)(ii). II.E.1.2 Auxiliary Feedwater System Automatic Initiation aid Flow g See SSAR Subsection Indication 1.9.3, items (1)(ii) and (2)(xii). 100.8-32 3 WCStirigh0LIS8
NRC REQUEST FOR ADDITIONAL INFORMATION dH HQ I Table 1.9-2 Listing of Unresobed Safety Issues and Generic Safety Issues Action Plan Applicable Itern/ Issue Title Screening Notes No. Criteria II.E.1.3 Update Standard Review Plan and Develop Regulatory Guide d ll.E.2.1 Reliance on ECCS e ll.E.2.2 Research on Small Break LOCAs and Anomalous Transients c II. E.2.3 Uncertainties in Performance Predictions a ll.E.3.1 Reliability of Power Supplies for Natural Circulation g See SSAR Subsection 1.9.3, item (2)(xiii). II.E.3.2 Systems Reliability e ll.E.3.3 Coordinated Study of Shutdown Heat Removal Requirements e II.E.3.4 Alternate Concepts Research c II.E.3.5 Regulatory Guide e ll.E.4.1 Dedicated Penetrations g See SSAR Subsection 1.9.3, item (3)(vi). II.E.4.2 Isolation Dependability g See SSAR Subsection 1.9.3, item (2)(xiv). II.E.4.3 Integrity Check c II.E.4.4 Purging g See SSAR Subsection 1.9.3, item (2)(xv). II.E.5.1 Design Evaluation b II.E.5.2 B&W Reactor Transient Response Task Force b II.E.6.1 Test Adequacy Study d II.F.1 Additional Accident Monitoring Instrumentation g See SSAR Subsection 1.9.3, item (2)(xvii). i l 1 1 I 100.8-33 W WB5tirighotise l i j
m_- -_ _ _ . . _ _ _ . _ _ . ._ _ __ _ _ I r NRC REQUEST FOR ADDITIONALINFORMATION i l gu ug i- l Table 1.9-2 i Listing of Unresolved Safety issues and Generic Safety Issues ! Action Plan Applicable Item / Issue Title Screening Notes No. Criteria i ll.F.2 Identification of and Recovery from Conditions leading to g See SSAR Subsection Inadequate Core Cooling 1.9.3, item (2)(xviii). l II.F.3 Instruments for Monitoring Accident Conditions g See SSAR Subsection l.9.3, item (2)(xix). j II.F.4 Study of Control and Protective Action Design Requirements a l II.F.5 Classification of Instrumentation, Control, and Electrical d f Equipment i 11.G.1 Power Supplies for Pressurizer Relief Valves, Block Valves, g See SSAR Subsection j and level Indicators 1.9.3, item (2)(xx). -{ 1 i II.II.1 Maintain Safety of TMI-2 and Minimize Environmental c j Impact il.H.2 Obtain Technical Data on the Conditions inside the TMI-2 b 1 Containment Structure l II.H.3 Evaluate and Feed Back Information Obtained from TM1 e 11.11.4 Determine Impact of TMI on Socioeconomic and Real Prop- d erty Values II.J. l .1 Establish a Priority System for Conducting Vendor Inspec- d tions II.J.1.2 Modify Existing Vendor Inspection Program d 11.J.1.3 Increase Regulatory Control Over Present Non-Licensees d 11.J.1.4 Assign Resident Inspectors to Reactor Vendors and Architect- d Engineers II.J.2.1 Reorient Construction Inspection Program d II.J.2.2 Increase Emphasis on Independent Measurement in Construc- d tion Inspection Program 100.8-34 ! l l 1 i l
1 1 i I NRC REQUEST FOR ADDITIONAL INFORMATION iill! ii21 1 m li' Table 1.9-2 Listing of Unresobed Safety Issues and Generic Safety Issucs Action Plan ' Applicable Item / Issue Title Screening Notes No. Criteria II.J.2.3 Assign Resident inspectors to All Construction Sites d II.J.3.1 Organization and Staffing to Oversee Design and Construc- e tion li.J.3.2 Issue Regulatory Guide e II.J.4.1 Revise Deficiency Reporting Requirements d II.K.l(l) Review TMI-2 PNs and Detailed Chronology of the TMI-2 f Accident U.K.l(2) Review Transients Similar to TMI-2 That Have Occurred at b Other Facilities and NRC Evaluation of Davis-Besse Event II.K.l(3) Review Operating Procedures for Recognizing, Preventing, f and Mitigating Void Formation in Transients and Accidents II.K. l(4) Review Operating Procedures and Training Instructions f II.K.l(5) Safety-Related Valve Position Description f II.K. l(6) Review Containment Isolation Initiation Design and Proce- f dures !!.K.l(7) Implement Positive Position Controls on Valves That Could b Compromise or Defeat AFW Flow II.K.l(8) Impicment Procedures That Assure Two Independent 100% b AFW Flow Paths ll.K.l(9) Review Procedures to Assure That Radioactive Liquids and f Gases Are Not Transferred out of Containment Inadvertently 11.K. l(10) Review and Modify Procedures for Removing Safety-Related f j Systems from Service l l 100.8-35 l 3 WB5thgfl3t!S8 ! l I I I l
l NRC REQUEST FOR ADDITIONAL INFORMATION illi itili 11 Table 1.9-2 Listing of Unresolved Safety issues and Generic Safety Issues Action Plan Applicable item / Issue Title Screening Notes No. Criteria II.K.l(11) Make All Operating and Maintenance Personnel . Aware of the f Seriousness and Consequences of the Erroneous Actions Leading up to, and in Early Phases of, the TMI-2 Accident II.K. l(12) One llour Notification Requirement and Continuous Commu- f nications Channels II.K.l(13) Propose Technical Specification Changes Reflecting imple- f mentation of All Bulletin items II.K.l(14) Review Operating Modes and Procedures to Deal with Sig- f nificant Amounts of liydrogen ll.K.l(15) For Facilities with Non-Automatic AFW Initiation, Provide f Dedicated Operator in Continuous Communication with CR to Operate AFW II.K.l(16) Implement Procedures That Identify PZR PORV "Open" f Indications and That Direct Operator to Close Manually at
- Reset
- Setpoint II.K.l(17) Trip PZR Level Bistable so That PZR Imw Pressure Will f Initiate Safety Injection ll.K.l(18) Develop Procedures and Train Operators on Methods of b Establishing and Maintaining Natural Circulation ll.K.l(19) Describe Design and Procedure Modifications to Reduce b Likelihood of Automatic PZR PORV Actuation in Transients II.K.l(20) Provide Procedures and Training to Operators for Prompt b Manual Reactor Trip for LOFW, TT, MSIV Closure, LOOP, LOSG Level, and LO PZR 12 vel II.K.l(21) Provide Automatic Safety-Grade Anticipatory Reactor Trip b for LOFW, TT, or Significant Decrease in SG Level 100.8-36
[ WC5tkl"f E S8
l NRC REQUEST FOR ADDITIONAL INFORMATION j Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable item / Issue Title Screening Notes No. Criteria II.K.1(22) Describe Automatic and Manual Actions for Proper Function- g See SSAR Subsection ing of Auxiliary IIeat Removal Systems When FW System 1.9.3, item (2)(xxi). Not Operable II.K.1(23) Desenbe Uses and Types of RV Level Indication for Auto- b matic and Manual Initiation Safety Systems II.K.l(24) Perform LOCA Analyses for a Range of Small-Break Sizes e and a Range of Time Lapses Between Reactor Trip and RCP Trip II.K.1(25) Develop Operator Action Guidelines e II.K.l(26) Revise Emergency Procedures and Train Ros and SROs e II.K. l(27) Provide Analyses and Develop Guidelines and Procedures for e Inadequate Core Cooling Conditions ll.K.l(28) Provide Design That Will Assure Automatic RCP Trip for e All Circumstances Where Required II.K.2(1) Upgrade Timeliness and Reliability of AFW System b II.K.2(2) Procedures and Training to Initiate and Control AFW Inde- b pendent of Integrated Control System II K.2(3) liard-Wired Control-Grade Anticipatory Reactor Trips b !!.K.2(4) Small-Break LOCA Analysis, Procedures and Operator b Training II.K.2(5) Complete TM1-2 Simulator Training for All Operators b II.K.2(6) Reevaluate Analysis of Dual-Level Setpoint Control b II.K.2(7) Reevaluate Transient of September 24,1977 b II.K.2(8) Continued Upgrading of AFW System e II.K.2(9) Analysis and Upgrading of Integrated Control System e 3 Westbgtlatise
NRC REQUEST FOR ADDITIONAL INFORMATION iiii! ti!!
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Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Itern/ Issue Title Applicable No. Screening Notes Criteria - II.K.2(10) Hard-Wired Safety-Grade Anticipatory Reactor Trips b II.K.2(11) Operator Training and Drilling b II.K.2(12) Transient Analysis and Procedures for Management of Small e Breaks II.K.2(13) Thermal-Mechanical Report on Effect of HPl on Vessel b Integrity for Small-Break LOCA With No AFW II.K.2(14) Demonstrate That Predicted Lift Frequency of PGRVs and b SVs Is Acceptable II.K.2(15) Analysis of Effects of Slug Flow on Once-Through Steam b Generator Tubes After Pnmary System Voiding II.K.2(16) Impact of RCP Seal Damage Following Small-Break LOCA g See SSAR Subsection With Loss of Offsite Power 1.9.3, item (1)(iii). II.K.2(17) Analysis of Potential Voiding in RCS During Anticipated b Transients ll.K.2(18) Analysis of Loss of Feedwater and Other Anticipated Tran- e sients ll.K.2(19) Benchmark Analysis of Sequential AFW Flow to Once- b Through Steam Generator t i II.K.2(20) Analysis of Steam Response to Small-Break LOCA b l II.K.2(21) LOFT L3-1 Predictions l b { II.K.3(1) Install Automatic PORV lsolation System and Perform Oper- g l ational Test See SSAR Subsection 1.9.3, item (1)(iv). II.K.3(2) Report on Overall Safety Effect of PORV Isolation System g See SSAR Subsection 1.9.3, item (1)(iv). l 100.8-38 [ WB5thgh0!!SO
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t NRC REQUEST FOR ADDITIONALINFORMATION A i Table 1.9-2 ! Listing of Unresolved Safety Issues and Generic Safety Issues i i Action Plan Applicable -! Item / Issue Title Screening Notes ! No. Criteria .l. II.K.3(3) Report Safety and Relief Valve Failures Promptly and Chal- f 'f lenges Annually 3 II.K.3(4) Review and Upgrade Reliability and Redundancy of Non- e i Safety Equipment for Small-Break LOCA Mitigation ; II.K.3(5) Automatic Trip of Reactor Coolant Pumps f II.K.3(6) Instrumentation to Verify Natural Circulation e ! II.K.3(7) Evaluation of PORV Opening Probability During Overpres- b sure Transient , II.K.3(8) Further Staff Consideration of Need for Diverse Decay Heat e Removal Method Independent of SGs j II.K.3(9) Proportional Integral Derivative Controller Modification g See SSAR Subsection Il 1.9.4, item II.K.3(9). ; t Anticipatory Trip Modification Proposed by Lome Licensees II.K.3(10) f to Confine Range of Use to High Power Levels II.K.3(11) Control Use of PORV Supplied by Control Components, Inc. f l Until Further Review Complete { II.K.3(12) Confirm Existence of Anticipatory Trip Upon Turbine Trip f II.K.3(13) Separation of HPCI and RCIC System Initiation Levels b-l II.K.3(14) Isolation of Isolation Condensers on High Radiation b II.K.3(15) Modify Break Detection Logic to Prevent Spurious Isolation b ; of HPCI and RCIC Systems II.K.3(16) Reduction of Challenges and Failures of Relief Valves - b Feasibility Study and System Modification ; II.K.3(17) Report on Outage of ECC Systems - Licensee Report and b - Technical Specification Changes
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100.8-39 , 1 r
NRC REQUEST FOR ADDITIONALINFORMATION Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable Itern/ Issue Title Screening Notes No. Criteria !!.K.3(18) Modification of ADS logic - Feasibility Study and Modifica- g See SSAR Subsection tion for Increased Diversity for Some Event Sequences 1.9.3, item (1)(vii). II.K.3(19) Interlock on Recirculation Pump Loops b II.K.3(20) Loss of Service Water for Big Rock Point b II.K.3(21) Restart of Core Spray and LPCI Systems on Low Level - b Design and Modification II.K.3(22) Automatic Switchover of RCIC System Suction - Verify b Procedures and Modify Design II.K.3(23) Central Water Level Recording e II.K.3(24) Confirm Adequacy of Space Cooling for llPCI and RCIC b Systems II.K.3(25) Efftet of Loss of AC Power on Pump Seals g See SSAR Subsection 1.9.3, item (1)(iii). II.K.3(26) Study Effect on RilR Reliability of Its Use for Fuel Pool e Cooling II.K.3(27) Provide Common Reference Level for Vessel Level Instru- b mentation l II.K.3(28) Study and Verify Qualification of Accumulators on ADS g See SSAR Subsection Valves 1.9.3, item (1)(x). , II.K.3(29) Study to Demonstrate Performance of Isolation Condensers b l with Non-Condensibles ) II.K.3(30) Revised Small-Break LOCA Methods to Show Compliance f with 10 CFR 50, Appendix K j II.K.3(31) Plant-Specific Calculations to Show Compliance with 10 CFR f 50.46 , i I 100.8-40 DN
NRC REQUEST FOR ADDITIONALINFORMATION Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable item / Issue Title Screening Notes No. Criteria II.K.3(32) Provide Experimental Verification of Two-Phase Natural e Circulation Models ll.K.3(33) Evaluate Elimination of PORV Function e ll.K.3(34) Relap-4 Model Development e ll.K.3(35) Evaluation of Effects of Core Flood Tank Injection on Small- e Break LOCAs II.K.3(36) Additional Staff Audit Calculations of B&W Small-Break e LOCA Analyses II.K.3(37) Analysis of B&W Response to Isolated Small-Break LOCA e II.K.3(38) Analysis of Plant Response to a Small-Break LOCA in the e Pressurizer Spray Line II.K.3(39) Evaluation of Effects of Water Slugs in Piping Caused by e HPI and CFT Flows II.K.3(40) Evaluation of RCP Seal Damage and Leakage During a e Small-Break LOCA II.K.3(41) Submit P.edictions for LOFT Test L3-6 with RCPs Running e II.K.3(42) Submit Requested Information on the Effects of Non-Conden- e ! sible Gases II.K.3(43) Evaluation of Mechanical Effects of Slug Flow on Steam e l Generator Tubes i II.K.3(44) Evaluation of Anticipated Transients with Single Failurc to b Verify No Significant Fuel Failure , II.K.3(45) Evaluate Depressurization with Other Than Full ADS b II.K.3(46) Response to List of Concerns from ACRS Consultant b [ WB5tlilgt10!!S8
i NRC REQUEST FOR ADDITIONALINFORMATION I 98 '!!Ei Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria II.K.3(47) Test Program for Small-Break LOCA Model Verification e Pmtest Prediction, Test Program, and Model Verification II.K.3(48) Assess Change in Safety Reliability as a Result of Implement- e ing B&OTF Recommendations ll.K.3(49) Review of Procedures (NRC) e II.K.3(50) Review of Procedures (NSSS Vendors) e II.K.3(51) Symptom-Based Emergency Procedures e II.K.3(52) Operator Awareness of Revised Emergency Procedures e II.K.3(53) Two Operators in Control Room e ll.K.3(54) Simulator Upgrade for Small-Break LOCAs e ll.K.3(55) Operator Monitoring of Control Board e II.K.3(56) Simulator Training Requirements e II.K.3(57) Identify Water Sources Prior to Manual Activation of ADS b III.A.I.l(l) Implement Action Plan Requirements for Promptly Improv- f ing Licensee Emergency Preparedness III. A.I. l(2) Perform an Integrated Assessment of the Implementation f III. A. I.2 Upgrade Licensee Emergency Suppon Facilities g See SSAR Subsection 1.9.3, item (2)(xxv). I!!.A.I.3(1) Maintain Supplies of Thyroid-Blocking Agent - Workers c Ill. A. I .3(2) Maintain Supplies of Thyroid-Blocking Agent - Public c Ill.A.2.l(l) Publish Proposed Amendments to the Rules d Ill. A.2. l(2) Conduct Public Regional Meetings d 100.8-42 YMM
NRC REQUEST FOR ADDITIONAL INFORMATION lEE hie Table 1.9-2 Listing of Unresobed Safety issues and Generic Safety Issues Action Plan Applicable Itan/ Issue Title Screening Notes No. Criteria Ill. A.2. l(3) Prepare Final Commission Paper Recommending Adaption of d Rules III. A.2. I(4) Revise Inspection Program to Cover Upgraded Requirenwnts d III. A.2.2 Development of Guidance and Criteria d Ill. A.3. I(l) Define NRC Role in Emergency Situations c Ill. A.3. I(2) Revise and Upgrade Plans and Procedures for the NRC c Emergency Operations Center III. A.3. l(3) Revise Manual Chapter 0502, Other Agency Procedures, and c NUREG-0610 III. A.3.1(4) Prepare Commission Paper c III. A.3.1(5) Revise Implementing Procedures and Instructions for Region- c al Offices III.A.3.2 Improve Operations Centers c Ill. A.3.3 Communications d III. A.3.4 Nuclear Data Link c III. A.3.5 Training, Drills, and Tests c III. A.3.6(1) Interaction of NRC and Other Agencies -International c III. A.3.6(2) Federal c III.A.3.6(3) State and local c III.B.1 Transfer of Responsibilities to FEM A c III.B.2(1) The Licensing Process c III.B.2(2) Federal Guidance c Ill.C. l(l) Review Publicly Available Documents d
' '843 W westhghouse
NRC REQUEST FOR ADDITIONAL INFORMATION f Mji Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria Ill.C.l(2) Recommend Publication of AdditionalInformation d III.C.l(3) Program of Seminars for News Media Personnel d Ill.C.2(1) Develop Policy and Procedures for Dealing With Briefing d Requests III.C.2(2) Provide Training for Member of the Technical Staff d III.D. I . l(l) Review Information Submitted by Licensees Pertaining to g See SSAR Subsection Reducing Leakage from Operating Systems 1.9.3, item (2)(xxvi). Ill.D. I .1(2) Review Information on Provisions for Leak Detection a Ill.D.I.l(3) Develop Proposed System Acceptance Criteria a Ill.D. I.2 Radioactive Gas Manageuent a Ill.D.I.3(1) Decide Whether Licensees Should Perform Studies and Make a Modifications III.D. I .3(2) Review and Revise SRP a Ill.D.I.3(3) Require Licensees to Upgrade Filtration Systems a Ill.D.I.3(4) Sponser Studies to Evaluate Charcoal Adsorber c lil.D.I.4 Radwaste System Design Features to Aid in Accident Recov- a ery and Decontamination Ill.D.2.l(1) Evaluate the Feasibility and Perform a Value-impact Analysis a of Modifying Effluent-Monitoring Design Criteria Ill.D.2.1(2) Study the Feasibility of Requiring the Development of Effec- a tive Means for Monitoring and Sampling Noble Gases and Radiciodine Released to the Atmosphere Ill.D.2.1(3) Revise Regulatory Guides a 100.8-44 MN
NRC REQUEST FOR ADDITIONAL INFORMATION iiii :: Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable item / Issue Title Screening Notes No. Criteria III.D.2.2(1) Perform Study of Radiciodine, Carbon-14, and Tritium c Behavior III.D.2.2(2) Evaluate Data Collected at Quad Cities e Ill.D.2.2(3) Dctermine the Distribution of the Chemical Species of Radio- e iodine in Air-Water-Steam Mixtures III.D.2.2(4) Revise SRP and Regulatory Guides e III.D.2.3(1) Develop Procedures to Discriminate Between Sites / Plants c Ill.D.2.3(2) Discriminate Between Sites and Plants *nat Require Consid- c eration of Liquid Pathway Interdiction Techniques Ill.D.2.3(3) Establish Feasible Method of Pathway Interdiction c Ill.D.2.3(4) Prepare a Summary Assessment c III.D.2.4(1) Study Feasibility of Environmental Monitors c III.D.2.4(2) Place 50 TLDs Around Each Site d III.D.2.5 Offsite Dose Calculation Manual e Ill.D.2.6 Independent Radiological Measurements d III.D.3.1 Radiation Protection Plans c III.D.3.2(1) Amend 10 CFR 20 d Ill.D.3.2(2) Issue a Regulatory Guide d III.D.3.2(3) Develop Standard Performance Criteria d III.D.3.2(4) Develop Method for Testing and Certifying Air-Purifying d Respirators III.D.3.3 In-plant Radiation Monitoring g See SSAR Subsection 1.9.3, item (2)(xxvii). l 100.8-45 ' [ W25thgh0l!SD i i i
l NRC REQUEST FOR ADDITIONAL INFORMATION inr m IH1 Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria III.D.3.4 Control Room Habitability g See SSAR Subsection 1.9.3, item (2)(xxviii). Ill.D.3.5(1) Develop Format for Data To Be Collected by Utilities Re- d garding Total Radiation Exposure to Workers III.D.3.5(2) Investigate Methods of Obtaining Employee Health Data by d Nonlegislative Means Ill.D.3.5(3) Revise 10 CFR 20 d IV. A.1 Seek legislative Authority d IV.A.2 Revise Enforcement Policy d IV.B.1 Revise Practices for issuance of Instructions and Information d to Licensees IV.C.1 Extend Lessons Learned from TMI to Other NRC Programs c IV.D.1 NRC Staff Training d IV.E.1 Expand Research on Quantification of Safety Decision-Mak- d ing IV.E.2 Plan for Early Resolution of Safety Issues d IV.E.3 Plan for Resolving Issues at the CP Stage d IV. E.4 Resolve Generic Issues by Rulemaking d IV.E.5 Assess Currently Operating Reactors c IV.F.1 Increased OIE Scrutiny of the Power-Ascension Test Pro- c gram IV.F.2 Evaluate the Impacts of Financial Disincentives to the Safety c of Nuclear Power Plants IV.G.I Develop a Public Agenda for Rulemaking d 100.8-46 [ Westirighot!$8
I 4 i I NRC REQUEST FOR ADDITIONALINFORMATION Table 1.9-2 Listing of Unresolted Safety Issues and Generic Safety Issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria IV.G.2 Periodic and Systematic Reevaluation of Exising Rules d IV.G.3 Improve Rulemaking Procedures d IV.G.4 Study Alternatives for Improved Rulemaking Process d IV.H.1 NRC Participation in the Radiation Policy Council d V. A.1 Develop NRC Policy Statement on Safety d V.B .1 Study and Recommend, as Appropriate, Elimination of Non- d safety Responsibilities V.C.1 Strengthen the Role of Advisory Committee on Reactor d Safeguards V.C.2 Study Need for Additional Advisory Committees d V.C.3 Study the Need to Establish an Independent Nuclear Safety J Board V.D.1 Improve Public and Intervenor Participation in the Ilearing d Process : V.D.2 Study Construction-During-Adjudication Rules d V.D.3 Reexamine Commission Role in Adjudication d V.D.4 Study the Reform of the Licensing Process d V.E.1 Study the Need for TMI-Related Legislation d V.F.1 Study NRC Top Management Structure and Process d V.F.2 Reexamine Organization and Functions of the NRC Offices d V.F.3 Revise Delegations of Authority to Staff d i V.F.4 Clarify and Strengthen the Respective Roles of Chairman, d 1 Commission, and Executive Director for Operations l 1 .8 m T westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION I f Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable item / Issue Title Screening Notes No. Criteria V.F.5 Authority to Delegate Emergency Response Functions to a d Single Commissioner - V.G.1 Achieve Single location,12mg-Term d V.G.2 Achieve Single location, Interim d Task Action Plan items A-1 Water llammer (former USI) g See SSAR Subsection 1.9.4, item A-1. A-2 Asymmetric Blowdown Loads on Reactor Primary Coolant g See SSAR Subsection Systems (former USI) 1.9.4, item A-2. A-3 Westinghouse Steam Generator Tube Integrity (former USI) g See SSAR Subsection 1.9.4, item A-3. A-4 CE Steam Generator Tube Integrity (former USI) b A-5 B&W Steam Generator Tube Integrity (former USI) b A-6 Mark 1 Short-Term Program (former USI) b A-7 Mark I long-Term Program (former USI) b A-8 Mark 11 Containment Pool Dynamic leads Long-Term Pro- b gram (former USI) A-9 ATWS (former US1) g See SSAR Subsection 1.9.4, item A-9. A-10 BWR Feedwater Nozzle Cracking (former USI) b A-11 Reactor Vessel Materials Toughness (former USI) g See SSAR Subsection 1.9.4, item A-11. A-12 Fracture Toughness of Steam Generator and Reactor Coolant g See SSAR Subsection Pump Supports (former USI) 1.9.4, item A-12. 100.8-48 [ WB5fkigh0t!S8
i i l l l 1 NRC REQUEST FOR ADDITIONAL INFORMATION Table 1.9-2 Listing of Unresolved Safety issues and Generic Safety Issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria A-13 Snubber Operability Assurance g See SSAR Subsection 1.9.4, item A-13. A-14 Flaw Detection a A-15 Primary Coolant System Dec<mtamination and Steam Genera- c tor Chemical Cleaning A-16 Steam Effects on BWR Core Spray Distribution b A-17 Systems Interactions in Nuclear Power Plants (former USI) e A-18 Pipe Rupture Design Criteria a A-19 Digital Computer Protection System d A-20 Impacts of the Coal Fuel Cycle d A-21 Main Steamline Break Inside Containment - Evaluation of a Environmental Conditions for Equipment Qualification A-22 PWR Main Steamline Break - Core, Reactor Vessel and a Containment Building Response A-23 Containment Leak Testing d A-24 Qualification of Class e Safety-Related Equipment (former g See SSAR Subsection USI) 1.9.4, item A-24. A-25 Non-Safety l_ cads on Class e Power Sources g See SSAR Subsection 1.9.4, item A-25. A-26 Reactor Vessel Pressure Transient Protection (former USI) g See SSAR Subsection 1.9.4, item A-26. A-27 Reload Applications d A-28 Increase in Spent Fuel Pool Storage Capacity g See SSAR Subsection 1.9.4, item A-28.
' 849 y.! wemene
NRC REQUEST FOR ADDITIONAL INFORMATION iii!! 11111 IE Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety issues Action Plan Applicable Item / Issue Title Screening Notes i No. Criteria i A-29 Nuclear Power Plant Design for the Reduction of Vulnerabil- c ity to Industrial Sabotage A-30 Adequacy of Safety-Related DC Power Supplies e A-31 RHR Shutdown Requirements (former USI) g See SSAR Subsection 1.9.4, item A-31. i 1 A-32 Missile Effects e A-33 NEPA Review of Accident Risks d See SSAR Subsection 1.9.4, item A-33. A-34 Instruments for Monitoring Radiation and Process Variables e During Accidents A-35 Adequacy of Offsite Power Systems g See SSAR Subsection ; 1.9.4, item A-35. A-36 Control of Heavy loads Near Spent Fuel (former USI) g See SSAR Subsection J 1.9.4, item A-36. A-37 Turbine Missiles a A-38 Tornado Missiles a 1 A-39 Determination of Safety Relief Valve Pool Dynamic Loads b l and Temperature Limits (former USI) A-40 Seismic Design Criteria - Short Term Program (former USI) g See SSAR Subsection 1.9.4, item A-40. A-41 Long Term Seismic Program c 4 i A-42 Pipe Cracks in Boiling Water Reactors (former USI) b l 1 A-43 Containment Emergency Sump Performance (former USI) g See SSAR Subsection l 1.9.4, item A-43. l 100.8-50
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NRC REQUEST FOR ADDITIONALINFORMATION a to 15 f Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria A-44 Station Blackout (former USI) g See SSAR Subsection 1.9.4, item A-44. A-4S Shutdown Decay lieat Removal Requirements (former USI) c A-46 Seismic Qualification of Equipment in Operating Plants g See SSAR Subsection (former USl) 1.9.4, item A-46. A-47 Safety Implications of Control Systems (former USI) g See SSAR Subsection 1.9.4, item A-47. A-48 Ilydrogen Control Measures and Effects of 11ydrogen Burns g See SSAR Subsection on Safety Equipment 1.9.4, item A-48. A-49 Pressurized Thermal Shock (former USI) g See SSAR Subsection 1.9.4, item A-49. B-1 Environmental Technical Specifications d B-2 Forecasting Electricity Demand d B-3 Event Categorization a B-4 ECCS Reliability e B-5 Ductility of Two-Way Slabs and Shells and Buckling Behav- c j ior of Steel Containments B-6 Loads, lead Combinations, Stress Limits e ; B-7 Secondary Accident Consequence Modeling a B-8 locking Out of ECCS Power Operated Valves a j i B-9 Electrical Cable Penetrations of Containment c l B-10 Behavior of BWR Mark III Containments b B-11 Subcompartment Standard Problems d 1 I 1 .841 T westhghotise
1 l NRC REQUEST FOR ADDITIONAL INFORMATION Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria 3-12 Containment Cooling Requirements (Non-LOCA) c B-13 Marviken Test Data Evaluation d B-14 Study of liydrogen Mixing Capability in Containment Post- e LOCA B-15 CONTEMPT Computer Code Maintenance a B-16 Protection Against Postulated Piping Failures in Fluid Sys- e tems Outside Containment B-17 Criteria for Safety-Related Operator Actions h (Medium) See SSAR Subsection 1.9.4, itern B-17. B-18 Vortex Suppression Requirements for Containment Sumps e B-19 Thermal-11ydraulic Stability c B-20 Standard Problem Analysis d B-21 Core Physics a - B-22 LWR Fuel h See SSAR Subsection 1.9.4, item B-22. B-23 LMFBR Fuel a B-24 Seismic Qualification of Electrical and Mechanical Compo- e nents B-25 Piping Benchmark Problems d B-26 Structural Integrity of Containment Penetrations c B-27 Implementation and Use of Subsection NF d B-28 Radionuclide/ Sediment Transport Program d l l 100.8-52 i i i
NRC REQUEST FOR ADDITIONAL INFORMATION mu m: .
- t
}
_ t Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria B-29 Effectiveness of Ultimate Heat Sinks h See SSAR Subsection 1.9.4, item B-29. B-30 Design Basis Floods and Probability d l B-31 Dam Failure Model a B-32 Ice Effects on Safety-Related Water Supplies h See SSAR Subsection 1.9.4, item B-32. B-33 Dose Assessment Methodology d B-34 Occupational Radiation Exposure Reduction e B-35 Confirmation of Appendix I Models for Calculations of d l Releases of Radioactive Materials in Gaseous and Liquid Effluents from Light Water Cooled Power Reactors , i B-36 Develop Design, Testing, and Maintenance Criteria for g See SSAR Subsection ! Atmosphere Cleanup System Air Filtration and Adsorption 1.9.4, item B-36. Units for Engineered Safety Feature Systems and for Normal Ventilation Systems B-37 Chemical Discharges to Receiving Waters d I B-38 Reconnaissance Level Investigations a B-39 Transmission Lines a B-40 Effects of Power Plant Entrainment on Plankton a B-41 Impacts on Fisheries a B-42 Socioeconomic Environment 4d Impacts d B-43 Value of Aerial Photographs for Site Evaluation d B-44 Forecasts of Generating Costs of Coal and Nuclear Plants d B-45 Need for Power - Energy Conservation e l
' '8~53 W westhghouse
I l Nhc REQUEST FOR ADDITIONAL INFORMATION l lill! :14 n: i
- V '
Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria B-46 Cost of Alternatives in Environmental Design a B-47 Inservice Inspection of Supports - Classes 1, 2, 3, and MC a Components B-48 BWR CRD Mechanical Failure (Collet Housing) c B-49 Inservice Inspection Criteria and Corrosion Prevention Crite- d ria for Containments B-50 Post-Operating Basis Earthquake Inspections a B-51 Assessment of Inelastic Analysis Techniques for Equipment e and Components B-52 Fuel Assembly Seismic and LOCA Responses e B-53 Load Break Switch g See SSAR Subsection 1.9.4, item B-53. B-54 Ice Condenser Containments c B-55 Improved Reliability of Target Rock Safety Relief Valves b B-56 Diesel Reliability h (High) See SSAR Subsection 1.9.4, item B-56. B-57 Station Blackout e B-58 Passive Mechanical Failures c , B-59 (N-1) Imop Operation in BWRs and PWRs d B 60 loose Parts Monitoring System c B-61 Allowable ECCS Equipment Outage Periods h (Medium) See SSAR Subsection 1.9.4, item B-61. B-62 Reexamination of Technical Bases for Establishing SLs, a LSSSs, and Reactor Protection System Trip Functions 100.8-54 DDlIS8
i l
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NRC REQUEST FOR ADDITIONAL INFORMATION i liii li
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l Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety ksues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria B43 Isolation of Low Pressure Systems Connected to the Reactor g See SSAR Subsection Coolant Pressure Boundary 1.9.4, item B43. B44 Decommissioning of Reactors f B45 hxline Spiking a B46 Control Room Infiltration Measurements g See SSAR SubsecGon 1.9.4, item B46. B47 Effluent and Process Monitoring Instrumentation e B48 Pump Overspeed During LOCA a B49 ECCS Leakage Ex-Containment e B-70 Power Grid Frequency Degradation and Effect on Primary c Coolant Pumps B-71 Incident Response e B-72 Health Effects and Life Shortening from Uranium and Coal d Fuel Cycles B-73 Monitoring for Excessive Vibration Inside the Reactor Pres- e sure Vessel C-1 Assurance of Continuous long Term Capability of Hermetic g See SSAR Subsection Seals on Instrumentation and Electrical Equipment 1.9.4, item C-1. C-2 Study of Containment Depressurization by Inadvertent Spray c Operation to Determine Adequacy of Containment Extemal Design Pressure C-3 Insulation Usage Within Containment e C-4 Statistical Methods for ECCS Analysis g See SSAR Subsection 1.9.4, item C-4. [ WBStirigflDilSO 1
l NRC REQUEST FOR ADDITIONAL INFORMATION Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable 1 Item / Issue Title Screening Notes No. Criteria C-5 Decay Heat Update g See SSAR Subsection 1.9.4, item C-5. C-6 LOCA Heat Sources g See SSAR Subsection 1.9.4, item C-6. C-7 PWR System Piping c C-8 Main Steam Line Leakage Control Systems Ib C-9 RHR Heat Exchanger Tube Failures a C-10 Effective Operation of Containment Sprays in a LOCA g See SSAR Subsection 1.9.4, item C-10. C-11 Assessment of Failure and Reliability of Pumps and Valves c C-12 Primary System Vibration Assessment c C-13 Non-Random Failures e C-14 Storm Surge Model for Coastal Sites a C-15 NUREG Report for Liquids Tank Failure Analysis a C-16 Assessment of Agricultural Land in Relation to Power Plant a-Siting and Cooling System Selection C-17 Interim Acceptance Criteria for Solidification Agents for g See SSAR Subsection Radioactive Solid Wastes 1.9.4, item C-17. D-1 Advisability of a Seismic Scram a D-2 Emergency Core Cooling System Capability for Future Plants a D-3 Control Rod Drop Accident c New Generic Issues 100.8-56
I i 4 NRC REQUEST FOR ADDITIONAL INFORMATION liii! iiiij i I!I Table 1.9-2 Listing of Unresolsed Safety issuts and Generic Safety Issues Action Plan Applicable item / Issue Title Screening Notes ~ No. Criteria
- 1. Failures in Air-Monitoring, Air-Cleaning, and Ventilating a Systems
- 2. Failure of Protective Devices on Essential Equipment a
- 3. Set Point Drift in Instrumentation c
- 4. End-of-Life and Maintenance Criteria e
- 5. Design Check and Audit of Balance-of-Plant Equipment e
- 6. Separation of Control Rod from Its Drive and BWR 11igh c Rod Worth Events
- 7. Failures Due to Flow-Induced Vibrations a
- 8. Inadvertent Actuation of Safety injection in PWRs e
- 9. Reevaluation of Reactor Coolant Pump Trip Criteria e
- 10. Surveillance and Maintenance of TIP isolation Valves and a Squib Charges
- 11. Turbine Disc Cracking e
- 12. BWR Ju Pump Integrity c
- 13. Small Break LOCA from Extended Overheating of Pressuriz- a er IIcaters
- 14. PWR Pipe Cracks c
- 15. Radiation Effects on Reactor Vessel Supports h (High) See SSAR Subsection 1.9.4, item 15.
- 16. BWR Main Steam Isolation Valve Leakage Control Systems e l
- 17. less of Offsite Power Subsequent to LOCA a j
- 18. Steam Line Break with Consequential Small LOCA e I i
100.8-57 ; [ Westhgtt0t!Se
NRC REQUEST FOR ADDITIONAL INFORMATION illi "i!! EI Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria
- 19. Safety implications of Nonsafety Instrument and Control e Power Supply Bus
- 20. Effects of Electromagnetic Pulse on Nuclear Power Plants c
- 21. Vibration Qualification of Equipment a
- 22. Inadvertent Boron Dilution Events e
- 23. Reactor Coolant Pump Seal Failures h (High) See SSAR Subsection 1.9.4, item 23.
- 24. Automatic Emergency Cere Cooling System Switch to Recir- a culation
- 25. Automatic Air Header Dump on BWR Scram System b
- 26. Diesel Generator leading Problems Related to SIS Reset on e Loss of Offsite Power
- 27. Manual vs. Automated Actions e
- 28. Pressurized Thermal Shock e
- 29. Bolting Degradation er Failure in Nuclear Power Plants h(High) See SSAR Subsection 1.9.4, item 29,
- 30. Potential Generator Missiles - Generator Rotor Retaining a Rings
- 31. Natural Circulation Cooldown e
- 32. Flow Blockage in Essential Equipment Caused by Corbicula e
- 33. Correcting Atmospheric Dump Valve Opening Upon Loss of e Integrated Control System Power
- 34. RCS Leak a
- 35. Degradation of Internal Appurtenances in LWRs a 100.8-58
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NRC REQUEST FOR ADDITIONAL INFORMATION Table 1.9 2 l Listing of Unresolved Safety Issues and Generic Safety issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria
- 36. Loss of Service Water c
- 37. Steam Generator Overfill and Combined Primary and Sec- e ondary Blowdown
- 38. Potential Recirculation System Failure as a Consequence of a injection of Containment Paint Flakes or Other Fine Debris
- 39. Potential for Unacceptable Interaction Between the CRD e System and Non-Essential Control Air System
- 40. Safety Concerns Associated with Pipe Breaks in the BWR b Scram System
- 41. BWR Scram Discharge Volume Systems b
- 42. Combination Primary / Secondary System LOCA e
- 43. Reliability of Air Systems f
- 44. Failure of Saltwater Cooling System e
- 45. Inoperability of Instrumentation Due to Extreme Cold Weath- g See SSAR Subsection er 1.9.4, item 45.
- 46. Loss of 125 Volt DC Bus e
- 47. Loss of Off-Site Power c
- 48. LCO for Class e Vital Instrument Buses in Operating Reac- e tors
- 49. Interkicks and LCOs for Redundant Class e Tie Breakers e
- 50. Reactor Vessel Level Instrumentation in BWRs c
- 51. Proposed Requirements for Improving the Reliability of Open g See SSAR Subsection Cycle Service Water Systems 1.9.4, item 51.
- 52. SSW Flow Bk>ckage by Blue Mussels e i
' '8'"
W westinghouse i 1
NRC REQUEST FOR ADDITIONAL INFORMATION Table 1.9-2 Listing of Unresolved Safety issus and Generic Safety Issues Action Plan Applicable item / Issue Title Screening Notes No. Criteria
- 53. Consequences of a Postulated Flow Blockage incident in a a BWR
- 54. Valve Operator-Related Events Occurring During 1978, e 1979, and 1980
- 55. Failure of Class e Safety-Related Switchgear Circuit Breakers a to Close on Demand
- 56. Abnormal Transient Operating Guidelines as Applied to a e Steam Generator Overfill Event
- 57. Effects of Fire Protection System Actuation h (Medium) See SSAR Subsection 1.9.4, item 57,
- 58. Inadvertent Containment Fk>oding a
- 59. Technical Specification Requirements for Plant Shutdown d when Equipment for Safe Shutdown is Degraded or Inopera-ble
- 60. Lamellar Tearing of Reactor Systems Structural Supports e
- 61. SRV Line Break Inside the BWR Wetwell Airspace of Mark c I and 11 Containments
- 62. Reactor Systems Bolting Applications e
- 63. Use of Equipment Not Classified as Essential to Safety in a BWR Transient Analysis
- 64. Identification of Protection System Instrument Sensing Lines c
- 65. Probability of Core-Melt Due to Component Cooling Water e System Failures
- 66. Steam Generator Requirements c 67.2.1 Integrity of Steam Generator Tube Sleeves d 100.8-60 3 WBSIL @ 8
NRC REQUEST FOR ADDITIONAL INFORMATION I Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable Itern/ Issue Title Screening Notes No. Criteria 67.3.1 Steam Generator Overfill e 67.3.2 Pressurized Thermal Shock e 67.3.3 Improved Accident Monitoring e 67.3.4 Reactor Vessel Inventory Measurements e 67.4.1 RCP Trip e 67.4.2 Control Room Design Review e 67.4.3 Emergency Operating Procedures e 67.5.1 Reassessment of SGTR Design Basis d 67.5.2 Reevaluation of SGTR Design Basis d 67.5.3 Secondary System Isolation a 67.6.0 Organizational Responses e 67.7.0 Improved Eddy Current Tests e 67.8.0 Denting Criteria e 67.9.0 Reactor Coolant System Pressure Control e 67.10.0 Supplement Tube Inspections d
- 68. Postulated Loss of Autiliary Feedwater System Resulting e from Turbine-Driven Auxiliary Feedwater Pumps Steam Supply Line Rupture
- 69. Make-up Nozzle Cracking in B&W Plants c
- 70. PORV and Block Valve Reliability g See SSAR Subsection 1.9.3, item (t)(iv).
- 71. Failure of Resin Demineralizer Systems and Their Effects on a Nuclear Power Plant Safety WB5tlflgt10l!SB
i l NRC REQUEST FOR ADDITIONA1.INFORMATION l r, '
.i Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues t Action Plan _
Applicab!c item / Issue Title Screening Notes , No. Criteria
- 72. Control Rod Drive Guide Tube Suppon Pin Failures a
- 73. Detached Thermal Sleeves a ,
- 74. Reactor Coolant Activity Limits for Operating Reactors a
- 75. Generic Implications of.ATWS Events at the Salem Nuclear d Plant ;
- 76. Instrumentation and Control Power Interactions a -i
- 77. Flooding of Safety Equipment Compartments by Back-flow e .l Through Floor Drains l
- 78. Monitoring of Fatigue Transient Limits for Reactor Coolant a ;
System
- 79. Unanalyzed Reactor Vessel Thermal Stress During Natural h (Medium) See SSAR Subsection i Circulation Cooldown 1.9.4, item 79. !
- 80. Pipe Break Effects on Control Rod Drive liydraulic Lines in a !
the Drywells of BWR Mark I and 11 Containments
- 81. Impact of Locked Doors and Barriers on Plant and Personnel a l Safety +
i
- 82. Beyond Design Basis Accidents in Spent Fuel Pools e i q 83. Control Room Habitability h See SSAR Subsection i 1.9.4, item 83.
- 84. CE PORVs c :
- 85. Reliability of Vacuum Breakers Connected to Steam Dis- a f charge Lines inside BWR Containments
- 56. Long Range Plan for Dealing with Stress Corrosion Cracking b 'f in BWR Piping !
6 100.8-62 f i i
NRC REQUEST FOR ADDITIONAL INFORMATION
!!H Hil m !!L Table 1;o-?
Listing of Unresolved Safety Issues and Generic Safety issues Action Plan Applicable item / Issue Title Screening Notes No. Criteria
- 87. Failure of HPCI Steam Line Without Isolation h (High) See SSAR Subsection 1.9.4, item 87.
- 88. Earthquakes and Emergency Planning c
- 89. Stiff Pipe Clamps a
- 90. Technical Specifications for Anticipatory Trips a
- 91. Main Crankshaft Failures in Transamerica DeLaval Emergen- c ey Diesel Generators
- 92. Fuel Crumbling During LOCA a
- 93. Steam Binding of Auxiliary Feedwater Pumps g See SSAR Subsection 1.9.4, item 93.
- 94. Additional Low Temperature Overpressure Protection for g See SSAR Subsection Light Water Reactors 1.9.4, item 94.
- 95. Loss of Effective Volume for Containment Recirculation c Spray
- 96. RilR Suction Valve Testing e
- 97. PWR Reactor Cavity Uncontrolled Exposures e
- 98. CRD Accumulator Check Valve Leakage a
- 99. RCS/RHR Suction Line Valve Interlock on PWRs f 100. OTSG Level b 101. BWR Water Level Redundancy c 102. Human Error in Events involving Wrong Unit or Wrong c Train 103. Design for Probable Maximum Precipitation g See SSAR Subsection 1.9.4, item 103.
' 8~S3 w wemeuse 1
NRC REQUEST FOR ADDITIONAL INFORMATION i Tab!: 1.9 2 Listing of Unresolied Safety Issues and Generic Safety hsues Action 11an Applicable Item / Issue Title Screening Notes No. Criteria 104. Reduction of Boron Dilution Requirements a 105. Interfacing Systems LOCA at BWRs b (High) See SSAR Subsection 1.9.4, item 105. 106. Piping and Use of Highly Combustible Gases in Vital Areas h (Medium) See SSAR Subsection 1.9.4, item 106. 107. Main Transformer failures a 108. BWR Suppression Pool Temperature Limits a 109. Reactor Vessel Closure Failure a 110. Equipment Protective Devices on Engineered Safety Features a 111. Stress Corrosion Cracking of Pressure Boundary Ferritic d Steels in Selected Environments 112. Westinghouse RPS Surveillance Frequencies and Out-of- d Senice Times 113. Dynamic Qualification Testing of Large Bore Hydraulic h (High) See SSAR Subsection Snubbers 1.9.4, item 113. I14. Seismic-Induced Relay Chatter e 115. Enhancement of the Reliability of Westinghouse Solid State c Protection System 116. Accident Management a 117. Allowable Time for Diverse Simultaneous Equipment Outag- a es 118. Tendon Anchorage Failure a 119.1 Piping Rupture Requirements and Decoupling of Seismic and d LOCA Loads 100.8-64 [ Wed1ghollS8
1 NRC REQUEST FOR ADDITIONAL INFORMATION Table 1.9-2 Listing of Unresolved balety luuts and Gennic Safety Inuts Action Plan Applicable item / Issue Title Screening Notes No. Criteria 119.2 Piping Damping Values d 119.3 Decoupling the OBE from the SSE d i19.4 BWR Piping Materials d 119.5 Leak Detection Requirements d 120. On-Line Testability of Protection Systems a 121. Hydrogen Control for Large, Dry PWR Containments h (High) See SSAR Subsection 1.9.4, item 121. 122.1.a Failure of Isolation Valves in Closed Position e 122.1.b Recovery of Auxiliary Feedwater e 122.1.c Interruption of Auxiliary Feedwater Flow e 122.2 Initiating Feed-and-Bleed c 122.3 Physical Security System Constraints a 123. Deficiencies in the Regulations Governing DBA and Single- a failure Criteria Suggested by the Davis-Besse Event of June 9, 1985 124. Auxiliary Feedwater System Reliability g See SSAR Subsection 1.9.4, item 124. 125.1.1 Availability of the STA a 125.1.2.a Need for a Test Program to Establish Reliability of the e PORV 125.1.2.b Need for PORV Surveillance Tests to Confirm Operational e Readiness 125.1.2.c Need for Additional Protection Against PORV Failun a 1 .8-e s W westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION (* s
=-
Table 1.9-2 Listing of Unresolved Safety Issum and Generic Safety Issues Action Plan Applicable Itan/ Issue Title Screening Notes No. Criteria 125.I.2.d Capability of the PORV to Support Feed-and-Bleed e 125.I.3 SPDS Availability c 125.I.4 Plant-Specific Simulator a 125.I.5 Safety Systems Tested in All Conditions Required by Design a Basis Analysis 125.I.6 Valve Torque Limit and Bypass Switch Settings a 125.I.7.a Recover Failed Equipment a 125.I.7.b Realistic 11 ands-On Training a 125.1.8 Procedures and Staffing for Reporting to NRC Emergency a Response Center 125.II.1.s Two-Train AFW unavailability a 125.II. l .b Review Existing AFW Systems for Single Failure e 125.l!.1.c NUREG-0737 Reliability Improvements a 125.II.1.d AFW/ Steam and Feedwater Rupture Control System /ICS a Interactions in B&W Plants 125.11.2 Adequacy of Existing Maintenance Requirements for Safety- a Related Systems 125.11.3 Review Steam /Feedline Break Mitigation Systems for Single a Failure 125.11.4 Thermal Stress of OTSG Components a 125.11.5 Thermal-llydraulic Effects of Imss and Restoration of Feed- a water on Primary System Components 100.8-66 [ Westhgh0t!S8
NRC REQUEST FOR ADDITIONAL INFORMATION
!. k Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable item / Issue Title Screening Notes No. Criteria 125.11.6 Reexamine PRA-Based Estimates of the Likelihood of a a Severe Core Damage Accident Based on less of All Feed-water 125.11.7 Reevaluate Provision
- to Automatically Isolate Feedwater c from Steam Generator l>Mng a Line Break 125.11.8 Reassess Criteria for Feed and-t' teed Initiation a 125.11.9 Enhanced Feed-and-Bleed Capability a 125.11.10 Hierarchy of Impromptu Operator Actions a 125.11.11 Recovery of Main Feedwater as Alternative to AFW a 125.11.12 Adequacy of Training Regarding PORV Operation a 125.11.13 Operator Job Aids a 125.11.14 Remote Operation of Equipment Which Must Now Be Oper- a ated Locally 126. Reliability of PWR Main Steam Safety Valves d 127. Testing and Maintenance of Manual Valves in Safety-Related a Systems 128. Electrical Power Reliability h (High) See SSAR Subsection 1.9.4. item 128.
129. Valve Interlocks to Prevent Vessel Drainage During Shut- a down Cooling 130. Essential Service Water Pump Failures at Multiplant Sites h (High) See SSAR Subsection 1.9.4. item 130. 131. Potential Seismic Interaction involving the iviovable In-Core d Flux Mapping System in Westinghouse Plants 132.. RIIR Pumps inside Containment a
)
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NRC REQUEST FOR ADDITIONAL INFORMATION _iii *: a: _ t Table 1.9-2 Listing of Unr: solved Safety Issues and Generic Safety issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria 133. Update Policy Statement on Nuclear Plant Staff Working d Hours 134. Rule on Degree and Experience Requirements c 135. Integrated Steam Generator Issues h (Medium) See SSA.R Subsection 1.9.4, item 135. 136. Storage and Use of large Quantities of Cryogenic Combusti- d bles On Site 137. Refueling Cavity Seal Failure a 138. Deinerting Upon Discovery of RCS Leakage a 139. Thinning of Carbon Steel Piping in LWRs d 140. Fission Product Removal Systems a 141. LBLOCA With Consequential SGTR a 142. Leakage Through Electrical Isolators in Instrumentation b (Mediurn) See SSAR Subsection Circuits 1.9.4. item 142. 143. Availability of Chilled Water Systems a 144. Scram Without a Turbine / Generator Trip a 145. Improve Surveillance and Startup Testing Programs a 146. Support Flexibility of Equipment and Components a 147. Fire-Induced Altemate Shutdown Control Room Panel Inter- a actions 148. Smoke Control and Manual Fire-Fighting Effectiveness a 149. Adequacy of Fire Barriers a 150. Overpressurization of Containment Penetrations a 1 1 100.8-68 W-WB5tlingh00$8 i l l
i
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I NRC REQUEST FOR ADDITIONAL INFORMATION l ild! *2i ! r p i l 1 Table 1.9-2 Listing of Unresolmi Safety Issues and Generic Safety Issues Action Plan Applicable Item / Issue Title Screening Notes No. Criteria 151. Reliability of Recirculation Pump Trip During an ATWS a 152. Design Basis for Valves That Might Be Subjected to Signifi- a cant Blowdown Loads 153. Loss of Essential Service Water in LWRs a 154. Adequacy of Emergency and Essential Lighting a lluman Factors Issues liFl.1 Shift Staffing f HF1.2 Engineering Expertise on Shift c HFl.3 Guidance on Limits and Conditions of Shift Work c IIF2.1 Evaluate Industry Training d HF2.2 Evaluate INPO Accreditation d ilF2.3 Revise SRP Section 13.2 d HF3.1 Develop Job Knowledge Catalog d IIF3.2 Develop License Examination Handbook d HF3.3 Develop Criteria for Nuclear Power Plant Simulators e HF3.4 Examination Requirements e HF3.5 Develop Computerized Exam System d HF4.1 Inspection Procedure for Upgraded Emergency Operating h (High) See SSAR Subsection Procedures 1.9.4, item HF4.1. HF4.2 Procedures Generation Package Effectiveness Evaluation d H F4.3 Criteria for Safety-Related Operator Actions e HF4.4 Guidelines for Upgrading Other Procedures c W85tlTigt10llS8
NRC REQUEST FOR ADDITIONAL INFORMATION gr nf Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable Itan/ Issue Title Screening Notes No. Criteria HF4.5 Application of Automation and Artificial Intelligence e HF5.1 local Control Stations h (High) See SSAR Subsection 1.9.4, item HF5.1.
}IF5.2 Review Criteria for Human Factors Aspects of Advanced h (High) See SSAR Subsection Controls and Instrumentation 1.9.4, item HF5.2.
IIF5.3 Evaluation of Operational Aid Systems e IIF5.4 Computers and Computer Displays e IIF6.1 Develop Regulatory Position on Management and Organiza- e tion IIF6.2 Regulatory Position on Management and Organization at e Operating Reactors HF7.1 Human Error Data Acquisition d HF7.2 Human Error Data Storage and Retrieval d HF7.3 Reliability Evaluation Specialist Aids d HF7.4 Safety Event Analysis Results Applications d HF8 Maintenance and Surveillance Program c Chernobyl Issues CH l.1 A Symptom-Based EOPs d Cdl.1B Procedure Violations d Cill.2A Test, Change, and Experiment Review Guidelines d CIII.2B NRC Testing Requirements d CHl.3 A Revise Regulatory Guide 1.47 d CHl.4A Engineered Safety Feature Availability d 100.8-70 DDI!S8
NRC REQUEST FOR ADDITIONAL INFORMATION r" ig. A Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable itern/ hsue Title Screening Notes No. Criteria Cill.4B Technical Specificatic Bases d Cl{l.4C low Power and Shutdown d Cill.5 Operating Staff Attitudes Toward Safety d CIII .6A Assessment of NRC Requirements on Management d Cl{1.7A Accident Management d Cil2. l A Reactivity Transients d Cil2.2 Accidents at Ixw Power and at Zero Power e Cil2.3A Control Room liabitability e CII2.3B Contamination Outside Control Room d CII2.3C Smoke Control d Cil2.3D Shared Shutdown Systems d Cll2.4A Firefighting With Radiation Present d C113. l A Containment Performance d Cil3.2A Filtered Venting d Cll4.1 Size of the Emergency Planning Zones a Cil4.2 Medical Services a CII4.3A Ingestion Pathway Protective Measures d CII4.4A Decontamination d Cll4.4B Relocation d CII5.1 A Mechanical Dispersal in Fission Product Release d Cil5.lB Stripping in Fission Product Release d 100.8-71 l g 1 1
NRC REQUEST FOR ADDITIONAL INFORMATION Table 1.9-2 Listing of Unresolved Safety Issues and Generic Safety Issues Action Plan Applicable item / Issue Title Screening Notes No. Criteria CH5.2A Steam Explosions d CH5.3 Combustible Gas d CH6.l A The Fort St. Vrain Reactor and the Modular HTGR a CH6.1B Structural Graphite Experiments d CH6.2 Assessment d Key
- a. Issue has been prioritized as Low, Drop or has not been prioritized.
- b. Issue is not an AP600 design issue. Issue is applicable to GE, B&W, or CE designs only.
- c. Issue resolved with no new requirements.
- d. Issue is not a design issue (Environmental, Licensing, or Regulatory Impact Issue; or covered in an existing .
NRC program).
- e. Issue superseded by one or more issues,
- f. Issue is not an AP600 design certification issue. Issue is applicable to current operating plants or responsibility of combined license applicant.
- g. Issue is resolved by establishment of new regulatory requirements and/or guidance.
- h. Issue is unresolved pending generic resolution (e.g., prioritized as liigh, Medium, or possible resolution identified).
100.8-72
NRC REQUEST FOR ADDITIONAL INFORMATION e :e l 4 !q r i Question 210.27 Provide a written response to the pre-application request for additional information on component capability and l reliability that was transmitted in a May 7,1992 letter, or provide a cross-reference to any response (s) that may l have already been formally submitted.
Response
- 1. The valves for ecch of the AP600 systems are listed in SSAR Table 3.2-3. This table identifies for each valve the safety class, seismic class, and principal construction code. The valve size, body type, and operator type are indicated in the P&lDs. In general, additional information, such as operator size and functional requirements has not been developed for AP600 valves. Such data will be developed in conjunction with the selection of specific valve vendors. See the response to Q210.28 for additional information.
- 2. Specific valve designs and vendors have not been selected for the AP600 valves. That relection will occur after design certification.
- 3. See the response to item 1.
- 4. See the response to item 1.
- 5. The planned activities to ensure adequate valve reliability for safety-related valves are as follows:
- Develop detailed valve specifications that adequately define the valve functional requirements.
- Review valve specifications with valve vendors and utilities; address valve operating problems and solutions to them.
- Westinghouse performs valve feasibility tests where required (ADS and IRWST check valves).
- Valve vendors perform valve qualification tests where required.
- AP600 plant startup tests will demonstrate proper valve performance in the plant.
- AP600 plant inservice tests will demonstrate continued proper valve performance with both at power and shutdown tests.
- A preplanned maintenance program will be developed for these valves and adjusted as necessary based on operating experience.
- 6. The valve reliabilities used in the AP600 PRA are taken from the ALWR URD Volume I!!, Chapter 1, Appendix A, "PRA Key Assumptions and Groundrules."
- 7. Periodic maintenance has not been defined for AP600 valves. Detailed vendor-specific information is required to be able to define the maintenance requirements. Inservice testing of safety-related valves is addressed in the responses to Q210.24 and Q210.25. Testing of non-safety-related defense-in-depth valves is less risk significant. The schedule for testing these valves is defined in SSAR Section 16.2.
21o.2 m W westinghoase
NRC REQUEST FOR ADDITIONAL INFORMATION
- 8. Process flow diagrams and tables are included in the AP600 SSAR in the description for each system. See the response to item 1 and the responses to Q 210.24 and Q210.25 on inservice testing.
- 9. See to the response to Q440.11 on the ADS test facility description and test matrix.
SSAR Revision: NONE 1 i I 1 210.27-2 3 WB5tingh00S8
I l l NRC REQUEST FOR ADDITIONAL INFORMATION Question 210.28 The staff, with the assistance of the Oak Ridge National laboratory (ORNL), has been assessing critical valves for the AP600 design. The staff has identified critical valves and their critical failure modes, and is attempting to develop estimates of the initial reliability of these types of valves. Table A is a list of the critical valves that the staff has identified. However, in order for the staff to complete its assessment, detailed technical information on many of these valves is required. The staff recognizes that some of this detailed information may not be available at the design certification stage of the review. However, pertinent assumptions that were used in determining valve design and reliability may be beneficial to the staff in helping it assess component operability and reliability. These assumptions should be provided where technical information is unavailable. Accordingly, provide the following information for the valves listed in Table, as appropriate: Information Requested on All Valves
- a. What is the process fluid composition (water, steam, raw water, borated water, etc.) assumed for these valves for normal and accident conditions?
- b. What is the maximum temperature cf the process fluid assumed for these valves!
- c. Provide the range of flow, pressure, and differential pressure assumed for these valves for both normal and accident conditions.
- d. Provide any available design information for these valves, e.g., the manufacturer and model number of the valve and actuator.
- e. Are there any design changes in the design of these valves as they are used in c<mventional nuclear generating stations?
- f. List the scope and frequency of planned periodic testing and preventive maintenance.
- g. Describe the planned qualification testing and its potential effect on replacement intervals.
l
- h. Is a method of bonnet over-pressurization protection provided for normally closed gate valves? j Infonnation Requested on Air-Operated Valves l The following information requests pertain to Valve Nos. PXS V108A/B, V014 A&B, V015A&B, PXS V002A&B, V003A&B, and RC V004A/B/C/D.
- i. What elastomeric compounds will be used for the various 0-rings, seats, diaphragms, etc.? l
- j. What lubricants will be used in the valve and solenoids? Provide the manufacturer and trade name, if available.
l 210.28-1 W~~ Westinghouse . l i l
i NRC REQUEST FOR ADDITIONAL INFORMATION er ajit IH
- k. Will the plant meet ISA Standard 57.3, " Quality Standard for Instrument Air,* if instrument air is used with these valves?
Information Requested on Motor-Operated Valves
- 1. What is the anticipated duty cycle of the valve actuator of the following valves: Valve Nos. PXS Vil7A/B, V118A/B, V101, RC V001 A/B/C/D, V002A/B/C/D, and V003A/B/C/D?
Infonnation Requested on Check Valves The following information requests pertain to Valve Nos. PXS V016A/B, V017A/B, V006A/B, V007A/B, V119A/B, V120A/B, V122A/B, V123A/B, V124A/B, and V125A/B.
- m. What is the minimum opening pressure differential?
- n. What kind of position indication system is used?
Infonnation Requested on Explosive Valses The following information requests pertain to Valve No. PXS V301 A/B.
- o. What is the manufacturer and model number of the valve and actuator?
- p. What is the predicted reliability based on operating experience and testing to date?
210.28-2 W westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION
?' S M }((
e Table A Critical Valves For the AP600 (See Q210.28) Valve No. and Description P&ID No. PXS V013A/B (CMT outlet) LO gate valve, manual PXS M6 001 PXS V014A/B (CMT outlet) NC globe AOV PXS M6 001 PXS V015A/B (CMT outlet) NC globe AOV PXS M6 001 PXS V016A/B (CMT outlet) NO tilt disc check valve PXS M6 001 PXS V017A/B (CMT outlet) NO tilt disc check valve PXS M6 001 PXS V002A/B (CMT pressure line) NC globe AOV PXS M6 001 PXS V003A/B (CMT pressure line) NC globe AOV PXS M6 001 PXS V006A/B (CMT pressure line) NO check valve PXS M6 001 PXS V005A/B (CMT pressure line) NO gate MOV PXS M6 001 PXS V030A/B (condensate drain) NC globe SOV PXS M6 001 PXS V031 A/B (condensate drain) NC globe SOV PXS M6 001 PXS V027A/B (accumulator outlet) NO gate MOV PXS M6 001 PXS V028A/B (accumulator outlet) check valve PXS M6 001 PXS V029A/B (accumulator outlet) check valve PXS M6 001 PXS V021 A/B (accumulator N2 line) NO SOV PXS M6 001 PXS V042 (accumulator N 2 line) NO AOV PXS M6 001 PXS V043 (accumulator N2 line) NO check valve PXS M6 001 PXS V04SA/B (accumulator N 2 line) NO SOV PXS M6 001 PXS Vil7A/B (IRWST to sump screen) NC gate MOV PXS M6 002 PXS VI18A/B (IRWST to sump screen) NC gate MOV PXS M6 002 PXS V119A/B (IRWST to sump screen) NC check valve PXS M6 002 PXS V120A/B (IRWST to sump screen) NC check valve PXS M6 002 PXS V121 A/B (IRWST injection) NO gate MOV PXS M6 002 21o.2n W westinghouse
a, NRC REQUEST FOR ADDITIONALINFORMATION ' a m l 1 Table A j (continued) -l Critical Valves For the AP600 - i
.(See Q210.28) !
Valve No. and Description P&lD No.- i PXS V122A/B (IRWST injection) check valve PXS M6 002 , PXS V123A/B (IRWST injection) check valve PXS M6 002 .; PXS V124A/B (IRWST injection) check valve PXS M6 002 i PXS V125A/B (IRWST injection) check valve PXS M6 002 - ; PXS V102A/B (passive RHR HX) LO gate valve PXS M6 002 '{ - PXS V103A/B (passive RHR HX) LO gate valve PXS M6 002 PXS V108A/B (passive RHR HX) NC globe AOV ' PXS M6 002 _l PXS V109A/B (passive RHR HX) LO gate valve PXS M6 002 ~ ! PXS V101 A/B (passive RHR HX) NO gate MOV PXS M6 002 - '! PXS Vill A/B (hot leg H 2drain) NC globe valve PXS M6 002 l PXS V130A/B (gutter loop seal) NO AOV PXS M6 002 , PXS V231A/B (CMT isolation) check valve -PXS M6 003 PXS V230A/B (CMT isolation) NC globe AOV PXS M6 003 PXS V232A/B (accumulator isolation) NC globe AOV PXS M6 003 - , PXS V301 A/B (pH adj. tank injection line) NC squib valve PXS M6 004 : ! PXS V315A/B (pH adj. tank, N 2system) NC vacuum breaker ' PXS M6 004 j RC V004A/B/C/D (ADPs) NC 12" gate AOV . RCS M6 001. .; RC V001 A/B/C/D (ADPs) globe DC MOVs - RCS M6 002 - ; RC V002A/B/C/D (ADPs) gate DC MOVs RCS M6 002 : ? RC V003A/B/C/D (ADPs) gate DC MOVs RCS M6 002 y PC V002A/B (containment cooling line) NO gate MOV PCS M6 001 'i PC V001 A/B (containment cooling line) NC butterfly AOV - PCS M6 001 -l SG V040A/B (main steam line) NO gate AOV SGS M6 001 ; i
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- ~ -ea, -
c.* m+ --
NRC REQUEST FOR ADDITIONAL INFORMATION
Response
The following is a response to the requested valve information, on a item by item basis. Note that in several of the responses, reference is made to AP600 valve requirements, which will be transmitted to the NRC by July 30, 1993. Information Requested on All Valves
- a. The process fluid composition (water, gas, ...) for both normal and accident valve conditions is contained in the valve requirements which will be transmitted to the NRC by July 30,1993.
- b. The maximum temperature for the process fluid is contained in the valve requirements. See item a.
- c. The range of flow, pressure, and differential pressure for normal and accident conditions is contained in the valve requirements. See item a.
- d. The specific valve designs and vendors have not been selected for the AP600 valves. That selection will occur after design certification. As a result, vendor-specific valve information (manufacturer and model numbers) is not available for these valves.
- e. Design change-s have not been identified for these valves, as compared with similar valves used in current plants. Ilowever, some changes are anticipated in the ADS stage 1, 2, and 3 valves as a result of the specification to increase margins and reliability. The specific design changes will not be known until vendors and valve designs are selected, which will occur after design certification.
- f. Inservice testing conditions and frequency are listed in the valve requirements (see item a). See also the responses to Q210.24 and Q210.25. The responses to these questions provide additional information on inservice testing. Preventive maintenance plans have not been developed for these valves because detailed, vendor-specific information is required. That information will become available after design certification, j l
- g. Valve qualification will be provided by the valve vendors after design certification. See SSAR Appendix 3D for additional information on the qualification of valves. This qualification will include testing where current operating experience and testing does not encompass the AP600 valve requirements. In addition, for the ADS l I
valves and the IRWST check valves, there are design certification tests that demcmstrate the feasibility of these valves. ;
- h. The valve requirements referenced in item a address Imnnet overpressurization. These requirements specify j that a passage be provided in the valve body that connects the bonnet area to the upstream side of the gate to l prevent bonnet overpressurization.
Information Requested on Air-operated Valves W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION
- i. The specific valve designs and vendors have not been selected for the AP600 valves. That selection will occur after design certification. As a result, vendor-specific valve information (elastomeric compounds) is not available for these valves.
- j. The specific valve designs and vendors have not been selected for the AP600 valves. That selection will occur after design certification. As a result, vendor-specific valve information (lubricants) is not available for these valves,
- k. The air-operated valves shown in the valve requirements (reference item a) use instrument air, which will meet ISA Standard S7.3.
Information Requested on Motor-operated Valves
- 1. The number of open/close duty cycles is listed in the valve requirements. See item a.
Information Requested on Check Valves
- m. The minimum opening pressure is listed in the valve requirements. See item a.
- n. The kind of position indication system has not been identified. This information will become available after vendors have been selected.
Information Requested on Explosive Valves
- o. The specific valve designs and vendors have not been selected for the AP600 valves. That selection willoccur after design certification. As a result, vendor-specific valve information (manufacturer and model numbers) is not available for these valves.
- p. The reliability used for these valves in the AP600 PRA is shown in the valve requirements (See item a). 'Ihis reliability is based on the ALWR URD Volume III, Chapter 1, Appendix A, 'PRA Key Assumptions and Groundrules," with a factor of 2 increase in failure rate due to the longer refueling cycle of the AP600, as compared with current plants.
SSAR Revision: NONE 210.28-6 W-Westinghouse ! i
NRC REQUEST FOR ADDITIONALINFORMATION
!j"' 1{j Response Revision 1 e
Question 251.11 Section 1 A of the SSAR states that a flywheel rupture will be contained within the stator shell. Provide an analysis and technical justifications supporting this statement. Response (Revision 11: The canned motor reactor coolant pump has an outer shell that comprises the pressure boundary. The shell is analyzed to demonstrate that in the event of a postulated flywheel fracture, the surrounding pump structure is sufficient to prevent missiles from leaving the pump. The analysis considers that portion of the shell, including the flange, and motor end cap around the flywheel assembly between the top and bottom elevations of the assembly as the barrier to missile generation. The Mi-structural analysis summary is i-dud:d documented in Reference 2 a Oyc' ? '= dure! an:!ysiwand is outlined wmmunal-below. 'he pump Oy > h: ! 1 mew ! :n:!ys "! he fma!!:.ed by June L The analysis of the capacity of the surrounding pump structure to contain the fragments of a postulated flywheel failure is done using the energy absorption equations of Reference 1. The containment of missile-like metal disk fragments is by a two-stage process. Stage 1 involves inelastic impact and transfer of momentum to include an effective target mass. To show that the fragments do not perforate the surrounding structure, the energy dissipated in plastic compression and shear strain and the k> cal impact area must be sufficient to account for the loss in kinetic ] energy of the system. For the nonperforation case, the process enters Stage 2, which involves dissipation of energy in plastic tension strain over extended volumes of shell material. For containment, the energy dissipated in plastic strain in Stage 2 must account for the residual kinetic energy on the system. In predictive calculations it is more i conservative to consider Stage 2. l l For the AP600 reactor coolant pump analysis, the uranium insert in the flywheel assembly is assumed to fracture ; at the design speed of 125 percent of normal speed. The worst-case scenario of fragment size and number was derived analytically, using methods in Reference 1, F: paper ~"ed ab: :, to determine the mass and velocity combination that would produce. the most severe impact on the surrounding pressure boundary components. The 1 following conservative assumptions are also made:
- 1. End plates and welds of the flywheel enclosure and the coolant surrounding the flywheel assembly have negligible energy-absorbing capability.
- 2. Only the mass in the stator shell and flange and the motor end cap between the elevation of the top and bottom of the flywheel assembly are considered to absorb energy.
- 3. Closure bolts and joint effects were not considered to be affected.
- 4. The minimum material properties were used.
251.11(R1)-1
l l I NRC REQUEST FOR ADDITIONAL INFORMATION . ) Response Revision 1 i, The analysis results show that the fragments impact the surrounding pump structure with a kinetic energy ofless j than 10 percent of the tensile energy-absorbing capability of the surrounding pump structure. Thus the components i around the flywheel contain the flywheel fragments using we-only a small portion of the energy-absorbing .l capability =:i!:5!:. He energy absorbed by the flywheel enclosure is alea-only a small fraction of the energy- l absorbing capability available in the pump structure. The energy absorbed by.the enclosure end-is not relied on to j i--5te containne+wf the fragments. ,
References:
- 1. Hagg, A. C., and Sankey, G. O., "The Containment of Disk Burst and Fragments by Cylindrical Shells,' ;
ASME Journal for Power, April 1974, pp.114-123. ! 2; WCAP-13734,
- Structural Analysis Summary for the AP600 Reactor Coolant Pump High Inertia Flywheel,"
-l
{ Proprietary), May,1993. ~ SSAR Revision: NONE j i i i r t I. i h h r 6 s 251.11(R1) W Westinghouse -
.t
-i
NRC REQUEST FOR ADDITIONAL INFORMATION l bk Question 280.4 i Provide the test data for the passive containment cooling system that was generatal from both the small and large I scale test facilities that are related to dry shell testing conditions.
Response
Four test data reports for the passive containment cooling system generated from small- and large-scale test facilities under dry shell test conditions are being reissued as WCAPs. The WCAPs will be transmitted to the NRC under i separate cover by May 31,1993 with Westinghouse proprietary 2 and 3 revisions. The WCAP numbers and report l titles are as follows:
- WCAP-13727 (P2) and WCAP-13728 (P3), "Ileavy Water Reactor Facility Project (HWRF) Small Scale Containment Cooling System Test Final Report" l
- WCAP-13732 (P2) and WCAP-13733 (P3), " Heavy Water Reactor Facility (HWRF) Small Scale Containment Cooling Test Preliminary Series 2 Test Results* j I
- WCAP-13742 (P2) and WCAP-13743 (P3), 'lleavy Water Reactor Facility Project, Phase 1 AP600 Small Scale Passive Containment Cooling System Test ' Dry' Test Results Applicable to the HWRF Project"
- WCAP-13725 (P2) and WCAP-13726 (P3), " Heavy Water Facility (HWRF) Large Scale Passive Containment Cooling System Baseline Test Data Report
- SSAtt Revision: NONE l
l l l 280.4-1 3 WB5tingh0USB l l l
NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 410.98 1 For Q410.95-Q410.104, demonstrate how the A P600 design meets applicable GDCs by providing failure modes and elTects analyses and other requested details, as identified in applicable SRP section(s) review methodology. ; l Section 9.4.2 of the SSAR, " Annex / Auxiliary Buildings Non-Radioactive HVAC System (VXS)," predominantly falls under the review guidelines of Section 9.4.3 of the SRP, " Auxiliary and Radwaste Area Ventilation System." 1 WCAP-13053 states that the VXS conforms to Position C.2 of R.G.1.29 and Positions C.1 and C.2 of R.G.1.140. Demonstrate how the VXS conforms with the guidelines of(1) Position C.2 of R.G.1.29 for the nonsafety-related portions of the system (to show how it meets GDC 2), and (2) Positions C.1 and C.2 of R.G.1.140 (to show how it meets GDC 60). Provide justification for taking an exception to Position C.1 of R.G.1.29 for the VXS since it also serves main steam components, electrical penetration and switch gear rooms, reactor trip switchgear rooms, M/G set room battery rooms, and battery charger rooms. Providejustification for the provisions in the VXS design of only one 100-percent-capacity air handling unit for the general area HVAC subsystem and two 50-percent-capacity air handling units for the equipment room HVAC subsystem and main steam compartment versus redundant capacity units to satisfy the single failure criteria. Identify what pressurization level is maintained in the Armex I building with respect to adjacent buildings, and provide corresponding fresh-air make-up flow rates to provide this pressurization level. Provide an equipment operability evaluation for all of the MSIV compartment equipment to demonstrate that this equipment can withstand a 104*F temperature environment. Address the habitability concems inside the MSIV compartment for this elevated temperature.
Response
The VXS is designed as a non-safety-related system since it serves no safety-related function and has no nuclear safety design basis. No safety-related equipment is located in the areas served by the general area HVAC and equipment room HVAC subsystems. The MSIV compartment HVAC subsystem is not required to support the operation of any safety-related equipment. The VXS is not required to support the functioning or operation of any equipment or systems listed in Position C.1 of Regulatory Guide 1.29. Therefore, Position C.1 of Regulatory Guide 1.29 is not applicable to the VXS. Compliance to Position C.2 of Regulatory Guide 1.129 is achieved since the VXS is not required to remain functional, and its failure by an SSE will not reduce the functioning of any plant feature included in items 1.a through 1.q of Regulatory Guide 1.129 Position C.1 to an unacceptable safety level. Portions of the VXS system located in areas containing safety-related components will be seismically supported in accordance with Regulatory Guide 1.29 Position C.2. The general area HVAC subsystem serves office, locker, and toilet areas of the annex 1 building and maintains the areas for personnel comfort only. Loss of building HVAC would result in an increased building temperature during the cooling season, but the building would not become uninhabitable since the only major building heat loads are "8" W westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION
.m !!!!!
n lighting and personnel. Building usage and occupancy are not essential. HVAC system maintenance could be scheduled during evening hours and the winter to minimize any effects from HVAC system shutdown. Therefore, the use of a single,100 percent HVAC unit is adequate for the intended system functional requirements and building usage requirements. The design of the MSIV compartment heating and cooling subsystem has been revised to utilize two 70 percent capacity air handling units. During normal plant operation, both AHUs will operate to maintain an area temperature of 104"F. Each AHU has sufficient capacity to maintain both main steam compartments at or below 122"F during normal plant operation with one AHU out of service. Safety-related equipment in the compartments will be designed for continuous operation at 122*F. The area is not normally occupied. During plant shutdown, one AHU can maintain the design temperature of 104*F in both compartments. The use of two 70 percent capacity AHUs provides adequate redundancy for system operation since one AHU can maintain the area at a temperature level that will support required equipment operation and required personnel access. The VXS equipment room HVAC subsystem has been revised to provide 100 percent redundancy for the air handling units and return / exhaust fans. Two 100 percent capacity air handling units and return / exhaust fans are provided for the switchgear rooms in the annex I building. Two 100 percent capacity air handling units and return / exhaust fans are provided for the remainder of the equipment rooms. If offsite ac power is lost, the equipment rooms will be cooled utilizing a once-through ventilation mode using outside air. In this mode of . operation, the rooms are maintained at or below 122*F. Equipment in the rooms designed to operate during a loss of offsite power is designed for continuous operation at this temperature. The annex I building is designed to be maintained at a slight positive pressure relative to adjacent areas by means of providing excess supply air to the building by the general area HVAC subsystem. Approximately 500 cfm of excess supply air will be provided. No specific pressurization level is intended to be maintained. Adjacent areas are maintained at negative pressures. The main steam compartments, designed for 104*F, are not normally occupied and will contain equipment designed for that ambient temperature. This design temperature is consistent with normal design practice for areas with frequent inspection / maintenance where there is no sensitive electronic equipment. SSAR Subsection 9. 4.2 will be revised as follows: SSAR Revision: Figures 9.4.2-1,9.4.2-2, and 9.4.2-3 will be revised as shown in the attached preliminary revision. (Subsection 9.4.2.2 ) The annex / auxiliary buildmg; nonradioactive HVAC system is shown in Figures 9.4.2.-1,9.4.2-2 and 9.4.2-3. 410.98-2 W-WOSilflghattse
NRC REQUEST FOR ADDITIONAL INFORMATION The system consists of the following three independent subsystems: 1- General area HVAC subsystem - Figure 9.4.2-1 2- Switchgear and equipment room HVAC sut> system - Figure 9.4.2-2 3- MSIV compartment HVAC subsystem - Figure 9.4.2-3 (Subsection 9.4.2.2.2 ) 9.4.2.2.2 Swi.tchgear and Equipment Floom HVAC System The switchgear and equipment room HVAC subsystem serves electrical and mechanical equipment rooms in the annex ! and auxiliary buildings. This subsystemEdivided int 6 two independent HVAC'systeiss, Lone:Whirig ths electrical'switchgear rooms arid one serving the remainder of the equipment rooms. ';The switchgear room HVAC system consists of two 100-percent-capacity supply air handling units, two 100-percent-capacity retum/ exhaust fans, a ducted supply and return air system, and automatic controls and accessories.i The. equipment room HVAC system consists of two 100-percent-capacity supply air handling units, two100-percent-capacity return air fans, two battery room exhaust fans. a ducted supply and return air system, and automatic controls and accessories. Each air handling unit contains a mixing section with automatic outdoor air and return air volume dampers, high efficiency filters, hot water heating coils, chilled water cooling coils, and a centrifugal fan. He air handling units are located in the mechanical equipment room in the annex I building at elevation 135'-3". Thpif handligtdiits for each(subsystem discharge into a common duct distribution system that is routed through the building to the various areas served. Air is returned to the air handling units from the rooms served (except the battery rooms) by a shared return duct system. The temperature of the air supplied by each air handling unit is maintained at 60'F by individual temperature controls with their sensors located on the air handling unit discharge duct. When outdoor air temperatures are below 60'F, each temperature controller modulates the mixing dampers on the inlet ofits air handling unit to mix return air and outside air in the proper proportion to maintain a mixed air temperature of 60*F. When outdoor temperatures are above 60*F, the mixing dampers automatically reposition for minimum outside air, and the temperature controller modulates the chilled water and hot water control valves to maintain the supply air at 60*F. Electric reheat coils are provided in the ductwork to areas requiring close temperature control, such as the non-class IE battery rooms.
~
In~ the event 'of a loss ~of offsite power, thiair handling unitfand retum/ exhaust fans se connected to thistandby power system to provide cooling to the diesel bus switchgear and de switchgear and inverters? This cooling permits the switchgear and; inverters to perform'their defense-in-depth functions in; support'of; standby power system operationc in'this mode of operation,' the switchgear and equipment rooms arel cooled uiill2ing once-through ventilation using outdoor air. During the once through ventilation mode, the switchgear and equipment rooms will be maintained at or belowJ122*F. ' EquipmentLin these areas that operates following a loss of offsite power is designed for continuous operation at this temperature. Each non-class 1E battery room is provided with an exhaust system to prevent the buildup of hydrogen gas in the room. Each exhaust system consists of an exhaust fan, an exhaust air duct, and a gravity back draft damper located in the fan discharge. Air supplied to the battery rooms by the supply air handling units is exhausted to atmosphere. W Westinghouse
I NRC REQUEST FOR ADDITIONAL INFORMATION 1 400 un l i The battery room exhaust fans can be connected to the standby power system for investment protection in the event of a loss of offsite power. (Subsection 9.4.2.2.3 ) The main steam and feedwater lines between the turbine building and the containment are routed through a separate compartment in the auxiliary building. This compartment is provided with a separate heating and cooling subsystem. The MSIV compartment heating and cooling subsystem consists of two 70-percent-capacity supply air handling units with ductsd supply air distribution, automatic controls, and accessories. Each unit contains low-efficiency filters, hot water heating coil, chilled water cooling coil, and a centrifugal fan. The units are located directly within the space served. The temperature of the MSIV area is maintained at or less than 104*F by a space thermostat that operates the chilled water control valve serving each unit. Thesultists sisd
~
so 'that a sitigle unit 'can rnaintain both M51V 'compartrusnRat'or belove122*K] This permit @ne tmit ta be take'n temporarily out of service for.sepair or main,tenance. During a plant outage the compartment is maintained at a minimum temperature of 50*F. The air handling units can be connected to the standby power system, for investment protection, in the event of a loss of offsite power. ( Table 9.4.2.-1) Switchgear and Equipment Area Normal O fieration Non-class lE battery rooms (annex building) 75* - 79'F Non-class IE battery charger rooms (annex building) 50* - 104*F HVAC equipment room (annex building) 50 * - 104
- F Electrical switchgear rooms, motor generator set room (annex building) 50* - 104*F Non-safety electrical penetration rooms (auxiliary building) 50' - 104*F Reactor trip SWGR rooms (auxiliary building) 50' - 104*F MSIV compartments (auxiliary building) 50* - 104*F Upset Conditions.(LOOP)
Electrical Switchgear Rooms ~ 122*F(maximum) Battery Charge Rooms - '122*F(maximum) I i W-Westinghouse l i l
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- 9. AUX 1LIARY SYSTEMS R ~ vision: 1 (Pr:liminary)
Effective: 05/19/93 4
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NRC REQUEST FOR ADDITIONAL INFORMATION pi idii Question 420.9 Section 7.1.6 of the SSAR references IEEE Standards 279, 384, 603, and 796 for the design of the AP600 instrumentation and control systems. These standards are hardware-related. There is no reference to software-related standards. Provide a discussion on conformance of the AP600 design to each of the following standards:
- a. IEEE Standard C63.12-1987 (ANSI), "American National Standard for Electromagnetic Compatibility Limits-Recommended Practice."
- b. MIL-STD-461C (1987),
- Electromagnetic Emission and Susceptibility Requirements for the Control of Electromagnetic Interference."
- c. MIL-STD-462 (1967),
- Measurement of Electromagnetic Interference Characteristics."
- d. IEC Publication 801-2, " Electromagnetic Compatibility for Industrial-Process Measurement and Control Equipment, Part 2: Electrostatic Discharge Requirements."
- e. IEEE Standard C62.1-1984, on impulse voltage.
- f. IEEE Standard C62.41-1980 (ANSI), " Guide for Surge Voltages in Low-Voltage AC Power Circuits."
- p. IEEE Standard C.62.45-2987 (ANSI), ' Guide on Surge Testing for Equipment Connected to low-Voltage AC Power Circuits."
- h. IEEE Standard 587, on surge protection.
- i. IEEE Standard 730.1-1989 (ANSI), ' Software Ouality Assurance Plans."
- j. IEEE Standard 828-1990, " Software Configuration Management Plans."
- k. IEEE Standard 829-1991, " Software Test Documentation."
- 1. IEEE Standard 830-1984 (ANSI), " Software Requirements Specifications."
- m. lEEE Standard 983-1986 (ANSI), " Software Quality Assurance Planning."
- n. IEEE Standard 1012-1986 (ANSI), " Software Verification and Validation Plans."
- o. IEEE Standard 1016.1-1987 (ANSI), " Software Design Descriptions."
- p. IEEE Standard 1028-1988 (ANSI),
- Software Resiews and Audits.*
l 420 m W westinghouse 1
NRC REQUEST FOR ADDITIONAL INFORMATION lis! !!ni
- q. IEEE Standard 1042-1987 (ANSI), " Software Configuration Management."
- r. IEC (International Electrotechnical C<munission) 880-1986, " Software for Computers in the Safety Systems of Nuclear Power Stations."
- s. ANSI ASC X3T9.5-1988, ' Fiber Distributed Data Interface (FDDI)."
- t. IEEE Standard 802.5-1985, " Token Ring Access Method and Physical Layer Specifications."
- u. ISO 7498-1984, "Open System Interconnection - Basic Reference Model."
- v. IEEE Standard 802.2-1985, " Standard for local Area Networks: logical Link Control."
Response
There are no regulatory requirements that invoke the codes and standards listed in this question. These codes and standards provide a means to promulgate the collective industry experience with designing electronic equipment to minimize the effects of electromagnetic interference (EMI), radio frequency interference (RFI), electrical surges, and electrostatic discharges, and with designing computer hardware and software to minimize human design and coding errors and to enhance interoperability. The standards and codes applicable to instrumentation and control equipment are identified in the system, hardware, and software design documents described in WCAP-13392 (NP), "AP600 Instrumentation and Controlliardware and Software Design, Verification, and Validation Process Report,* for specific instrumentation and control systems. Instrumentation and control equipment function and operation, in accordance with regulatory requirements and guidance, will be proven by the design, verification, and validation program, functional testing, and qualification testing, as appropriate. SSAR Revison: NONE 420.9-2 W westinghause
NRC REQUEST FOR ADDITIONAL INFORMATION iHi! liu W ii Question 420.1o Electromagnetic interference (EMI)/ radio frequency interference (RFI), including surge and electrostatic discharge, is an issue applicable to safety-related digital system design. Provide a discussion describing how the AP600 design conforms to the following standards and guidance (The response need not be limited to these items.)(Section 1.8):
- a. MIL-STD-461(A,B,C), " Electromagnetic Emission and Susceptibility Requirements for the Control of Electro-me.gnetic Interference."
- b. MIL-ri D-462,
- Electromagnetic Interference Characteristics Measurement."
- c. MIL-STD-1399,
- Interface Standard for Shipboard Systems, DC Magnetic Field Environment."
- d. SAMA PMC 33.1-1978, " Electromagnetic Susceptibility of Process Control Instrumentation."
- e. NCR Information Notice IN 83-83, "Use of Portable Radio Transmitters Inside Nuclear Power Plants."
- f. NUREG CR-3270,
- Investigation of Electromagnetic Interference (EMI) Levels in Commercial Nuclear Power Plants."
- g. ANSI /IEEE Standard C37.90.1-1989, "IEEE Standard Surge Withstand Capabihty Tests for Protective Relays and Relay Systems."
- h. ANSl/IEEE Standard C37.90.2-1987, "IEEE Trail Use Standard Withstand Capability of Relay systems to Related Electromagnetic Interference from Transceivers.*
- i. IEC Standard 801-1, " Electromagnetic Compatibility for Industrial-Process Measurement and Control Equipment-General Introduction."
- j. IEC Standard 801-3,
- Electromagnetic Compatibility for Industrial-Process Measurement and Control Equipment-Radiated Electromagnetic Field Measurement."
- k. IEC Standard 801-4, " Electromagnetic Compatibility for Industrial-Process Measurement and Control Equipment - Electrical Fast Transient / Burst Requirements."
- 1. IEEE Standard 1050-1989, *IEEE Guide for Instrumentation and Control Equipment Grounding in Generating Station."
- m. IEEE Standard 572-1985, "lEEE Standard for Qualification of Class H: runnection Assemblies for Nuclear Power Generating Stations.*
- n. IEEE Standard 518-1982, "IEEE Guide for the Installation of Electrical Equipment to Minimize Electrical Noise Inputs to Controllers from External Sources."
420.10-1 W Westinghnuse
NRC REQUEST FOR ADDITIONAL INFORMATION
!!i tin M jg
- o. IEC Standard 801-2, "EMC for Industrial-Process Measurement and Control Equipment, Part 2:
Electrostatic Discharge Requirements."
Response
There are no regulatory requirements that invoke the codes and standards listed in this RAI. These codes and standards provide a means to promulgate the collective industry experience with the determination of the effects of electromagnetic interference (EMI), radio frequency interference (RFl), electrical surges, and electrostatic discharges on electronic equipment. The standards and codes applicable to instrumentation and control equipment are identified in the system, hardware, and software design documents described in WCAP-13392 (NP), *AP600 Instrumentation and Control Hardware and Software Design, Verification, and Validation Process Report," for specific instrumentation and control systems. Instrumentation and control equipment function and operation, in accordance with regulatory requirements and guidance, will be proven by the design, verification, and validation program, functional testing, and qualification testing, as appropriate. SSAR Revision: NONE 420.10-2 3 Westinghouse
NRC REQUEST FOR ADDITIONALINFORMATION Ein
'I Ouestion 420.19 An assessment ofIEEE Standard 796-1983 "lEEE Microprocessor System Bus," was performed by the Lawrence Livermore National Laboratory (LLNL) for the NRC. A draft technical report is attached with this request for additional mformation (RAI). Address the following concems that are raised by the LLNL repon (Section 7.1.2):
- a. The bus design described by lEEE Standard 796-1983 is based on an 18 year old design standard, its computational capability is limited (1-4 million bus cycles per second). Replacement parts availability is also a concern.
- b. The bus design described by IEEE Standard 796-1983 is sensitive to external noise and should be used in well-shielded enclosures with well-shielded power supplies only,
- c. Applications should be certified only for specific computer or computer-related PCBs located in specific slots. Addition or relocation of PCBs in board cage slots will require recertification.
- d. The bus design described by IEEE Standard 796-1983 has minimum support for multiprocessing applications, and the software bears the burden for ensuring correct system synchronization. This requires highly skilled programmers and is difficult to do conectly. Software errors in this area may result in common-mode failures extending over multiple systems.
- e. The bus width and address space limitation posed by IEEE Standard 796-1983 cause additional software complexity.
- f. The bus described by IEEE Standard 796-1983 starts from the least-significant byte first. Bit ordering would be a concern if processors of different bit order were intermixed on the bus.
- g. Potential for conflict exists if PCBs are configured for incompatible bus master exchanges. If a bus priority arbiter is used, it represents a single point of failure for the system.
- h. There are three opportunities for PCB configuration error in interrupt sequence settings.
- i. The lack of transmission line termination in the IEEE Standard 796 bus specification contributes to a delay in settling time.
- j. The bus described by IEEE Standard 796-1983 uses obsolete integrated circuit technology (TTL) and was designed at a time when transmissico line theory was not being applied to microprocessor bus design.
- k. The bus described by IEEE Standard 796-1983 uses two obsolete connectors. Edge connectors are more susceptible to contamination and mis-insertion than the two-part connectors for modern buses.
- 1. The bus described by IEEE Standard 796-1983 has no parity or error correction.
W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION
Response
1 The IEEE Standard 796-1983 bus was selected by Westinghouse for microprocessor-based products because ofits history of successful use in industrial applications. It met our criterion for a proven technology. It is supported by numerous vendors, which ensures the availability of compatible products well into the future. Westinghouse has used this technology to configure a wide range of systems from small stand-alone applications in the process and fossil plant industries to complete I&C systems for nuclear plants. These systems are configured using multiple IEEE Standard 796-1983 bus-based subsystems operating in parallel without synchronization, and communicating with one another using data links and data highways. WestinghouseIEEE Standard 7961983 bus-based systems have been subjected to seismic, environmental, and EMI/RFI testing compatible with nuclear Clast 1E applications. The , supporting software has completed extensive verification and validation testing. The hardware and software test programs have been approved by both the USNRC and international regulators and the results confirmed by independent testing in the case of the British Sizewell application. The assessment of IEEE Standard 796-1983, "lEEE Microprocessor System Bus," performed by the Lawrence Livermore National Laboratory (LLNL) states: "There is no reason to believe that current plans for the AP600 protection system cannot be accommodated within the processor and address space limitations posed by IEEE 796-198.3 bus width.* Responses to the specific items from Section 7.1.2 of the LLNL report are as follows:
- a. De bus design desaibed by IEEE Standard 796-1983 is based on an 18 year old design standard. Its computational capability is limited (1-4 million bus cycles per second). Replacement pans availability is also a concern.
The IEEE Standard 796-1983 bus is a proven technology. Its capability is adequate for the intended purpose. Westinghouse is not aware of any specific parts availability problems.
- b. De bus design described by lEEE Standard 796-1983 is sensitive to external noise and should be used in well-shicided enclosures with well-shielded power supplies only. ,
The Westinghouse implementations use well-shielded enclosures with well-shielded power supplies. Tests have been performed to ensure that the systems meet their specifications. See the response to Q420.56.
- c. Applications should be centped onlyfor specific computer or computer-related PCBs located in specifc '
slots. Addition or reforadon of PCBs in board cage slots will require recertifcation. IEEE Standard 796-1983 makes no restriction on kication of cards in specific slots,
- d. ne bus design deswibed by IEEE Standard 796-1983 has minimum supportfor multiprocessing applications, and the software bears the burden for ensuring correct system syndsronization. His 420.19-2 W Westinghotise
NRC REQUEST FOR ADDITIONAL INFORMATION 7 $$ !11 requires highly skilled programmers and is dificult to do correctly. Software errors in this area may result in common-modefailures extending owr multiple systems. Westinghouse does not use the IEEE Standard 796-1983 bus in multiprocessing applications and does not synchronize the operations of the different processors.
- e. 1he bus width and address space limitationposed by IEEE Standard 796-1983 cause additionalsoftware wmplexity.
Wcstinghouse is not aware of any limitation beyond that normally experienced during the design of any microprocessor-based system. Our design engineers are familiar with the characteristics of the bus and the supported microprocessors.
- f. 1he bus described by IEEE Standard 796-1983 startsfrom the least-signtficant bytefirst. Bit ordering would be a concern ofprocessors ofdiferent bit order were intmnized on the bus.
Westinghouse design engineers are acquainted with the bit ordering of Intel products and other products with which they interface. Processors of different bit order do not share the bus.
- g. Potentialfor conflict exists if PCBs are configuredfor incompatible buz master exchanges. If a bus priority arbiter is used, it represents a single point offailurefor the system.
Westinghouse does not use the IEEE Standard 796-1983 hus in a multimaster configuration. There is no bus priority arbiter. The postulated failure does not constitute a single point of system failure in Westinghouse designs since the system is configured using multiple, separate computer subsystems within a safety division. The divisions are, in turn, configured with a four-way redundancy.
- h. There are three opportunitiesfor PCB wnfiguration error in interrupt sequence settings.
Westinghouse design engineers are familiar with the use of the interrupt sequence settings on the circuit boards. In addition, it is our practice to avoid the use of interrupts in our systems, further minimizing the likelihood of incorrect settings.
- i. 1he lack of transmission line termination in the IEEE Standard 796 bus specification contributes to a delay in settling time.
Westinghouse design engineers are familiar with the settling time characteristics of the IEEE Standard 796-1983 bus and account for it in the design. J. 7he bus described by IEEE Standard 796-1983 uses obsolete integrated circuit technology (17L) and was designed at a time when transmission line theory was not being applied to micropronssor bus design. 420.wa W wesunsouse
NRC REQUEST FOR ADDITIONAL INFORMATION l E The computational capability of the IEEE Standard 796-1983 bus is adequate for the needs of the systems in which it is used.
- k. The bus described by IEEE Standard 796-1983 uses tne obsolete conneaors. Edge connectors are more susceptible to contamination and mis-insertion than the tav-part connectorsfor modern buses.
The connectors defined by IEEE Standard 796-1983 are adequate for their intended purpose. Westinghouse is not aware of any problem in obtaining IEEE Standard 796-1983 connectors. Numerous computer subsystems configured with these connectors are in operation. Systems configured with the IEEE Standard 796-1983 connectors have been both seismically and environmentally qualified. Many systems in power plant and processing plant environments have been in operation for over 10 years with no evidence of this problem.
- l. 1he bus described by lEEE Standard 796-1983 has no parity or error correction.
Westinghouse design engineers are familiar with the lack of error correction on the IEEE Standard 796-1983 bus. The software has been designed to address this characteristic. SSAR Revision: NONE 420.19-4 3 Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION 92 !!!!! L !!! Question 420.36 Identify and justify all the protection system actuated equipment that are not tested during reactor operation in accordance with the guidelines of Regulatory Guide 1.22 (Section 7.1.2.12).
Response
The response to Q210.24 describes the inservice test strategies for the AP600 safety-related equipment. This response includes all equipment actuated by the protection system and identifies any of this equipment that is not tested in accordance with Regulatory Guide 1.22 (Section 7.1.2.12). SSAR Revision: NONE i 420.36-1 j W
=
Westinghouse , l
NRC REQUEST FOR ADDITIONAL INFORMATION n!!! "i" ! y? _ t Ouestion 420.39 Clarify whether the only interface between the integrated protection system and the integrated control system is through the integrated control cabinets signal selector subsystem. If this is not the case, identify and describe other interface methods. (Section 7.1.3.1.1)
Response
The signal selector subsystem is the main interface for plant data that originates in the integrated prc.::etion cabinets and used by the plant control syst.m. There is also a direct interface, used for the safety-related rod withdrawal black shown in Sheet 4 of Figure 7.2-1 of the AP600 SSAR, between the integrated protection cabinets and the rod control cabinets. This interface uses an isolated contact closure signal. Isolated sensor signals are provided from the protection and safety monitoring system to the diverse actuation system as discussed in SSAR Subsection 7.7.1.11. SSAR Revision: NONE W Westinghouse
l \ l 1 NRC REQUEST FOR ADDITIONAL INFORMATION i l l Question 420.44 l Describe the detailed design for transmission of post-accident monitoring information from the protection system to the main control room. Provide failure modes and effects analyses to demonstrate that a postulated single failure , (such as data link component) will not disable the post-accident monitoring information. (Section 7.1.4.2.19) l
!1 Response: )
1 A description of the qualified data processing system (QDPS) is included in WCAP-13391(NP), "AP600 l Instrumentation and Control liardware Description,* Rev. O, May 15,1992 (Reference 3 of SSAR Subsection 7.1.6) 1 as Section 4.7. 'Ihe qualified data processing system is part of the protection and safety monitoring system and ! uses the same hardware and software building blocks as the integrated protection cabinets. A failure modes and effects analysis will be performed using the methods of WCAP-13662(NP), " Advanced Passive l Plant Protection System FMEA," after further detailed design on the QDPS is completed. The results of this FMEA will be incorporated into a future revision of this WCAP. The detailed design description and the updated FMEA will be submitted to the NRC by December 15, 1993. j SSAR Revision: NONE l l l l 1 3 Westinghouse l l l
NRC REQUEST FOR ADDITIONAL INFORMATION nin r y;i Ouestion 420.52 Describe the extreme environmental and energy supply conditions that the AP600 protection and safety monitoring system is designed to withstand. Rese conditions should include but not be limited to (Section 2.3 of WCAP-13382):
- maximum and minimum temperature range
- maximum humidity range a maximum and minimum power supply voltage and frequency
- totalloss of HVAC
- smoke or fire in the area
- the environment during a station blackout event
- EMI/RFI, surge and electrostatic discharge e seismic vibration
Response
- Maximum and minimum temperature range The protection and safety monitoring system equipment is designed for normal operating temperature limits of 41' to 104*F.
The protection and safety monitoring system equipment is designed for abnormal operating temperature limits as shown in SSAR Figure 3D.5-1.
- Maximum humidity range The protection and safety monitoring equipment is designed for a maximum relative humidity range as listed in SSAR Table 3D.5-4 and shown in Figure 3D.5-1.
- Maximum and minimum power supply voltage and frequency !
ne protection and safety monitoring cabinets are provided with ac power between 112 and 125 Vac and between 57 and 63 Hz. l
- Total loss of IIVAC l 1
Passive heat sinks are sufficient for 72 hours of required cabinet operation if normal HVAC is lost. l 42032a W westinghouse
NRC REQUEST FOR ADDITIONALINFORTAATION W L ili i
- Smoke or fire in the area The protection and safety monitoring system is designed to tolerate the loss of a single division due to smoke and fire. The HVAC is designed to prevent propagation of fire or smoke from the affccted division to the other divisions. Smoke purge capability is provided by the nuclear island nonradioactive ventilation system (VBS)
HVAC system, discussed in SSAR Section 6.4.
- The environment during a station blackout event During a station blackout event, the passive HVAC features operate. The temperature rises are shown in SSAR Figure 3D.5-1.
- EMI/RFI, surge and electrostatic discharge EMI/RFI requirements are addressed in the response to Q420.59.
The protection and safety monitoring system is designed to pass surge withstand requirements as defm' ed in IEEE Standard 472-1974 (ANSI C37.90.1-1989), except for nuclear instrumentation inputs. Nuclear instrumentation inputs are designed to withstand a surge of 1000 feet of triaxial cable charged to the maximum detector voltage and current and to withstand an overload condition of 10 times the channel nominal full-range signal. Since the cabinets are grounded and normally closed, electrostatic discharge is a maintenance issue. Grounding straps are provided for maintenance personnel. Maintenance instructions include appropriate precautions.
- Seismic vibration The safety-related instrumentation and control cabinets will be qualified to the seismic response consistent with their building location.
SSAR Revison: NONE 420.52-2 W WBStingh0llSS l i
NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 420.53 All instrumentation and control cabinet outputs will de-energize during a loss of power event. Provide a failure modes and effects analysis (FMEA) for the integrated protection cabinets power supply arrangement inside the Nuclear Island / Annex building. (Section 7.1.2.10 of the SSAR and Section 2.9 of WCAP-13382)
Response
Power to the protectier. and safety monitoring system is supplied by the Class IE de and UPS system. A failure modes and effects analysis for the Class 1E de and UPS system is provided in Table 8.3.2-7. SSAR Revision: NONE hY WB5tlDgh0USS l
NRC REQUEST FOR ADDITIONAL INFORMATION A Question 420.86 The AP600 alarm system includes 5000 input alarms, and 40 support workstations. Describe the software design that can achieve a 2-second time response even during the avalanche of alarms during an upset condition. (Section 18.9.2.4.16)
Response
The current application software sequentially processes all 3000 points in one second. Since all points are processed every second, there is no difference in system response whether the alarm activity is high or low, i.e., there is no avalanche condition as far as the system is concerned. The system will process 5000 points in less than two seconds. SSAR Revision: NONE r 420.86-1 W- WestinEhouse
NRC REQUEST FOR ADDITIONALINFORMATION r* Ein Response Revision 1 Question 450.2 Section 6.4 of the SSAR states that the VES is capable of providing emergency ventilation and pressurization of the MCR. Also, it states that the VES is sufficient to maintain a 1/8-inch water gauge positive pressure differential with respect to the adjacent areas, preventing infiltration of containment air into the main control room envelope while the VES is in operation. Table 15.6.5-2 identifies 0.3 cfm unfiltered air inleakage from ingress and egress. The staff considers a 0.3 cfm unfiltered inleakage for the entire control room envelope unrealistic as judged from the experience of existing operating plants to date. Reassess the unfiltered infiltration inside the control room envelope and provide a credible infiltration inleakage. This should be supported by an approved methodology for calculation and should be able to be tested and verified periodically. Also, provide (1) a list of areas considered part of main control room envelope, (2) the entire MCR envelope volume, (3) the expected revised unfiltered infiltration rate in the entire MCR envelope, and (4) value credited for the entire MCR envelope infiltration rate in accident dose calculations. Explain in detail how the MCR envelope is isolated during accident conditions in order that it does not exceed the revised value of the unfiltered infiltration rate used in accident dose calculations. Identify the permanent measures to be implemented, including sealing the MCR envelope and periodic verification and testing provisions. If sealants are used, demonstrate their acceptability and qualification to maintain needed isolation through the proposed design plant life. Response (Revision 1): The responses to the numerous questions within 450.2 are provided in the following paragraphs. QUESTION: Table 15.6.5-2 identifies 0.3-cfm unfiltered air inleakage from ingress and egress. The staff considers a 0.3-cfm unfiltered inleakage for the entire control room envelope unrealistic as judged from the experience of existing operating plants to date. Reassess the unfiltered infiltration inside the control room envelope and provide a credible infiltration inleakage. This should be supported by an approved methodology for calculation and should be able to be tested and verified periodically. RESPONSE: During emergency operation, the unfiltered inleakage for current plants is made up of two components: damper leakage and ingress / egress. For the AP600, the makeup air is provided by the compressed air storage tanks. Therefore, as long as the MCR envelope is maintained at a 1/8-inch WG positive pressure, the only source of unfiltered inleakage is through the vestibule doors due to ingress / egress. The traditional value for single-door inleakage is 10 cfm or greater. This value can be reduced significantly er uminatal by using a double-door vestibule. Although the initial estimate of potentialleakage rate was 0.3 cfm for the vestibule door configuration, a more detailed estimate ofinfiltraion due to personnel ingress / egress ' -
- deu in ;vogw+ is presented in the following paragraphs. h-methe&'rgy, nu =p 'm, -d r.=!!n of 'h:
andp :" N pn i&J b 'he NRC $ ".fvil 'O. '993. W-Westinghouse
NRC REQUEST FOR ADDITIONAL.INFORMATION Response Revision 1 QUES 770N: Provide the following: (1) a list of areas coraidered part of the MCR envelope (2) the entire MCR envelope volume (3) the expected revised unfiltered infiltration rate in the entire MCR envelope (4) the value credited for the entire MCR envelope infiltration rate in accident dose calculations
RESPONSE
(1) The main control area, offices, kitchen, and restroom krated on the nuclear island. (2) 42,260 ft3 (value includes the reduction in volume due to equipment, furniture, etc.) (3)-OJ+fm (ir" "' v:!u: ;=ent!y undc .v ! Aen = d!= :cJ ab' v:),0.24 cfm; += pena de ^ p? 30,1993 . the basis for this value is provided in the following paragraphs. (4) Unda :=!=:r, = di=und ab ., ::spn;= J= ^.pr!! 'O, '993 See the following paragraphs. The following provides additional information regarding the methodology and assumptions utilized in calculating MCR unfiltered air inleakage and in evaluating MCR radiological consequences. The AP600 MCR envelope is maintained at least at a 1/8-inch WG positive pressure; therefore, the only path avallable for unfiltered air inleakage is through the MCR vestibule doors from personnel ingress / egress. He AP600 MCR design includes a double-door vestibule, which significantly reduces the unfiltered air inleakage rate. The following paragraphs provide the methodology and assumptions utilized in calculating the MCR unfiltered air inleakage rate. De MCR unfiltered air inleakage rate is calculated from the number of vestibule door openings that occur during the accident. An assumption is that there will be no vestibule door openings (i.e., no special provisions made for emergency response personnel entering the MCR, or control room operators leaving the MCR during the accident) exceps at shift changes. He MCR inleakage rate is influenced by the amount of air assumed to be exchanged with the MCR air during each vestibule door opening._ The following assumptions are utilized in calculating MCR unfiltered air inleakage:
- Vestibule door openings occur at shift changes.
- First shift change occurs at 4 hours (after the beginning of the accident), with a change of shift every 12 hours thereafter.
Shift duration is 12 hours (does not include shift tumover time).
- Debriefing by MCR staff during shift changeover is 1/2 hour.
- When changing shifts, the MCR staff enters / exits the MCR vestibule together.
The personnel ingress / egress process is strictly administratively controlled. The MCR ingress / egress sequence is as follows. For MCR ingreu the outer MCR vestibule door is opened, the entire staff enicrs the vestibule, the outer vestibule door is closed, the inner vestibule door is opened, and the staff enters the MCR. During staff debriefing, the MCR inner vestibule door is maintained in the open position. The vestibule volume becomes part of the MCR I 450.2(R1)-2 i i i
NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 envek>pe, precluding unfiltered air inleakage into the vestibule. For MCR egress the entire staff enters the open vestibule, the inner vestibule door is closed, the outer vestibule door is opened, the staff exits, and the outer vestibule door is closed. This process occurs with each shift change. Assuming a full vestibule air volume exchange with each vestibule door opening sequence, the average unfiltered air inleakage rate to the MCR, fo. the duration of the accident, is 0.24 cfm. He total unfiltered air volume introduced into the MCR with each shift change is 174.0 ft .3 The AP600 MCR radiological consequences have been analyzed based on an MCR unfiltered air inleakage rate of 0.3 cfm (SSAR Table 15.6.5-2); therefore, the calculated rate of 0.24 cfm is bounded by the current analyses. See the response to Q450.1 for the normal MCR inleakage through wall, penetrations, ducts, etc. QUESTION: Explain in detail how the MCR envelope is isolated during accident conditions so that it does not exceed the revised value of the unfiltered infiltration rate used in the accident dose calculations. RESPONSE: Upon receipt of a Ili-2 MCR radiation signal, the isolation dampers for the VBS receive a signal to close, thereby isolating the non-safety-related portions of the VBS outside the MCR from the MCR envelope. The VBS isolation dampers are part of the safety-related, .eismic Category 1 portion of the VBS ductwork that penetrates the MCR envelope. These redundant, bubble-tight dampers have spring-return actuators that fail-safe on the loss of electrical power. They are constructed, qualified, and tested in accordance with AMCA 500 and ASME N509, Section 5.9. Upon receipt of the same Hi-2 MCR radiation signal, the VES isolation valves receive a signal to open. The VES is made up of two completely redundant trains of compressed air tanks. Opening of either isolation valve provides 100 percent of the air flow required to pressurize the MCR envelope to at least 1/8 inch WG with respect to the surrounding areas. Under these conditions, the only path available for unfiltered inleakage is the MCR vestibule doors due to ingress / egress. QUESTION: Identify the permanent measures to be implemented, including sealing the MCR envelope and periodic verification and testing provisions. If sealants are used, demonstrate their acceptability and qualification to maintain needed isolation through the proposed design plant life. RESPONSE: The const.uction and sealing techniques of the main control room envelope are discussed in the response to Q450.1. The verification and testing provisions are provided through memm.e initial and periodic steps in& ding N identified in:
- Initial Test Program, as discussed in AP600 SSAR Subsection 14.2.8.1," Main Control Room Ventilation System *
- Technical specification requirements, as specified in SSAR Section 16.1, Item 3.7.6 3 Westirighouse
NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 SSAR Table 15.6.5-4 will be revised as follows: SSAR Revision: NONE 450.2(R1)-4 W Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 471.7 Section 12.3.1.1.1 of the SSAR. states that the steam generator manways are sized for easier entrance and exit of workers with protective clothing, and to facilitate the installs. tion and removal of tooling. Provide the size of the manways and discuss the ease of personnel entering and exiting the steam generators while wearing full face air-supplied hooded respirators.
Response
The AP600 steam generator manways are 18 inches in diameter. This is an increase of approximately 2 inches over most currently operating Westinghouse PWRs, which have 16-inch manways. The AP600 steam generator channel heads and access platforms are designed to facilitate the use of robotic devices such as ROSA III for activities including installation of nozzle dams, eddy current inspection tube plugging and sleeving. He trend in operating plants is the increased use of robotics to reduce the amount of personnel entries to perform these tasks. The AP600 is designed to use the latest robotic devices and t'so provides larger manways that will ease entry of personnel wearing full-face, air-supplied hooded respirators, should the necessity for personnel entry be required. SSAR Revision: NONE 477 W westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 471.8 Section 12.3.1.1.1 of the SSAR states that tanks are provided with overflow lines diverted to waste collection systems to collect contamination within plant structures. Describe the arrangement of the overflow lines into the waste collection system. Are they hard-piped into a sump, directed to a funnel, or directed to a floor drain channel?
Response
With the exception of the inside-containment refueling water storage tank (IRWST) and the tanks in the liquid radwaste system, overflow lines from tanks containing radioactive liquid are hard piped to a local sump. For example, the spent resin storage tanks have a hard piped connection to the sump in the radwaste building. The IRWST overflow lines are hard piped to the refueling cavity, which in turn has hard piped drains to the containment sump. This arrangement is descrital in SSAR Subsection 6.3.2. The overflow lines from tanks in the liquid radwaste system, except for the waste holdup tanks, are piped to local funnels (which serve as siphon breakers) and these funnels are piped to the waste holdup tanks. The waste holdup tanks overflow lines are interconnected such that if one tank overflows when the other has volume available, the full tank will overflow to the other tank. Ultimately both tanks overflow to the room containing them, which is water-tight and has connections to allow it to be emptied after an overflow event. SSAR Revision: NONE W-WB5 thigh 0USB
NRC REQUEST FOR ADDITIONAL INFORMATION jf tu!l h Question 471.12 Section 12.3.1.1.1 of the SSAR states that components exposed to high temperature reactor coolant contain specific limitations on the cobalt impurity content as give in Table 12.3-1 of the SS AR. The table states that for bearings and hard-facing materials, the maximum weight percent average value is approximately 60. Explain why this value is so high compared to the cobalt limitation for other components.
Response
As stated in SSAR Table 12.3-1, the maximum cobalt content of bearing and hard-facing materials is currently not limited. Cobalt based hardfacing materials such as stellite may be specified for specific applications where necessary for reliability considerations. In other applications, low or no-cobalt materials are used. The value of 60 percent is not intended as an average for all bearing and hard-facing materials in the plant. It is rather an approximate maximum cobalt content for the specific applications where stellite or other cobalt-based materials are used. SSAR Revision: NONE c1.12-1 W westinghouse
NRC REOUEST FOR ADDITIONAL INFORMATION Ouestion 471.16 Section 12.4.1.1 of the SSAR states that, based on review of current plant operations, and on AP600 design changes and reliability improvements, it is estimated that 100 worker-hours per year are required to be spent in the containment to monitor the plant during power operations. Justify this 100 work-hour estimate and describe in detail the types of operations that it covers. Describe what protective equipment will be needed (respirators, ice vests, etc.) for the containment entries, and also include person-rem estimates of these operations.
Response
The SSAR statement that "..it is estimated that 100 worker-hours per year are required .
- describes an assumption that was employed in deriving ORE estimates for the AP600 plant. There are currently no requirements for maintenance and/or inspections inside the containment when the plant is at power. The 100 worker-hours are based on a review of current operating plant practices and procedures and are considered to be a conservative estimate considering the AP600 design features and reliability improvements.
A survey of operating plant personnel indicates an extremely wide variation in the containmsat access frequency during power operation.. It ranges from a scheduled daily inspection tour, at one plant, to only when and if required (normally less than once per month) at most of the other plants that were contacted. The location, frequency, and duration of containment access at power at current operating plants are influenced by plant technical specifications, containment design, unscheduled repairs, and specific utility practices. For example, the identification of leaks from the reactor coolant pressure boundary is an important issue in satisfying the plant technical specification limits on allowable leakage. This identification process can involve access to search for leaks and, based on operating plant experience, is the most common reason for access at power. Operations and activities that have been associated with entries during power operation at operating plants and that are possible reasons for containment entry for the /. '600 include:
- Surveillance Routine patrols Health physics surveys System leak identification
- Unscheduled maintenance Repair / replace of transmitters Valve operator repairs Isolate minor system / component leaks Valve adjustments Fan cooler repairs Repair / replace radiation monitoring detectors The basis for the value assumed in the AP600 ORE assessment is access by two workers for 1/2 hour at an average frequency of twice per week (average over 50 weeks per year), for an annual total of 100 worker-hours. This 471.16-1 W-Westinghouse
I I l l NRC REQUEST FOR ADDITIONAL INFORMATION
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1 1 estimate is expected to be conservative for the AP600, since the experience at many operating plants is less than once per month. I t The estimated annual personnel exposure associated with such activities is a small fraction of the total ORE estimated for reactor operations and surveillance. That is, for an average radiation field of 15 millirem /hr, the total , containment entry personnel exposure is 1.5 man-rems, or approximately 10 percent of the 13.8 man-rems estimated for reactor operations and surveillance. Based on operation with a low level of fuel defects and minimal primary system leakage into containment, respiratory protection is not expected to be required; this has been confirmed by 1 operating plant experience. The expected temperatures at the operating deck and other accessible areas would not j require special cooling / ice vests. However, the particular protective equipment requirements, including personal l cooling equipment, will be determined by the operating planth health physics department, as predicated by the radiation levels and environmental conditions that exist in the containment. I SS AR Subsection 12.4.1.1 will be revised as follows: { SSAR Revision. When the plant is at power, the containment radiatica fields are significantly higher than at plant shutdown. The frequency and duration of at -power containment entries is dependent on the plant operator. Based on review of current plant operations and on the AP600 design changes and relie5ility improvements, it is zu::=:cd assumed that 100 worker-hours per year are equ:: d : =dniair the p!=: du-ing : pc e a rad!!!nn spent in the containment during power operations. 1 i 471.16-2 l W
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NRC REQUEST FOR ADDITIONAL INFORMATION s.h 8 Ouestion 471.17 Section 12.5.3.7 of the SSAR states that special coatings are applied to the walls and floors of areas containing radioactive fluids to aid in decontamination. Describe which areas of the plant will be treated with these coatings, and, if the entire wall surface will not be treated, describe how far up the walls these coatings will be applied.
Response
SSAR Subsection 6.1.2 provides a complete discussion of coatings inside and outside containment. SSAR Revision: NONE 471.17-1 3 WB5fitigh00se ! l
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NRC REQUEST FOR ADDITIONAL INFORMATION im !!E IF 11[ Question 630.7 Section 16.1.1 of the SSAR states that the technical specificatim (TS) LCOs are provided in accordance with the criteria of the Commission's Policy Statement on Technical Specification improvements, and are consistent with draft NUREG-1431, " Standard Technical Specifications (STS) - Westinghouse Plants," to the maximum possible extent, except where differences are justified because of design reifferences. The staff has recently completed the resolution of commenu on the improved STS that were issued to the Westinghouse Owners Group for proof and review on June 30,1992. As a result, the draft NUREG-1431 that Westinghouse used as the foundation for the AP600 application has changed substantially with the issuance of the Revision 0 STS. Therefore, identify specific differences in the proposed technical specification requirements for the AP600 from those contained in the June 30,1992 version of NUREG-1431.
Response
The screening criteria contained in the Commission's Policy Statement on Technical Specification Improvements were applied to all AP600 plant systems to identify the required technical specifications. Any applicable changes to the Westinghouse Standard Technical Specifications resulting from the ongoing industry effort will be incorporated into the AP600 technical specifications as appropriate by December 31,1993. The AP600 includes pasdve, safety-related systems and components that do not exist in current plants and are not represented in the Westinghouse Standard Technical Specifications. The functions performed by these AP600 passive, safety-related systems and components meet the screening criteria, thus requiring technical specifications. The associated AP600 passive, safety-related system and component technical specifications are based on the ; philosophy used to develop the Westinghouse Standard Technical Specifications for systems and components that perform similar functions on current plants. Some technical specifications included in the Westinghouse Standard Technical Specifications because they meet the screening criteria for current plants are not required to be included in the AP600 technical specifications. The functions performed by these systems and components do not meet the screening criteria; therefore, corresponding technical specifications are not required. A preliminary comparison of the AP600 Technical Specifications with the NRC Standard Technical Specifications in NUREG-1431 (Revision 0) has been completed. The comparison identified three categories of differences that are listed in the following table along with specific examples for clarification. [ WB5fiflgh00S8
NRC REQUEST FOR ADDITIONAL.INFORMATION ...l m!i u % _ t
- 1. Plant-specific design characteristics or terminology that affects plant operation in accordance with technical specifications
- Revised mode definition temperature limits
- Use of AP600-specific safety limits
- Use of AP600-specific moderator temperature coefficient and peaking factors
- Different completion times proposed for various AP600 components
- Change in the fuel cycle length
- Passive systems
- Safety-related functions and components of non-safety-related systems
- Different AP600 reactor trip system and engineered safety features instrumentation
- RCS total loop flow rates vs individual cold leg flow rates
- Different low temperature overpressure protection requirements
- Revision to LCO 3.0.3 treatment for safe shutdown conditions
- 2. Plant-specific design characteristics, terminology, or references that do not affect plant operation in accordance with technical specifications
- Gray rods
- No reactor coolant pump seals
- No safety-related diesel-generators
- On-line power distribution monitoring system
- Fixed incore detectors
- Use of AP600-specific references 630.7-2 W
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NRC REQUEST FOR ADDITIONALINFORMATION
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- 3. Technical, editorial, and grammatical corrections and clarifications
- Use of different wording or examples
- Wording clarification
- Reference a limit in other parts of a specification
- Discussion on the calibration of temperature sensors
- Addition of missing completion times
- Clarification of surveillance requirements by adding notes on when or how to conduct surveillance requirements
- Cross-reference to an additional LCO
- Addition of a condition for a required action
- Clarification of mode applicability
- Clarification of conditions for required actions
- Correction to table format by adding a format line
- Addition of a shutdown margin requirement for the physics test LCO
- Addition of references to core operating limits report
- Clarification on the quadrant power tilt ratio calculation
- Allowable change in battery electrolyte lindts during charging SSAR Revision: NONE 630.7-3 W Westinghouse
NRC REQUEST FOR ADDITIONALINFORMATION
!!!!: !!ili Question 830.9 Section 16.1.1 states that, in addition to excluding LCOs for systems classified as 'non-safety systems * (as this classification is used for the AP600 design), other technical specification changes are included as a result of (1) redefining completion times and surveillance frequencies using a combination of deterministic criteria and PRA evaluation; (2) specifying shutdown completion times and mode definitions based on the availability of passive systems; and (3) specifying "7BD* for information required to be specified in the TS but where detailed design, equipment selection, or other efforts are incomplete.
Provide specific justificaticn for the changes to the completion times and surveillance intervals in accordance with the basis for the staff's evaluation of related topical reports. In addition, submit the "TBD" information on the technical specifications presented in Chapter 16 of the SSAR, or provide a schedule for doing so.
Response
A majority of the completion times and surveillance frequencies included in the AP600 technical specifications submitted in the SSAR are identical to those included in NUREG-1431, Revision 0. Where differences exist, the AP600 completion times and surveillance frequencies were selected based on plant-specific operation differences or design features such as the 24-month refueling cycle. The AP600 surveillance frequencies also , reflect assumptions used in the AP600 PRA evaluation. The response to Q630.07 provides a summary of AP6v0 design differences that impact the AP600 technical specifications. The following table provides information relevant to those items identified as TBD in the AP600 technical specifications. Note that the setpoints in Tables 3.3.1-1,3.3.2-1, and 3.3.6-1 will not be completed prior to design certification since the calculation of plant-specific setpoints requires specific equipment selection. Setpoi.it values previously provided in these tables are nominal values provided as examples. s
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NRC REQUEST FOR ADDITIONAL INFORMATION mi tim
- 5. .
Tech Spec Page Type Topic Proposed Resolution Commitment Date 3.1.2 16.1-64 Parametric Shutdown margin linut Shutdown margin limit is 1.6% (Subsection Complete Value 15.4.6). 3.1.5 16.1-89 Parametric Rod drop time Rod drop time is 2.4 seconds (Ssubsection Complete Value 15.0.5). 3.3.1. Actions Pages 16.1-188, Completion cts to restore 3/4 RTS Completion time to restore 3/4 reactor trip Complete D, E,1. K, L, 189, 190, 191, Time channels to operable system channels to operable status is 24 hours. M, Q, and R 192,193, and status and associated 232,233,234, bases 237,238,239, 240,243,244 3.3.1, Table 16.1-200 and 228 Setpoint Negative flux rate trip Instrumentation equipment specifications are COL 3.3.1-1, Item setpoint and DAL needed to develop AP600-specific setpoints.
- 16. and associated Bases
- B3.3.1 16.1-250 Parametric Example time for delta- The example time for delta-T and calorimetric Complete Value T and calorimetric difference to exceed < < 1% is 24 hours.
difference to exceed
<<1%
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NRC DEQUEST FOR ADDITIONAL INFORMATION Tech Spec Page Type Topic Proposed Resolution Commitment Date 3.3.2, Actions 16.1-251, 252 Completion cts to restore ESF The completion times to restore one logic Complete C and E and and 296,297 Time logic groups or gmup on the inoperable division to operable associated channels to operable status (Action C.2) and to restore the Bases status (ESF functions inoperable channel to operable status (Action implemented using 2/2 E.2) are both 12 hours. logic) 3.3.2, Actions 16.1-252 and 297 Completion cts to restore ESF The completion time to restore one logic group Complete D and F and Time logic groups or on 3/4 divisions to operable status (Action D.2) associated channels to operable is 12 hours. The completion times to restore Bases status (ESF functions one logic group on all four divisions to implemented using 2/4 operable status (Action D.3) and to restore 3/4 logic) channels to operable status (Action F.2) are both 72 hours. 3.3.2, Action 16.1-252 and 298 Completion CT to restore The completion time to restore the protection Complete G and Time protection logic cabinet logic cabinet to operable status is 72 hours, associated to operable status Bases 3.3.2, Table 16.1-259 and 284 Setpoint DAL for CMT level Instrumentation equipment specifications are COL 3.3.2-1, Items needed to develop AP600-specific setpoints. 7d/e/flg and associated Bases W Westinghouse
l NRC REQUEST FOR ADDITIONAL INFORMATION Ei" El l Tech Spec Page Type Topic Proposed Resolution Commitment Date 3.3.2. Table 16.1-259 and 285 Setpoint Setpoint and DAL for 4 Instrumentation equipment specifications are COL 3.3.2-1, item kV bus undervoltage needed to develop AP600-specific serpoints. 7h and associated Bases 3.3.2, Table 16.1-262 Setpoint Containment fligh 1 Equipment specifications are needed to develop COL 3.3.2-1, Items and 2 radioactivity AP600-specific setpoints. 13 and 15 B3.3.2, items 16.1-284,285 Setpoint ADS Stages 2/3/4 time Equipment specifications are needed to develop COL 7e/f/g delays AP600-specific setpoints. B3.3.2, Item 16.1-285 Setpoint ADS actuation delay Equipment specifications are needed to develop COL 7h time following 4-kV AP600-specific setpoints. bus undervoltage 3.3.3, Actions 16.1-303 Completion CT to thermocouples to 3.3.3, Actions E, F, and G are being revised 7/1/93 E/F/G Time operable status consistent with NUREG-1431. The submittal will include the completion times. W Westinghouse
NRC REQUEST FOR ADDITIONAL.INFORMATION mu unt s" *s e Tech Spec Page Type Topic Proposed Resolution Commitment Date 3.3.4. Table 16.1-318 Parametric Remote shutdown Remote J.2down workstation channels will be COL 3.3.4-1 Value, workstation function specified based on the results of the completed Function and required channels function-based task analysis described in Identification Chapter 18. A preliminary list of functions required to be included on the remote shutdown workstation is being developed in response to Q420.102. 3.3.5. Table 16.1-326 Setpoint VES main control room Equipment specifications are needed to develop COL 3.3.6-1, high radiation isolation AP600-specific setpoints. Items I and 2 signal setpoint B3.4.2 16.1-346 Parametric Time to complete a The time to complete a rapid shutdown Complete Value rapid reactor shutdown following a violation of the minimum temperature for criticality is 8 hours. 3.4.5 LCO 16.1-365 and 367 Parametric Pressurizer volume @ A calculation to determine the volume is 7/1/93 and Bases Value 92% of indicated level underway. span B3.4.9 16.1-394 Par 2 metric Alarm range for Design and analysis efforts are underway to 12/31/93 Value increases in determine the alarm range. unidentified RCS leakage flow rates from the containment sump and air cooler
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NRC REQllEST 009 ADDITIONALINFORMATION su !q w y Tech Spec Page Type Topic Proposed Resolution Commitment Date SR 3.4.10.2 16.1-400 Parametric l-131 sample after how 'this statement will be deleted when the AP600 Complete Value many hours of Mode 1 technical specifications are revised in operation.;>_15% rated accordance with NUREG-1431. Revision 0, thermal power 3.4.12, 16.1-409 and 413, Completion cts for inoperable Determination of completion times and 12/31/93' Actions A and 414 Time ADS valves- surveillance frequencies for the passive fluid B and Bases systems is in progress. SR 3.4.12.1 16.1410 and 414 Surveillance Frequency for at-power Determination of completion times and 12/31/93 and Bases Frequency ADS valve operability surveillance frequencies for the passive fluid surveillance test systems is in progress. 3.4.13 LCO, 16.1-415, 416, Parametric LTOP vent area A calculation to determine LTOP vent area is 7/1/93. Action C, SR 417 and 420,423 Value underway. 3.4.13.3, and Bases SR3.4.14.1 16.1-425 Parametric . RCS flow during. Analysis to determine the minimum required 7/1/93 Value Modes 3/4/5 RCS flow during Modes 3/4/5 is underway. I' 3.5.1 Actions 16.1-429,430 and Completion cts and SFs for Determination of completion times and 12/31/93 A/B/C, all 433,434,435. Time, accumulators surveillance frequencies for the passive fluid SRs, and - Surveillance systems is in progress. Bases Frequency 630.9-6.
- - - .. - - .- - . ~ . - - - - . . . - -.= . =.- . . . = . - . .-- = - . . . - - . ..... - .
e , NRC REQUEST FOR ADDITIONALINFORMATION i Texh Spec Page Type Topic Proposed Resolution Commitment Date 3.5.2 Actions 16.1-437, 438.439 Completion cts and SFs for core Determination of completion times and 12/31/93 A/B/C/ DIE, and 442,443, Time, makeup tanks surveillance frequencies for the passive fluid S R3.5.2.1, 444,445,446 Surveillance systems is in progress. SR3.5.2.2 Frequency and Bases 3.5.3, Actions 16.1-447,448 and Completion cts and SFs for Detennination of completion times and 12/31/93 A/B/C, 451,452,453 Time, passive residual heat surveillance frequencies for the passive fluid S R3.5.3.1, Surveillance removal heat systems is in progress. SR3.5.3.2, Frequency exchangers and Bases 3.5.4, Actions 16.1454,455 and Completion cts and SFs for Determination of completion times and 12/31/93 A/B/C/D/E, 457,458,459 Time, IRWST surveillance frequencies for the passive fluid all SRs, and Surveillance oystems is in progress. Bases Frequency B3,6.1 and 16.1-464 and 473 Parametric Allowable containment ne allowable containment pressure for the Complete B3.6.2 Value pressure for limiting limiting design basis accident is 45 psig DBA (Section 6.2). B3.6.4 16.1-496 Parametric Containment analysis ne initial containment pressure analysis Complete Value initial containment assumption is 1 psig (Section 6.2). pressure B3.6.5 16.1-501 Design Basis Clarify what the TBD The limiting design basis accident is the main Complete Accident represents steam line break. 630.H W Westinghouse
r 3 NRC REQUEST FOR ADDITIONAL INFORMATION
.jjy !!G Tah Spec Page Type Topic Proposed Rer.olution Commitment Date B3.6.6 16.1-507 Parametric PCS analysis respmse The PCS analysis response time to achieve full Complete Value time to achieve full flow is 1I minutes (Ssubsection 6.2.1.1).
flow B3.6.7 16.1-514 Parametric Times to reach 3.5% A volume percent of 3.5 is reached during the Complete l Value and 4% post-LOCA seventh day after the accident, assuming no hydrogen recombiners (Subsections 6.2.4.2.2 and concentrations without 6.2.4.3.3). A volume percent of 4.5 is reached recombiners during the twelfth day after the accident, assuming no recombiners. Thus, initiation of a recombiner by the eleventh day aller the accident would prevent the concentration from exceeding 4.5%. 3.6.8, 16.1-518,519 and Parametric pil adjustment tank Analysis to determine pli adjustment tank 12/31/93 SR3.6.8.1, 522 Value volume limits volume limits is underway. and Bases B3.7.6 16.1-562 Paramettic MCR temperature after The MCR temperature rise is limited to Cornplete Value a loss of VBS < 15cF for 72 hours following the loss of VBS (Subsection 6.4.2.2). 3.8.1, Action 16.1-569 and 575 Completion CT to restore The completion time to restore inoperable Class Complete A and Bases Time inoperable Class 1E de 1E de power subsystem to operable status is 2 power subsystem to hours. operable status W Westinghouse
NRC REQUEST FOR ADDITIONALINFORMATION Hit IE _ k Tah Spec Page Type Topic Proposed Resolution Commitment Date 3.8.3, Action 16.1-569 and 592 Completion CT to restore ne completion time to restore 3/4 Class IE Complete A and Bases Time inoperable Class 1E inverters to operable status is 24 hours. inverter to operable status 3.8.5, Action 16.1400 and 605, Completion CT to restore ne completion time to restore inoperable Class Complete A and Bases 606,607 Time inoperable Class IE lE ac instrument and control bus to operable AC I&C bus to status is 2 hours and also 4 hours from operable status discovery of failure to meet the LCO. 3.8.5 Action B 16.1-600 and 605, Completion CT to restore Class 1E The completion time to restore 3/4 Class 1E de Complete and Bases 606,607 Time de electrical power electrical power distribution subsystems to distribution subsystem operable status is 24 hours. to operable status 3.8.7, Table 16.1-618 and 623, Parametric Class IE battery Component selection is needed to determine the COL 3.8.7-1 and 624 Value specific gravity limits values. Baws SSAR Revision: Proposed revisions as indicated in the preceding table.
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