ML20028H394
| ML20028H394 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 12/20/1990 |
| From: | Scaletti D Office of Nuclear Reactor Regulation |
| To: | Marriott P GENERAL ELECTRIC CO. |
| References | |
| NUDOCS 9101020308 | |
| Download: ML20028H394 (23) | |
Text
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l December 20, 1990 Docket Nc. STN 50-605 Patrick W.
Marriott, Manager Licensing & Consulting Services GE Nuclear Energy General Electric Company 175 Curtner Avenue San Jose, California 95125
Dear Mr. Marriott:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE GENERAL ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE ADVANCED BOILING WATER REACTOR DESIGN During the course of the review of your application for certifi-cation of your Advanced Boiling Water Roantor Design (ABWR), we have identified a need for additional information.
Our request for additional information, contained in Enclosure 1, addresses the CP/ML Rule, 10 CPR 50.34(f), and the unresolved and generic safety issues (USIs and GSIs) discussed in Appendices 19A and 19B of the ABWR Standard Safety Analysis Report.
Provided as is a list of USIs and GSIs which includes those that the staff believes to be applicable to the ABWR design.
You should ensure that your treatment of the USIs and GSIs is consistent with that list.
Also, there exists a set of USIs and GSIs identified in NUREG-0933 as resolved issues.
We request that you identify and review those issues to determine if the specified resolution is applicable to the ABWR design.
If the resolution is not applicable, please discuss the ABWR resolution.
We request that you provide your responses to the enclosures by February 15, 1991.
If you have any concerns regarding this request, please call me on (301) 492-1104.
Sincerely, orginal signed by Dino C.
Scaletti, Project Manager giO10{j$og 90 00 Standardization Project Diroc. tate cg OD R
Division of Advanced Reactors PDR and Special Projects P
Office of Nuclear Reactor Regulation
Enclosures:
As stated cc w/ enclosures:
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Docket No. STN 50-605 Patrick W. Marriott, Manager Licensing & Consulting Services GE Nuclear Energy General Electric Company 175 Curtner Avenue San Jose, California 95125 Dear Mr. Marriott
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION REGARDING THE GENERAL ELECTRIC COMPANY APPLICATION TOR CERTIFICATION OF THE ADVANCED BOILING WATER REACTOR DESIGN During the course of the review of your application for certifi-cation of your Advanced Boiling Water Reactor Design (ABWR), we have identified a need for additional information.
Our request for additional information, contained in Enclosure 1, addresses the CP/ML Rule, 10 CFR 50.34(f), and the unresolved and generic safety issues (USIs and GSIs) discussed in Appendices 19A and 19B of the ABWR Standard Safety Analysis Report.
Provided as is a list of USIs and GSIs which includes those that the staff believes to be applicable to the ABWR design.
You should ensure that your treatment of the USIs and GSIs is consistent with that list.
Also, there-exists a set-of USIs and GSIs identified in NUREG-0933 an~ resolved issues.
We request that you identify and review those issues to determine if the specified-resolution is applicable to the ABWR design.-
If the resolution is not
.applicabic, please discuss the ABWR resolution.
We request that you provide your responses to the enclosures by February 15, 1991.
If you have any concerns regarding this request, please call me on (301) 492-1104.
Sincerely, 0
Ql' Mdh
'~
Dino C.
Scaletti, Project Manager Standardization Project Directorate Division of Advanced Reactors and Special Projects Office of Nuclear Reactor Regulation Enclosures As stated cc w/ enclosures See next page
.. - _. _ _ _._,_~-._.
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I Mr. Patrick W. Marriott D0ci.et No. STN E0-605 General Electric Coripany
(
cc: Mr. Robert Mitchell General Electric Compariy 176 Curtner Avenue San Jose, Californie 95114 Hr. L. Gifford, Program fiarager Perulatory Programs GE Nucleer Energy 12300 Twir. brook Parkwey
$uite 315 Rockville, Maryland 20852 Director Criteria & Standards Divisich OfficeofRadiationPrograms V. S. Environmental Protection Agency 401 M Street, S.W.
Washington, D.C.
20460 i:
Daniel F. Giessing Division of Nuclear Regulation and Safety Office of Converter Reactor Deploytent, NE-12 Office cf Nuclear Enerp Washington, D.C.
20545 l
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l RE0 VEST FOR ADDITIONAL INFORMATION - ABWP APPENDICES 19A AND 19B.
CP/ML RULE 10 CFR SC.34(F) AND USI's AND G,l',s_
190.2.1 Quality Assurance -ExpandedOALIST[!.F.1)
Classification of Instrumentation, Control, and Electrical Equipment [II.F.5]
The ALWR Resolution Summary for these issues states:
1.
The designer shall identify any structures, systems, or components (items) that are not safety related but for which provisions be, yond normal indus-try practice are judged to be needed to provide desired reliability and availtbility.
2.
At the same tine, specific surveillerce/ maintenance provisions (appropriate for the specific item and desired reliability and availability) shall be identified for those iter:s.
The NRC evaluation is that ALWRs should have a Peliability Program to ensure that the facility is operated and maintained within enveloping PRA assumptions throughout its life.
The NRC anticipates that these new (Reliability Progrem) requirements will effectively subsume the 1.F.1 and II.F.5 issues and these issues can be considered resolved.
The ABWR Resolution states:
1.
The ABWR application of quality system requirements satisfies the ALWR resolution.
2.
An interface requirenent (Section 190.3.1) is included to ensure that quality system requirements will be provided during construction and operation.
3.
Tierefore, this issue is resolved for the ABWR, REQUEST FOR ADDITIONAL INFORMATION 1.
It is not clear to the staff that the AbhR 55AR describes how points 1 and 2 of the ALWR Resolution Summary (above) are to be satisfied.
That is how is the ABWR designer identifying items for which provisions beyond normal industry practice are judged to be needed? And how are specific surveillance / maintenance provisions being identified for those items? SSAR Table 3.2-1 is used to show the quality assurance that is applied to plant items. The table indicates that a quality assurance program meeting 10 CFR 50 Appendix B either does or does not apply.
In some instances, where Appendix B does not apply, there is reference to a footnote regarding quality assurance.
Such references are neither wide-spread enough nor specific enough to really meet an objective of the classification system which is to assign appropriate Quality Control and Quality Assurance measures.
The SSAR should be clarified in this regard, or justification should be given for not doing so.
For example, footnote "u" regarding quality assurance for non-safety-related fire protection items should make it clear that a quality assurance program meeting the guidance of Branch Technical Position CMEB 9.5-1 (NUREG-0800) will be applied to each such item.
Similarly, for non-safety-related radioactive waste management items, a footnote should make it clear 1
that a quality assurance program meeting the guidance of Regulatory Guide 1.143 will be applied during design and construction.
The safety parameter display system (or its equivalent), though not safety-related, should have a quality assurance program beyond normal industry practice applied, and this should be clear in Table 3.2-1.
Generic Letter 85-06, " Quality Assurance Guidance for ATWS Equipment That Is Not Safety Related," is also applicable.
If GE has not already done so, it should ascertain whether there are other ABWR plant items within the scope of points 1 and 2 of the ALWR Resolution sumary (above) and revise Table 3.2-1 accordingly if required. Then the ABWR Resolution should reference Table 3.2-1 to show how GE has resolved THI issues I.F.1 and II.F.5 for the ABWR.
198.3.1 Quality Assurance Program The statement in the ABWR SSAR, " Applicants referencing the ABUR design shall have a Quality Assurance Program satisfying the requirements of Section 19.B.2.1(2) including the right to impose additional environmental require-ments," does not appear to accurately reflect the reouirements of 198.2.1(2).
The "right to impese additional environmental requirements
- is not as encompass-ing as "the right to impose additional requirem(nts to supplement the 10 CFR 50, Appendix B requirements."
It is not clear what is meant by " environmental" requirements.
REOUEST FOR ADDITIONAL INFORMATION 2.
Clerify that applicants referencing the AWR design shell have the right to impose additional requirennts to supplement i
the 10 CFR 50, Appendix B requirements, or justify not doing so.
19A.2.41 Procedures for Feedback of Operating, Design and Construction Experience (I.C.1)
The response to this item states:
" Interface requirement, see Subsection 19A.3.6."
Subsection 19A.3.6 states:
...(ReferenceSubsection19A.2.4.)"
The response to Subsection 19A.2.4 states:
... This requirement is not applicable to the ABWF:.
It applies only to PWP-type reactors." This-series of references takes us from an NRC requirement that rocedures provide for evaluation and feedback of related experience in a t me y manner to the ABWR designers and constructors to a conclusion that there is no requirement.
I REQUEST FOR ADDITIONAL INFORMATION 3.
Clarify the response to 19.2.41, paying particular attention to the references, or justify not doing so.
19A.2.42 Expand QA List (1.F.1)
The response states:
- This issue is addressed in Appendix 19B."
REQUEST FOR ADDITIONAL INFORMATION 4.
Since Appendix 198 has many pages, please narrow the reference to item 198.2.1 in Appendix 198.
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210.51 19B.2,2 In-SituTestingofValves-TestAdequacyofStudy(GenericSafety Issue II.E.6.1)
The information in this section should be revised to more nearly reflect the current status of this issue. GSI II.E.6.1 originally consisted of the following sub-issues:
1.
In-situtestingofmotoroperatedvalves(MOV) 2.
In-situ testing of pressure isolation valves (PIV)
I 3.
Reevaluation of thermal overload protection devices for motor operated valves 4
In-situ testing of check valves Sub-issues 1, 2 and 3 are no longer-considered to be a part of II.E.6.1.
Sub-issue 1 was subsumed by the staff's evaluation of responses to Generic Letter 89-10, " Safety-Related MOV Testing and Surveillance".
Sub issue 2 was subsumed by Generic Safety Issue 105, " Interfacing Systems LOCA in Light-Water Reactors". Sub-issue 3 is considered to be resolved for the ABWR on the basis of the unconditional commitnient in the SSAR Table 1.8-20 to Regulatory Guide 1.106, " Thermal Overload Protection for Electric Motors on Motor-Operated Valves".
Sub-issue 4 remains unresolved at this time. During a meeting on April 7,1966 between the staff and industry representatives, it was agreed that industry would initiate an aggressive program to resolve the chec cvalve issue. Since that time, the Institute of Nuclear Power Operation (INPO), the Electric Power Resear:h Institute (EPRI), the Nuclear Industry Check Valve Group (NIC) and the ttaff have made some progress in addressing this issue.
However, as stated in a letter from Mr. James Taylor, NRC Executive Director for Operations, to Mr. Z. T. Pate, President of INPO, dated April 20,.1990, the staff continues to find weaknesses in the efforts of individual licensees to improve the performance of check valves. To assist the staff in its continuing evaluations of industry actions, INPO was requested to provide the status of its activities and perspectives regarding the resolution of the check valve issue.
The staff has.not yet received a complete response to this request.
The staff does not agree that the information in the 'ABWR Resolution" of Subsection 19B.2.2 in the SSAR is sufficient to resolve this issue for the ABWR.
i The exceptions to position indication of check valves will require some clarifi-cation.
However, the staff prefers that this type of information be included as a part of the ASME SEction XI Inservice Test Program for safety-related pumps and valves which is discussed in the SSAR Subsection 3.9.6.
Therefore, GE is requestedtoreviseSubsection198.2.2relatedtosub-issue 4toreflectamore broad commitment to the collective industry and NRC activities relative to implementation of the resolution of issues on in-situ testing of check valves.
In addition, the staff will need to complete its review of the ABWR Inservice Testing Program before this issue can be considered resolved.
Since sub-issue _1 has been subsumed Subsection 19B.2.2 should also include a
. commitment to provide a response to Generic Letter 89-10 which will be appli-cable to the ABWR.
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l 19B.2.4 Nuclear Powtr Plant Design for the Reduction of Vulnerability to L
Industrial 5eb0tage (U51-AF9) i 1.
Reliance on Security Measures In Appendix'19B.2.4 of the Standard Safety Analysis Report, ABWR resolution of GSI A-29 relies upon separation of trains and usual access constraints and security provisions for protection against insider sabotere.
Separation of trains has-benefit against an external adversary, but it does not assure addi-tional protection against an insider with authorized access to both trains.
Card readers on (oors between trains could present a deterrent effect against i
insidtr sabotage, but locked interior doors also have the potenti61 for inter.
fering with rapid access under emergency conditions.
Separation of trains also allows the possibility of plant management to group plant personnel into teams attheBWR-6inCofrentes, Spain)(. such as used to en1ance quality of operations that each work on only one train No information is presented on how security access controls would be used to deter insider sabotage at an ABWR, nor to assure that these controls would not interfere with safety, furthermore, Appen-dix 19B.2.4 of the ABWR Standard Safety Analysis Report does not identify design features that decrease reliance on physical security programs for protection-
)
1 against an insider.
)
2.
Lack of Cottnitment Appendix 19B.2.4 does not contain a commitment regarding the ABWR resolution of GS! A-29.
Of the 13 pages devoted to GS1 A 29, about 12 pages present BhR requirements of the draft EPRI ALWR Requircutnts Document, most of which are not relevant to CS! A-29. Although the etsif has some as yet unresolved ques-tions about the EPRI.rtquirements, the section titled "ABWR Resolution
- does not conrnit to meeting these draf t requirements.
It-states the ABWR is " con-sidered to comply with" and "is consistent with* the ALWR requirements.
It
. also states that the-ABWR complie>5 "with 10 CFR 73 and therefore with the ALWR requirements,' whereas compliance with existing 10 CFR Part 73 neither implies compliance rith the somewhat different ALWR requirements nor commits to any improvements over existing designs that also comply with 10 CFR Part 73.
Chap-ter 9, Section 5.2.2.1 of the ALWR Requirements Document is quoted on page 19B.2 11 as:
... a sabetage vulnerability analysis shall be conducted prior to finalizing design.
Such eralysis shall be in accordance with the criteria and assumptions in-Section 5.1.3.*
Further, Section 5.2.4,2 (on page 198.2-12) states:
"The design of the security system shall-include an evaluation of its impact on plant operation, testing,'and maintenance.'
Neither of these ALWR required evaluations is contained in the ABWR Standard Safety Analysis Report.
In addition, ABWR Response 910.10 says that issues dealing with plant internal L
security corinunications are outside the scope of the ABWR Standard Plant design, althoughtheALWRRequirementsDocumentcitedinAppendix19B.2.4(19) includes such consnunication requirements.
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3.
Inadequate Responses to Prior Request Responses to a staff request for additional information are given on pages 20.3-200 through 20.3 202 of the ABWR Standard Safety Analysis Report.
Several of these responses are inadequate or incorrect.
For example Response 910.7 incorrectly implies that Appendix 19B.2.4 includes a discussion of the insider and outsider sabotage actions that would be necessary to cause significant core damage or Part 100 release levels.
4 No Ana_1ysis of Issues of insider Sabotace The Commission's Severe Accident Policy Statement included the statement:
"The Commission also recognizes the importance of such potential contributors to severe accident risk as human performance and sabo-tage. The issues of both insider and outsider sabotage threats will be carefully analyzed and, to the extent practicable, will be empha-s ued in the design and in the operating procedures developed for new plants To comply with this position and to resolve GSI A-29, the staff anticipated that a systematic analysis of sabotage vulnerabilities would be performed to identify combinations of components which, if tampered with would lead to core damage.
Thiswouldhaveallowedanexaminatienoftheresultstoseeifdesignchanges were feasible to ensure that the terpering would be detected before the entire set could be disabled, or to see if it was feasible to " harden" critical con o-nents to make them less susceptible to tampering.
Such an analysis has not acen presented.
19B.2.14 Beyond Design Basis Accident in Spent Tuel Pools {GI-82)
Generic sefety issue C2 "Beyond Design Basis Accidents in Spent Fuel Pools" is concerned with the loss of the pool water which may result in a fire in the pool l
causing a release of fission products.
In the ABWR resolution, it is indicated l
that the spent fuel pool will be designed to withstand a design basis earthquake without pool drainage, and will be arranged to prevent cask movement over the i
pool, which will be accomplished through the use of a separate cask loading pit.
Was a cask drop in the cask loading pit considered? Since the cask loading pit-is adjacent to the s)ent fuel pool, discuss the effect of such a cask drep in the loading pit on tle refueling pool.
In addition it appears that the fuel pool is near the staging area for the reactor vessel head, indicate the effect on the fuel pool of vessel head drop on the adjacent staging area.
198.2.17 ProbableMaximumPrecipitation(GI-103)
GenericSafetyissueNo.103"DesignForProbableMaximumPrecipitation"(PMP) is concerned with the difference in the determination of PMP, By using the-recently developed NOAA/NWS procedures which are believed to be more realistic, PMP estimates larger than those obtained by previously used methodologies may lead to higher flood levels. Therefore, in ABWR resolution on Page 19 B-2-47, specify that the recently developed NOAA/NWS procedures will be used for determining PHP for a specific site.
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I, IgB.2.15 Interfacing Systems LOCA et BWRs - RHR Suction Valve Testing (GI-R, 1 4)
Recent CWR operating experience indicates that the isolation valves between the RCS and low pressure interfacing systems m6y not adequately protect against overpressurization of low pressure systems.
i for AEWRs, pressure isolation valve instrumentation and controls are provided 4
to (1) prevent opening shutdown cooling connections to the vessel in any loop when the pool suction valve discharge valve, or spray valves are open in the
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same loop (2) prevent opening the slutdown connections to and from the vessel whenever the RC5 pressure is above the shutdown range, (3) automatically close shutdown connections when RCS pressure rises above the shutdown range, and (4) prevent-operation of shutdown suction valves in the event of a signal that the water level in the reactor is low.
The ABWR has been designed to minimize the possibility of an interf acing system LOCA in the following ways.
The low pressure systems directly interfacing with the RCS are designed with 500 psig piping which provides for a rupture pressure of approximately 1000 psig.
In addition, tht high/ low-pressure motor-operated isolation valves have safety-grade redundant pressure interlocks. Also, the motoroperatedemergencycorecoo1In i
when the reactor tw at low pressure.g system (ECCS) valvets will only be tested All inboard check valves on the ECCS will be testable and have position indication. Additionally, design criteria used by GE require that all pipe designed to 1/3 or greater of reactor pressure requires two malfunctions to occur before the pipe would be subjected to reactor syster pressure.
The pipe designed to less than 1/3 reactor pressure requires at least three malfunctions before the pipe would be subjected to reactor system pressure.
Position Since ABWR low pressure systems arc designed only for 500 psig rather than the full RCS design pressure of 1250 psig, the /,BWR design should provide (1) the capabilityfor-leaktestingofthepressureisolationvalves,(2)valveposition
- indication that is available in the control room when isolation valve operators aredeenergizedand(3)high-pressurealarmstorerncontrolroomoperatorswhen rising RCS pressure approaches the design pressure of 6ttached low-pressure sys-tems ano both isoletion valves are not closed.
It is the staff's position that GE should confirm that the above design features are incorporated into the ABWR
- design, GI-96 was related to pWR$ which considers the failure of the low pressure isolation valves between the RCS and RHR system in PWRs.
The issues contained in GI-96 now are incorporated into GI 105.
19.A.2.6 Reduction of~Challenses and Failures of Relief Valves -
feasibility Study anc, Sy. ten Modification (ll.K.3.16) position 2
The-record of relief-valve failures to close for all boilingly 30 in 73 reactor-water reactors (SWR $) in the past 3 years of plant operation is approximate years (0.41failuresperreactor-year).
This has demonstrated that the failure i
of a relief. valve to close would be the most likely cause of a small-beak loss-of-coolantaccidentLOCA). The high failure rate is the result of a high-6
relief-valve challenge rate and a relatively high failure rate per challenge (0.16 failures per challenge). Typically five valves are challenged in each event. Thisresultsinanequivalentfailurerateperchallengeof0.03.
The challenge and f ailure rates can be reduced in the following ways:
Additional anticipatory scram on loss of feedwater, Revised relief-vtive actuation setpoints, Increased emergency core cooling (ECC) flow, Lower operating pressures, Earlier initiation of ECC systems, Heat removal through emergency condensers, Of fset v61ve setpoints to open fewer valves per challenge, Installation of additional relief valves wit 1 a block-or isolation 4 valve feature to eliminate opening of the safety / relief valves (SRVs), consistert with the ASME Code,
'9) i.9 creasing the high steam line flow setpoint for main steam line isohtion valve (MS!V) closure, Lowering the pressure setpoint for MSIV closure, Reducing the testing frequency of the itSIVs, More-stringent valve leakage criteria, and Early removal of leaking valves.
An investigation of the feasibility of rt:ducing challenges to the relief valves by use of the aforcrentioned methods should be conducted. Other methods should also be included in +he feasibility study. Those changes which are shown to reduce relief-valve ch.11entes without compromising the performance of the relief valves or other systems should be implenented.
Challenges to the relief valves should be reduced substanti6lly (by an order of magnitude).
Resolution The staff requirts the following additional information to complete the review on this item:
In the 130 position for this item 13 possible ways are listed for reducing the challenge and f ailure rates of safety relief valves. These items that will be implemented for the ABWR should be listed an6 the basis for concluding that theyreducethechallengeratesubstantiallyshouldbeprovided.
19A.2.30 identification of and Recovery from Conditions Leading to ICC (II.F.2)
Position Licensees shall provide a description of any additional instrumentation or controls (primary or backup) proposed for the plant to supplement existing instrumentation (including primary coolant saturation monitors) in order to provide en unambiguous, easy-to-interpret indication of inadequate core cooling (ICC). A description of the functional design requirements for the system shall also be included. A description of the procedures to be used with the proposed equipment, the analysis used in developing these procedures, and a schedule for installing the equipment shall be provided.
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i Resolution The staff requires the following additional information to complete the review on this item:
GE refers to Topical Report SL1-8211, " Review of SWR Reactor Yessel Water Level Heasureraent System," in their response. Generic Letter No. 84-23, " Reactor Vessel Water Level Instrumentation for BWRs
- discusses in detail the potential improvements required as a result of the topical re? ort review.
How the ABWR design satisfies the requirements of GL No. 84-23 siould be explained in detail.
II.K.3.25 - Effect of loss of Alternating Current Power un pump Seals (II.K.3.25)
Position The licensees should dttermine, or, a plant-specific basis, by analysis or expe-rin:ent, the consequences of a loss of cocling water to the reactor recircula-tion pump seal coolers. The purp seals should be designed to withstand a com-p1tte loss of alternating-current (ac) power for et least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
Adequacy of the seal design should be demonstrated.
Resolution The intent of this item's position is to prevent excessive loss of reactor coolant irrentory following en enticipated operationel occurrence.
Loss of AC pcwer for this is construed to be loss of offsite power.
The staff requires the following additional information from GE to ccmplete the ieview on this itt*:
Confirm that failure of the following systems will not generate a LOCA.
a Recirculation Motor Cooling System (RPC) b Recirculttion Motor Seal Purge System (RMSp) c Recirculation Motor Inf16 table Shaf t Seal Subsystem (RMISS) 19B.2.12 Bolting Degradation or Failure (GI-29) 1.
The applicant should define bolting in detail. Bolting in this context should include bolts, studs, embedments, machine / cap screws, threaded fasteners, and associated nuts and washers.
E.
Define high strength bolting and tredium strength bolting in terms of material and u chenical properties.
3.
provide bolting n.anufacture process (e.g., beat treated, quenched, ternpered, etc.)
4.
Identify specific safety related items (e.g., equipment and piping systems) where the high strength bolting or ndium strength bolting will be used.
5.
Discuss how to avoid the intergranular stress corrosion cracking (IGSCC) of bolting in a BWR bydrogen environment.
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6.
Identify thread lubricants th6t will be used and identify chemical compound (s) in them.
7.
The applicar.t discussed the ALWR Resolution initiated by the Atomic Industrial Forum / Metal properties Council Task Group and BWR Requirements in the EPRI ALWR Requirements Document.
It is unclear whether the applicant will follow the resolutions and requirements.
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19A.E.26 and 19A.27 Containment Isolation Dependability (!!.E.4.2) and Cortafnment Purging (II.E.4.4).
430.300. Justify non compliance of the current ABWR design with position 7 of (THI-!!. THI issue !!.E.4.2.
This design does not include containment iso.
E.4.2) lation on a high containment radiation signal for the containment purge and vent isolation velves as required by position 7.
430.301. Explain whether the reopening of is016 tion valves is perforraed on a (TM1-l!. valveby-valvebasis(whichisacceptebleaccordingtotheguidance E.4.2) provided for TMI issue II.E.4.2) or as a ganged reopening (which is i
notacceptable).
430.302. Discuss the administrative controls that will be in effect to assure (TMI-II.
that closed purge isolation valves cannot be inadvertently opened.
E.4.2) a 430.303.
Explain the technical basis for the 72 (versus 24) hour technical s seci.-
. (TMl-II.
fication limit allowing the large diameter purge lines to be open a >ove E.4.4) 15% power at the beginning and end of the fuel cycle.
Since taese valves can se open during power operation verify that the large ditmeter purge Ifne isolation valves can suc,cessfully perform their intended (22*)
functionunderaccidentconditions(containrentdesignpressures).
l' 430.304 Ev61uate che adequacy of two-inch at-power purge lines for reiteving (TMI-II.
primary containment excessive pressure (resulting from a combination of E.4.4) compresses air system leaks steam leaks, and elevated containter.t atmosphere tercperatures associated with hot days or degraded contain-ment HVAC performance) during normal operation in light of current reactor operating experience in which many plant operators are forced to periodically open the large containment purge lines at power to maintain normal containment pressure.
Include an evaluation of the ability of the the two-inch lints to maintain normal containment pressure, the pot.sible need to o purge and vent isolation valves,perate the large (22") containment and identify the size of a small purgu line required to preclude operation of the 22" centainment purge ifnes during power operation.
430.305. Also note a discrepancy between Figure 6.2-39a sheet 1 and Table (6.2.5) 6.2-7 of Amendment 11. Yalve T31-F007 is listed as a 2" valve in Table 6.2-7.
However, Figure 6.2-3ga has been modified and this valve now appears to be en the same lir.e as the 14" rupture dist to be used for the contairwnt overpressure protection system.
Other 1
information provided for this valve in the figure and the ttble is l
also contradictory. Podify the figure and teble to accurately i
i w
r,-,.
,.----,---,-e-
.w...--
~-m.m.
.,..~.-.s.
-iw-,
m
represent both the 2" at-power purge lirts and the containment overpressurization protection flow path.
430.300. Widely separated primary containment penetrations for the drywell l
(TMI.!!. end wetwell purge systems (supply side penetrations X-80 and X 240; E.4.2) and exhaust side ptrietrations X 81 and X 241, SSAR Figure 6.2-39a, i
sheet 1) have common prinary containnent outboard isolation valves.
Esplain how the above configurations comply with GDC 56 which requires the outboard isolation valve to be located as close to the contain-3 t
ment (i.e., the drywell or wetwell in this case) as practical.
l 19A.2.12 Evaluation of Alternative Hydrogen Control Systems (Item 1.xii) j 19A.2.21 Hydrogen Control System Freliminary Design (THI Issue !!.B.8) l 19A.2.40 pedicated Penetrations (for Hydrogen Recombiners) (TMI lssue 11.E.4.1) i l
198.2.6 Hydrocen Control Heatures and Effects of Hydrocen Burns on Safety l
ToiTiicent Weneric Safety Issue A 48) l 430.307. Section 19B.2.6 of the ABWR $$AR reflects the EPRI Requirerents Document (CP/HL positions on hydrogen generation, that is containment concentrttions i
1.xii) resultingfromanactivefuel-cladexidatIonof751andthecontain-rent atmosphere hydrogen concentration requirement is a concentration of less than 13%.
These requirements ere less conservative than the i
requirements of 10 CFR 50.3a(f), namely 200f and 101.
- Also, provide
{
an analysis and supporting docursentation demonstrating that tie hydro-gen control system will be able to maintain containment atmosphere j
within acceptable limits and that the hydrogen recombiners will func-i tion in an extremely hydrogen rich environment, using the hydrogen i
gereration retes and allowable concentrations of 10 CFR 50.34(f).
a 430.308.
There is a discrepancy between ABWR $$AR section 19A.2.12 which states 1
(CP/ML that perranently installed recombiners are provided and section 1.xii) 6.2.1.1.1 which states that portable recombiners will be available for
]
use after a LOCA signal is generated (Section 6.2.5.2.7 also states that recombiners will be located on skids in the secondary contain-4 ment). Clarify the type of recombiners to be included in the ABWR design.
If portable recombiners are to be used and located outside primary containment, provide information detailing how containment integrity is to be maintained during system operation. Of specific j
concern is the possibility that leaks in the portions of the system outside containnent-could result in a flannable mixture and uncon-l trolled combusion. Additionally, EPRI ALWR requirements-presented in section 19B.2.6 require that the
- Plant Designer shall define a suitable scheme
- for removing residual hydrogen from the containment after an accident. This concern has not been addressed and it appears that there would not be sufficient oxygen in containment af ter an eccident to recombine all the hydrogen.
Thus, either an anaylsis-demonstrating that recombiners are sufficient to remove the hydrogen present af ter a beyond design basis accident, or a method for ensuring 1
j that purging containment, using either the Containment Overpressure 10 1
- _ _., _ _. -.-_ _.~.
i 5
l i
Protection System or purging through the Standby Gas Treatment System, i
will not result in a mixture that would be flamable upon contact with I
air, is required.
2 430.309. Section 6.2.5 of the ABWR SSAR asserts, but does not demonstrate, (TMI-II.
that mixing of drywell and suppression chamber atmospheres by natura1 8.8) circulation occurs and would be enhanced by containment sprays. There is no justification for the assertion that cornbusible mixtures will not form locally.
Provide the analysis justifying the assertion that combustible mixtures will not form locally.
430.310. There is no consideration of the effects of accidents beyond design (THI-II. basis, as required by this TMI issue.
The ABWR SSAR states that there B.8) are no design basis events that result in core uncovery or core heatup sufficient to cause significant inetal-water reaction, and therefore uses the Regulatory Guide 1.7 design basis metal-water reaction instead of addressing hydrogen generated by a 100% metal-water reaction.
Provide an anaylsis of the capebility of the hydrogen control system to miti-gate the effects of beyond design basis hydrogen and oxygen generation rates.
430.311. The design evaluation of the inerting system uses NE00-22155 rather j
(TMI-II.
than Regulatory Guide 1.7 oxygen generation rates.
The NED0-22155 B.8) generatics rates are not acceptable for use in licensing submittals.
(See the July 6, 1989 NRC SER for NED0-22155.)
Incorporate the appro-priate oxygen generation rates in the hydrogen control system analysis.
430.312. The containment isolation valves for the Flammability Centrol System (TMI-II.
have been identified in Table 6.2-7 of the ABWR SSAR. However, no other E.4.1) information regarding these conteinment penetrations have been provided.
The discussion of the recombiner system to be provided, as part of ABWR SSAR section 6.2.5, should include a description of the dedicated pene-trations for this system.
Include in your discussion details such as i
(1) how long af ter 1.0CA and at what concentration level of hydrogen the recombiner has to be activated, (2) line sizes as related to flo,w requirernents, (3) duration of recombiner operation, and (4) interface requirements for referencing applicants with regard to the recombiners (e.g., development of procedural provisions to assure ovailability of possibly shared portable hydrogen recombiners between sites on a timely basis and coordination of surveillance programs in accordance with SRP Section 6.2.5 acceptance criterion II.12). Also, include the flamability control system in SSAR Table 3.2-1, 19A.2.29 AdditionalAccidentHonitorinoInstrumentation(II.F.1) i 430.313.
Regarding additional accident monitoring instruir.entation identified (II.F.1) in NUREG-0737. THI Action Item II.F.1, Attachments 1 through 6, discuss compliance with the positions stated in the following clari-fications:
a.
Clarification (1), (2) and (4) for noble gas effluert monitor (NUREG Pages II.F.1-2, 3 and 6). Note that SSAR Sections 7.5 and 11.5 discuss only some positions identified in the clarifications.
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i b.
Clarifications (1) through (4) for sampling and analysis of plant effluents as they relate to ABWR scope (NUREG Pages !!.F.1 7, 8 and 9);
identify any applicable interface requirernent for the referencing applicant, c.
Clarifications (1), (3) and (5) for containment high range radiation monitor (NUREG Pages II.F.1-11 and 13).
d.
Clarification (5) for conteintrent pressure monitor (NUREG Page !!.F.1-14).
e.
Clarification (5) for the containment water level inonitors (NUREG Page II.F.1-16).
Include suppression pool water level in SSAR Subsection 7.5.2.1.
The staff finds GE's justification for considering the drywell sump level monitors as Category 3 rather than as Category I as required by Regulatory Guide 1.97, Rev. 3, unacceptable. Address the above concern, f.
Clarification (3) for containment hydrogen monitor (NUREG page II.F.1-18).
Also, clarify whether the monitors have the capability to operate from
-5 psig to design pressure as required by Regulatory Guide 1.97, Rev.3.
19A.2.47 Organizatien and Staffing to Oversee Design and Construction (II.J.3.1) 630.1.
include informatIon on "sition, part (B) you state that you will In item 19A.2.47 NRC po the technical resources director by the appli-cent." This is either a typographical error or a misinterpretation of the NRC position.
It is our position thet the information should include "the technical resources directed by the applicant." Please clarify your statement or provide a basis for the change to the content cf the information that we require.
19.B.2.18GenericImplicationsofMainTruisformerFailure(GI107) l 19.C.2.24 Electrical power Reliability (GI 128) 1 435.63 Description and enalysis demonstrating compliance of the offsite preferred power system to regulatory requirements has not been addressed in the ABWR SSAR. Provide a description and analysis
)
demonstrating compliance for the offsite power system within the ABWR standard plant scope from the utility /ABWR interface to the Class IE distribution system input terminals. Also, provide interface requirements for the offsite power system outside the l
ABWR standard plant scope from the utility /ABWR interface out to the utility grid system.
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435.64 Provide descriptive information and anaylsis or reference in the l
ABWR SSAR, where the descriptive design information or anelysis can I
be found, which demonstrates that the ABWR design is consistent with i
the ALWR resolution for generic issues described in Section 19B.2.18 and 198.2.24, 435.65 It is the staff position that transformers associated with the pre-ferred offsite circuits be separated by the maximum extent practical (preferably on different sides of a building) in order to minimize 12
the comon-cause effects of fire, missiles, or environmental effects on their operation.
I'rovide a description and interface requiree r.t. which demonstrates compliance with this staff posiilon.
435.66 It is the staff position that interconnectors between redundant divisions through safety or non safety buses shall be ciaintained with two norm 611y osen and interlocked devices that are separate and independent such tiat single failure or operator error cennot cause tfe interconnec-tion of or challegt to redundant divisions.
Provide a design des-cription, interface reouirements, and/or analysis which demonstrates compliance with this staff position.
435.07 Hon-safety computers and transient recorder loads shown on figure 8.3-5 have provisions included in their power supply design for automatically transferring these loads from class IE division 1
- 3.
In addition, it appears that the power supply may also in, provision for automatic transfer of these leads between divis and 2.
The design does not appear to meet regulatory positic' and 4c of Regulatory Guide 1.6 and thus does not appear to races independence requirements of criterion 17 of Appendix A to 10 C.
T' art 50, the intent of position 1 of Regulatory Guice 1.75, or the intent of generic issue 128.
Ex) lain how the design tutets the staff requirentots or provide design cienges such thet tie ABWR electric system design will meet the independence requirement of criterion,?
or identify this design as being in non compliance with criterion 17 and provide justification, a35.68 Identify all safcty and non-stfety loads that can be powered from more then one Class 1E divisions AC or DC power supplies.
135.69 ABWRResolution(3)ofSection19B.2.24indicatesthatalternate1,C power for Lbttery chargers is supplied through a series of physically separated breakers frce a different division.
These interdivisional breakers are in series, mechanically interlocked and Lept normally open during plant operation, providejustificatIonthatthisdesign is more reliabic than one where the alternate AC supply comes only from the division the DC system is associated with.
In addition provide physical layout drawings, enalysis, and/or interface requ, ire-ments which demonstrate that w1en the alternate AC is used the inde.
pendence and single failure requirements of criterion 17 are still met.
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)
I HIGH ANP MEDIUM PRIORITY UNRESOLVED GS1's TO BE j
~ ADUE55ED lh E\\0 TUT 10hM3 LM DE51Gb FEVIEW5'
)
flumber Title Priority Branch Application 25 Pediation Effects on Reactor Yesse1 High EMEB All Supports J
3 23 Reactor Coolant Pump Seal Failures High SRXB All 29 Lolting Degradation or Failures in NPP Migh EMCB PWR E7 Effects of Fire Protection System Medium SPLE All Actuatien of Safety Related Equipment 70 PORY and Block Valve Reliability Medium EMEB PWR l
79 Unanalyzed RY Thermal Stress During Medium EMCB
- PVP, Natural Convection Cooldown 1
I
- B7 Pailure of N CI Steam Line Without High SRXB BWR Isolation 94 Addittunal Low temperature Overpressure Figh SRXB PWR Protection !bsues for LWRs 205 1hterfacing Systems LOCA at BWRs High EMEB BWR 200 Piping & Use of Highly Combustible Medium SPLE All 1
Gases in Vital Areas D3 Dynamic Qualification Testing or Large High EMEB All Bore Hydttulic Snubbers 221 Fydrogen Control for Large, Dry PWR High SPLB PWR Containments 22B Electrical Fower Reliability High SELB All 230 Essential Service Kater Pump Failures High SPLB All at Multiplant Sites 4
335 Steam Generator and Steam Line Overfill Medium EMCB PWR L
B-17 Criteria for Safety Related Operator Medium LHFB All i
Actions B 55 1rtroyed Reliability of Target Rock Medium EMEB All I
Sefety Relief Yalves P 56 Diesel Reliability High SELB All B-61 Allowable ECCS Equipment Outage Periods Medium OTSB All HT4.4 Cuide11nes for Upgrading Other High LHFB All Procedures HTS.1 Local Control Stations High LHFB All liFS.2 Review Criteria for Human Factors High LHFB All Astects of Advance Controls &
Instrumeritation 3.D.3 Safety System States Monitoring Medium SICB All i
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'Althougn Generic Issue C-B has recently been resolved with no new requirements established and has been dropped from this list the staff is reviewing matters concerning the absence of the main steam isolatlon valve leakage control system frm the ABWk otsign.
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NEARLY.RE$0LVED'G$15 TO BE ADDRESSED 1h It0LtrT10t.Aa Li(R DE513t: F.EVIEWS jo Number Title Guidance Branch Application i
75 Ger.eric Implications of $alem ATWS GL $3 28' LQAB All
]
Events B3 Control Roern Habitability 6/29/84 Memo LHFB All B4 CE FORVs NUREG 1289 SRXB CE 22.0.4.1 Revise Deficiency Reporting Nrly Resivd DOEA All Requirenents
- l f
i 1
I l
i i
1
- *Plo new technical issues are expected to result from remaining closure activities.
EnclosureR FESOLVED GSI'S AND V51'5 TO BE ADDRESSED IN EYOLUT10 NARY LWR DESIGN hEVIEWS FLL khlCh GUlFlhCE HA5 ELEh 155UED lh TFE F0Lii 0F GEhEEl0 LETTEF5, hREG5, EEHLib, EULLEilh5, ETC.
I, umber Title Resolution Branch Application 25 Auton.a.
Air Header Dump on BWR Crder dtd SRXB BWR Scram System 1/9/81 40 Safety Concerns Associated with NUREG-0803 SRXB CWR Pipe Breaks in the BKR Scram System 51 Proposed Peqirments for Improved GL-89-13 SPLB All Reliebility of Open Cycle SKS 67,03.3 leproved Accident ttonitoring GL 82-33 SICB All 86 Lcng Range Plan for Cealing with GL 88-01, EMCB BWR Stress Corrosion Cracking in El' REG-0313 EWR Piping S3 Steam Bincing of Auxiliary GL 88-03 SPLB PWR Feejwater Fumps 1EB 85-01 103 Desirn for Probable Maximum GL 89-22 ESGB All Prscipitatier A-02 Asyr.etric BicSdown Loads on NUREG-0609 EMCB PWR Rtactor Primary Coolant Systeri A 03-05 Stfam Generator Tube Integrity NUREG 0844 EMCB PWR A 10 Ett feedwater Nozzle Cracking NUREG 0619 EMCB CWR A-42 Pipe Cracks in FWRs GL 88 01 EMCB BWR HUREG-313 A-47 Safety Irplications of Control GL 89-19 51CB All Systems C-01 Assurante of Centinuous Long Term NUREG-05tP SPLB All Capalility of Herraetic Seals on Instrumentation and Electrical Eouipment II.E.6.1 Insitu Testing of Valves NUREG-1296 EMEB All GL 89-10 11.K.1(5)
Safety-Related Valve Position Hl' REG-0800 LHFB All Des cription
- 31. K.1( 2 0)
Review and Modify Procedures for NUREG 0800 LHFP All Removing Safety-Related Systems from Service II.K.1(13)
Pro:cse Technical Specification NUREG 0933 OTSB All C1anges Reflecting implementa-tien of All Bulletin items 11.K.1(20)
Provide Procedures and Training IEB 79-05B LHFB B&W to Operators for Prompt Manual Reactor Trip for LOTW, TT.
MSIV Closure LOOP, LOSG Level, and LO PZR Level l
3 RESOLVED GS1'S AND US1'S TO Et ADDRESSED IN EVOLUTIONARY LKP t! SIGN REY 1tv.'s FOR WhlCl,GUl[iAhCE hA5 EEEh 155M D lh ThE F0iy 0F GEhERIC LETTEF.5,
_hUEEG5, EEiDET5, EULLETIh5, ETC. (EGNTINUED) 1urber Title Resolution Branch Applicaticti 21.K.1(21)
Provide Automatic Safety Grade lEB 79 05B LHTBB B&W Anticipatory Reactor Trip for LOTW, TT, or Significant Decrease in Steam Generator Level 22.K3(22)
Describe Automatic ! Panual IEE 79-08 SPLB BWP Actions for Proper function-ing of Auxiliary Heat Removal System when Feed-water System not Operable 11.K.1(23)
Describe Uses & Types of RV IEB 79 08 SRXB BWP.
Level Indicated for Automatic
$1CB and P.arval Initiation of Safety Systt 22.K.1(28)
Provide Design that will Assure NUREG.0800 SRXD PWR Autorcatic RCP Trip for All Circur. stances where Required 1
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Enclosurt.K GSI'$ AND l'51'S TO BE 1.DDRESSED 1h EVOLUT10N#1Y LWR DESIGt: PEVIEWS F0t WH]CH l
7mviIch E;.5 FisvLTip li. TFE 155Ula.cE OF A F.ULE, CT TFE DEviLOTVENT OF A 5Ti!.i/it REYlEn FLAli FEV151Cf, EEGULAT0E) GUIDE. OF EE\\l510h
~
i TO A EEGut ATOM GL10E'
_ Number Title Ptsolution Branch Applicetion
)
45 IriePerability of Instrurentation Due SRP-7.1.5,7 S!CB All to Extreme Cold Weather IE4 Auxiliary Feed *ater System Reliability SRP 10.4.9 SPLD PWR A.C1 Water Herrner SRP 3.9.3 EMEB All A 07 l' ark I Long Term Program (fortner USI)
SRP 6.2.1.1c SPLB BWR NUREG 0661 Supp. 1 A 0B Mark !! Containment Pool Dyramic SRP E.2.1.1c SPLE EWR Loads - Long Term Program NUREG 0B08 EliEB A 09 ATWS Rule 50.(2 SICB All A-12 Fracture Toughness of Steam Generator SRP 5.3.4 EMTB PWR j
and Reactor Coolant Pump Supports j
A.13
$ nutter Operability Assurance SRP-3.9.3 EMEB All A 24 (n,alification of Class 1E Safety Rule 50.49 SPLB All 4
Related Equipment A 25 Non Sefety Loads on Class IE Power R.G.-l.75 SELB All i
Sources A 26 Reactor Vessel Pressure Transient 3RP 5.2 EttEB PWR Protection A 31 RHR Shutduwn Pequirements SRP-5.4.7 SRXB All l
l A*35 Acequacy of Offsite Power Systems SRP-B.3.1 SELB All l
A 3E Control of Heavy Loads Near Spent SRP 9.1.5 SPLB All Fuel Pool A.39 Determination of SRV ?ool Dynamic SRP-6.2.1.1c $PLB BWR Loads &TemperatureLimits(in BWRContainment)
A-40 Seismic Design Criteria - Short Term SRP-2.5.2 ESGB All Program SRP 3.7.)
i SRP-3.7.2 SRP-3.7.3 l
[ 43 Containment Energency Sump Perfor-SRP 6.2.2 SRXB All I
mance R.G. l.B2 i'
A 44 Station Blackout Rule 50.63 SELB All P.G. 1.155 A-4B H Control tieasures & Effects of H Rule 50.44 SPLB All 2
g Burns on Safety Equipment A.49 Pressurized Thermal Shock Rule 50.65 EMEB PWR R.G.-l.154 B.10 Behavior of Et!R Mark III Containments SRP 6.2.1.1C SPLB BWR TTor brevity, Titl-2 issues resolved through NUREG-0737 efforts are not included on list. These items are handled in the review of the design through the SRP. THI 2 issues listed resulted from NUREG 0660 efforts.
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[NCLOSURE 2 G5!_'s AM 051'5 TO PE ADUISSED 1h EY0tVT10NAtY tWR 0[51GN FtY1EWS F0P UhltH
~TE 5 sGJ ; 0 N 95 F E 5 a T E D J t. T EE 155 U AhM F A QL E. 07EUE L 0f lie hT O F A Su';;;iD EEi]Eu rLitrTgy;57pn ;,gMU DTTTROTTE h 15I0h TO A TE LL AldUIDE (EU*lTMEU*
huder Title Resolution Branch Applicatiet B 3f tevelop Design, Testirs & Maintenance R.G.-l.140 SPLB All Criteria for Atacsthere Cleanup R.G.-1.52 System Air Filtrt. tion and Adsorption Systems and for Normal Ventilation Systems L-E3 Isolation of Low Pressure System SRP 3.9,0 SPLB All Connected to PCF0 B EC Control Room Infiltration Peasurements SRP 9.4.1 PRPB All I.17 Interim Acceptente Criteria for Solid Rule 01.56 SPLB All Agents for Radioettive Wastes 1.A.1.4 Lcng Term Upgrading of Operating Rule 50.54 LHFE All F'ersonnel ar.o Staffing 1.A.4.1(2)
Interim Charges in Training Simulators R.G. l.149 LHFB All 1.A.4.2(1) Research on Training Simulators R.G. l.149, LHFB All Rev.1 1.A.4.2(2) Upgrade Training Simulator Standards R.G.-l.149 LHFB All 2.A.4.2(3) Regulatory Guide on Training R.G.-l.149 LHFB All Simulators 1.A.4.2(4) Review Simulators for Conferrance to Rule 55.45 LHFB All Criteria 1.C.9 Long Term Progrer., Plan for Upgrading SRP 13.5.2, LHFB All of Frocedures Rev. 1 SRP-13.5.2, App. A 1.D.5(2)
Plant Status & Post-Accident R.G.1.97,
$1CB All Monitoring Rev. 2 1.F.2(2)
Include QA Fersonnel in Review &
SRP 17 LQAB All Approval of Plant Procedures 1.F.2(3)
Include QA Personnel in all Design, SRP-17 LQAB All Construction. Installation, Testing end Operational Activities 1.F.2(0)
Increase the Size of the Licensee's QA SRP 17 LQAB All Staff 1.F.2(9)
Clarify Organization Reporting Levels SRP 17 LQAB All for the QA Organization 1.G.2 Scope of Preoperational & Low-Power SRP-14 LQAB All Testing Program 11.B.8 Rulemaking Proceeding on Degraded Rule 50.34(f) SPLB All Core Accidents 11.E.1.3 Update SRP and Develop Regulatory SRP-10.4.9 SPLB All Guidance - Auxilary Feedwater System l
l
'For brevity TMI-2 issues resolved through NUREG 0737 efforts are not included on list.
These items are hat,dled in the review of the design through the SRP. TMI-2 f
issues listed resulted from NUREG-0660 efforts.
7-_ _. _ ____________ _ _ _
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j ENCLOSURF. 2
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i G5!'S Atit U$1'$ TO BE ADDRESSED lt; EVOLUT!0t;ARY LL'R DESIGl? REVIEWS FOR WHICH I
_FE5DLUT10h hAE FE5ULTED lh lhE 1550AhCE OF A RULE, 0h ThE DE\\ELOFMENT OF A STAhDAED EEVIEel FLAh EEV1510f;. EEGUL ATORY GUIDE, OR EEVISION 10 A FEGULATOEt GUIDE (C0hTIKUED)*
Hurter Title Resolution Branch Arrlication 11.F.3 Instruments for Monitoring Accident R.G.I.97,
$1CB All Conditions Rev. 2 i
!!!.D..
Set Criteria Requiring Licensees to R.G.I.97, PRPB All 3.3(2)
Evaluate Need for Additional Rev. 2 Survey Equipment SRP 12.3 SRP-12.5 111.D..
Issue Rule Chatige Providing Acceptable Rule.
FAPB All 3.3(3)
Hethods for Calibration of 20.501(c)
Fadiation Monitoring Instrumenta-1 tion 111.D.-
Issue a Regulatory Guide R.G.-B.25 PRPB All 3.3(4)
HF1.1
$hift $taffing R.G.-l.114, LHFB All Rev. 2 4
i SRP-13.1.2 2
1 iPor brevity. THI ? 1ssues resolved through NUREG-0737 efforts are not included on
-list. These itec<s are handled in the review of the design through the SRP, TMi-2 issues listed resulted from NUREG-0660 efforts.
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