ML13317A611
| ML13317A611 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Dresden, Palisades, Oyster Creek, Haddam Neck, Ginna, San Onofre, Yankee Rowe, La Crosse, Big Rock Point  |
| Issue date: | 05/07/1981 |
| From: | Lainas G Office of Nuclear Reactor Regulation |
| To: | COMMONWEALTH EDISON CO., CONNECTICUT YANKEE ATOMIC POWER CO., CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.), DAIRYLAND POWER COOPERATIVE, JERSEY CENTRAL POWER & LIGHT CO., NORTHEAST UTILITIES, ROCHESTER GAS & ELECTRIC CORP., SOUTHERN CALIFORNIA EDISON CO., YANKEE ATOMIC ELECTRIC CO. |
| References | |
| TASK-***, TASK-RR NUDOCS 8105130109 | |
| Download: ML13317A611 (42) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 MAY o~s LETTER TO ALL SEP LICENSEES US' Gentlemen:
SUBJECT:
DELETION OF SYSTEMATIC EVALUATION PROGRAM TOPICS COVERED BY THREE MILE ISLAND NRC ACTION PLAN, UNRESOLVED SAFETY ISSUES, OR OTHER SEP TOPICS Topics in the Systematic Evaluation Program (SEP), Phase II, that are being implemented as part of the Three Mile Island (TMI) NRC Action Plan, or Unresolved Safety Issues (USIs), or are duplicated in part by another SEP topic, are being deleted from the SEP to minimize dupli cation. Enclosure (1) is the list of SEP.topics being deleted and their related TMI, USI, or other SEP topic reference. Enclosure (2) is the original SEP topic definition and the basis for our determination that the SEP topic review can be deleted from the SEP program. NUREG-0485, the SEP Summary Status.Report, will be revised to delete those topics identified in Enclosure (1).
Enclosure (3) is the list of SEP topics not included in Enclosure (1) which have previously been identified as generic. The ongoing generic activity related to each of these topics is also identified (i.e., the multi-plant generic activity number and title or the NRR generic activity number and title).
The NRC review of the issues identified in Enclosure (1) will be per formed by the staff responsible for the TMI Action Plan item or the Unresolved Safety Issue. The review and implementation of TMI Action Plan items and USIs are being conducted for all operating reactors separate from the SEP program. Since a number of TMI issues, as well as USIs, will be resolved during the same time frame as the completion of our assessments, the staff will consider the status and corrective actions for TMI and USI items and will, to the extent practicable integrate them into our overall assessment. This would assure that corrective actions required as a result of the SEP Integrated Assessment are coordinated to the extent possible with TMI and USI requirements and not unnecessarily impact plants.
1S
/75 64, IT~
Of,
-2 The topics identified in Enclosure (3) rely upon the completion of the related generic activity. Many of these related generic reviews are complete and Safety Evaluation Reports have been issued.
Each licensee will be informed by separate correspondence of the status of Enclosure (3) topics and what further action, if any, is requested. The results of these generic topic reviews, i.e., Enclosure (3) topics, will be included in the Integrated Assessment.
Sincerely, Gus C. Lainas, Assistant Director for Safety Assessment Division of Licensing
Enclosures:
As stated cc w/enclosures:
See next page
ENCLOSURE 1 SEP TOPICS BEING DELETED SEP TMI, USI, or TOPIC NO.
TITLE 11-2.8 Meteorological Measurements Program TMI II.F.3 Instruments for Accident Conditions TMI III.A.1 Emergency Preparedness -
Short Term II-2.D Meteorological Data in Control Room TMI II.F.3 Instruments for Accident Conditions TMI III.A.1 Emergency Preparedness - Short Term TMI I.D.1 Control Room Design Reviews II-8.D Core Supports & Fuel Integrity USI A-2 Asyiiunetric Blowdown Loads 111-9 Support liteyrity USI A-12 Steam Generator & Reactor Coolant Pump Support USI A-7 Mark I Containment USI A-24 Qual. of Class IE Equipment USI A-46 Seismic Qualification SEP 111-6 Seismic Design Considerations SEP V-1 Codes and Standards III-11 Component Integrity USI A-46 Seismic Qual. of Equip. in Operating Plants SEP 111-6 Seismi-c Design Considerations USI A-2 Asynmetric Blowdown Loads 111-12 Environimental Qual. of Safety Equip.
USI A-24 Qual. of Class 1E Equipment V-3 Overpressurization Protection USI A-26 Reactor Vessel Pressure Transient Protection V-4 Piping & Safe End Integrity USI A-42 Pipe Cracks in BWRs V-8 Steam Generator Integrity USI A-3,4,5 Steanm Generator Tube Integrity V-13 Water Hammer USI A-1 Water Haiimier
-2 SEP TOPIC
TITLE VI-2.A Pressure-Suppression Type 8WR Containments USI A-7 Mark I Containment VI-2.B Subcompartment Analysis USI A-2 Asymmetric Blowdown Loads VI-5 Combustible Gas Control TMI II.B.7 Analysis of Hydrogen Control USI A-48 Hydrogen Management VI-7.E ECCS Sump Design USI A-43 Containment Emergency Sump Performance VI-8 Control Room Habitability TMI I1. J.3.4 Control Room Habitability VII-4 Effects of Failure in Non-Safety Related Systems on ESF USI A-47 Safety Implications of Control Systems USI A-17 Systems Interaction VII-5 Instruments for Radiation and TMI II.F.1 Additional Accident Instrumentation Process Variables During TMI II.F.2 Inadequate Core Cooling Accidents TMI II.F.3 Instruments for Accident Conditions IX-2 Overhead Handling Systems (cranes)
USI A-36 Control of Heavy Loads Near Spent Fuel Pool X
Auxiliary Feedwater System TMI II.E.1.1 Auxiliary Feedwater System Evaluation XIll-1 Conduct of Operations TMI I.C.6 Correct Performance of Operating Activities TMI III.A.1 Emergency Preparedness -
Short Term TMI III.A.2 Emergency Preparedness -
Long Term
TITLE XY-21 Spent Fuel Drop Accidents USI A-36 Control of Heavy Loads Near Spent Fuel Pool XV-22 Anticipated Transients Without Scram USI A-9 Anticipated Transients Without Scram XV-23 Tube Failures in Steam Generators USI A-3,4,5 Steam Generator Tube Integrity USI A-9 Anticipated Transients Without Scram XV-24 Loss of all AC Power USI A-44 Station Blackout Enclosure
ENCLOSURE 2 DEFINITION TOPIC:
11-2.5 Onsite Meteorological Measurements Program
- 1. Definition:
To review the onsite meteorological measurements program to determine the extent that the licensee complies with 10 CFR Part 50, Appendix E, and Appendix 1.
- 2. Safety Objective:
To assure that adequate meteorological instrumentation to quantify the off-site exposures from routine releases is availaole and maintained.
- 3. Status:
Onsite meteorological measurements programs are being reviewed as a part of the Appendix I evaluations.
- 4.
References:
- 1. 10 CFR 50, Appendix E, Appendix I
- 2. R. G. 1.97, Rev. 1
- 3. R. G. 1.23
- 4.
SRP Section 2.3.3
- a. TMI Task Action Plan -
Task II.F.3, Instrumentation for Monitoring Accident Conditions Task II.F.3 requires that appropriate instrumentation be provided for accident monitoring with expanded ran7es and a source term that considers a damaged core capable of surviving the accident environment in which it is located for the length of time its function is required.
Regulatory Guide 1.97, Revision 2, "Instru mentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Condition During and Following an Accident,"
issued December 1980, contains :he required meteorological instrumentation to quantify the off-site exposure.
- b. THI Task Action Plan - NUREG 0660 -
Task III.A.1, Improve Licensee Emergency Preparedness -
Short Term Task III.A.1 requires the evaluation of 10 CFR Part 50, Appendix E backfiL requirements in accordance with NUREG 0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plan and Preparedness in Support of Nuclear Power Plants."
Backfit requirements include review of the Onsite Meteorological Measurement Program.
The evaluation required by Task II.F.3 and II-.A.1 are identical to SEP Topic II-2.B; therefore, this SEP topic has been deleted.
0 DEFINITION TOPIC:
II.2.0 Availability of Meteorological Dat. in the Control 'Room 1.Deflnition:
Data from the onsite meteorological program should be available in the control room.
- 2. Safety Objective:
To assure that the licensee has appropriate meteorological logical data displayed in the control room to assess conditions during and following an accident to allow for:
(1) early indication of the need to initiate action necessary to protect portions of the off site public; and (2) an estimate of the magnitude of the hazard from potential or actual accidental releases.
- 3. Status:
No wort currently being done on this subject for operating plants.
- 4.
References:
- 1. 10 CFR 50, Appendix E, Appendix I
- 2. R. G. 1.97, Rev. 1
- 3. R. G. 1.23
- 4.
SR? Section 2.3.3
- a. TMI Task Action Plan -
NUREG-0660 - Task II.F.3 Task II.F.3,"Instrumentation for Monitoring Accident Conditions" requires that appropriate instrumentation be provided for accident monitoring with expanded ranges and a source term that considers a damaged core capable of surviving the accident environment in which it is located for the length of time its function is required.
Regulatory Guide 1.97, Revision 2 "Instrumentation for Light-Water Coolant Nuclear Power Plants to Assess Plant and Environs Conditions during and Following an Accident" issued December 1980, contains the required meteorological instrumentation to quantify the off-site exposure.
- b. TMI Task Action Plan - NUREG-0660 -
Task III.A.1 Task III.A.1,"Improve Licensee Emergency Preparedness Short Term" requires the evaluation of 10 CFR Part 50, Appendix E backfit requirements in accordance with NUREG 0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plan and Preparendess in Support of Nuclear
g 0
-2 Power Plants".
Backfit requirements include review of the Onsite Meteorological Measurement Program.
- c. TMI Task Action Plan - NUREG-0660 - Task I.D.1 Task I.D.1, "Control Room Design Reviews" requires that operating reactor licensees and applicants for operating licenses perform a detailed control room design review to identify and correct design defici encies.
This review will include an assessment of control room layout, the adequacy of the information provided, the arrangement and identification of important controls and instrumentation displays, the usefulness of the audio and visual alarm systems, the information recording and recall capability, lighting, and other considerations of human factors that have an impact on operator effect iveness.
The evaluations required by Tasks II.F.3, III.A.1 and I.D.1 are identical to SEP Topic II-2.D; therefore, this SEP Topic has been deleted.
OEFINITION TOPIC:
111-8.0 Core Supports and Fuel Integrity
- 1. Definition:
Abnormal loading conditions on the core supports andd fuel assemblies due to seismic events or LOCAs could cause fuel damage due to impact between fuel assemblies and uper and lower grid plates or lateral impact between fuel assemblies and the core baffle wall.
The resulting damage could result in loss of coolable heat transfer geometry, make it impossible to insert control rods, or cause releases of radioactive materials due to fuel pin failure.
- 2. Safety Objective:
To assure that all credible loading conditions on core supports and fuel assemblies will not result in unacceptable fuel damage or distortion.
- 3. Status:
DOR is currently reviewing the dynamic loads imposed on the fuel assemblies during a LOCA. Independent analyses are being conducted by staff consultants.
- 4.
References:
USI A-2, Asymmetric lowdown Loads on Reactor Primary Coolant System,.iNUREG-0649 USI A-2 requires that an analysis be performed by licensees to assess the design adequacy of the reactor vessel supports and other structures to withstand the loads when asymmetric LOCA forces are taken into account. The staff has completed its investigation and concluded that an acceptable basis has been provided in NUREG-0609, "Asymmetric Blowdown Loads on PWR Primary Systems," January 1981, for performing and.reviewing plant analyses for asymmetric LOCA loads.
The structural accept ance criteria specified in NUREG-0609 are as follow:
-2 The structural integrity of the primary system including the reactor pressure vessel, RPV internals, primary coolant loop, and components must be evaluated against appropriate acceptance criteria to determine if acceptable margins of safety exist. Allowable limits and appropriate loading combinations are set forth in standard review plans (SRPs), which are listed in the table that follows.
The staff recognizes that in some specific cases, where "as built" designs are being reevaluated for asymmetric LOCA loads, these design limits may be exceeded. Acceptance of alternative allowable limits will be based on a case-by-case evaluation of the safety margins.
Load combination criteria in general were not addressed as part of this study.
Currently the staff requires that seismic (SSE) and LOCA response be combined, along with responses due to other loading as specified by the SRP.
An accept able method for combining elastically generated seismic and LOCA responses is provided in NUREG-0484.
Acceptable methods for combining response generated by an inelastic LOCA analysis and elastic seismic analyses will be evaluated on a case-by-case basis.
Item SRP Reactor pressure vessel 3.9.3 Reactor internals 3.9.5, 3.9.1 Primary coolant loop piping 3.9.3 ECCS piping 3.9.3 RPV, SG, pump supports.
3.8.3 Biological shield wall 3.8.3 Steam-generator compartment wall 3.8.3 Neutron-shield tank 3.8.3 Since USI A-2 also requires the investigation of seismic and LOCA response be combined, the evaluation required by USI A-2 is identical to SEP Topic II1-3.0; therefore, this SEP ToPic has been deleted.
gIlk DEFINITION TOPIC:
111-9 Support Integrity
- 1. Oefinition:
Review the design, design loads, and materials integrity including corrosion and fracture toughness and the inservice Inspection programs of supports and restraints including bolting for the reactor
- vessel, steam generator, reactor coolant pump, torus and other class 1, 2 and 3 safety related components and piping systems.
2.. Safety Objective:
-To assure adequate support and/or restraint of safety related systems and comoonents under normal and accident loads so that they will not be prevented from performing their intended functions because of support failures.
- 3.
Status:
OR has ongoing programs to review component supports.
Current emphasis is on primary system supports and on piping system supports and restraints (snubbers).
- 4.
References:
- 1. ASMIE Section III
- 2. Pink Book Generic Topics 3-5 and 3-43
- a. USI A-12, Fracture Toughness of Steam Generator and Reactor Coolant Pump Supports, NUREG -0510, 0606 The original scope of USI A-12 was the review of the steam generator and reactor coolant pump supports of. pressurized water reactors. However, the staff has exoanded the review to include other support structures, such as boiling water reactor (BWR) vessel supports, BWR pump supports, pressurized water reactor (PWR) vessel supports and PWR pressurizer supports (NUREG-0577, Section.1.3).
This expanded review will be undertaken in accordance with the guidance of Section 4 of NUREG-0577.
- b. USI A-7, MARK I Containment Long-Term Program, NUREG-0649 Support integrity of the Torus is being evaluated under USI A-7. Under this task, a short 'term program that evaluated Mark I containment has provided assurance that the Mark I containment system of each operating BWR
-2 facility would maintain its integrity and functional capability during a postulated LOCA. A longer term program for BWR facilities, not yet licensed, is planned wherein the NRC staff will evaluate the loads, load combinations, and associated structural acceptance criteria proposed by the Mark I Owners group prior to the performance of plant-unique structural evaluations.
The Mark I Owners group has initiated a comprehensive testing and evaluation program to define design basis loads for the Mark I containment system and to estab lish structural acceptance criteria which will assure margins of safety for the containment system which are equivalent to that which is currently specified in the ASME Boiler and Pressure Vessel Code. Also included in their program is an evaluation of the need for structural modifications and/or load mitigation devices to assure adequate Mark I containment system structural safety margins.
- c. USI A-24, Environmental Qualification of Safety Related Equipment, NUREG-0371 Snubber operability and degradation of seals is covered under USI A-24.
- d. USI A-46, Seismic Qualification of Equioment in Operatina Plants, NUREG-0705 Mechanical snubbers are covered under USI A-46.
- e.
SEP Topic III-6, Seismic Design Considerations Snubbers are evaluated for capacity under SEP Topic 11-6.
- f. SEP Topic V-1, Codes and Standards Inservice Inspection requirements for supports is covered under SEP Topic V-1, which refers to 10 CFR 50.55a.
SEP olants currenJy have surveillance Technical Specifications on snubbers.
The evaluation required by USI A-12, A-7, A-24, A-46, SEP Topics 111-6 and V-1 is identical to the evaluation reauired by SEP Topic 111-9; therefore, this SEP topic has been deleted.
DEFINITION TOP!C:
111-11 Component Integrity
- 1. Definition:
7eview licensee's criteria, testing Procedures, and dynamic analyses employed to assu.re the structural integrity and functional operability of safety related mechnanical equioment uncar faulted conditions and accicent l0acs.
Incudec are mecnanical eupment such as pumps, valves, fans, pume irives, heat exchancer tube buncles, valve actuatrs, ba:tery and instrument racks, control consoles, cabinets, panels, and cable trays.
- 2.
Safety CbJective:
To confi r the anility of safety related mechanical equipment having experienced proolems to function as neeced during and after a faulted or accident condition.
The ca:ability of safety related mechanical ecuipment to perfor-necessary protective actions is essential for plant safety.
- 3. Status:
n7s review is not currently uncerway in DOR.
- 4.
References:
- 1. 10 ;CF, EU.E-I:a
- 2.
10 CE O, opendix A, GDC 2, 4, 14, 1i
- 3. Standard Review Plan 3.3.2
- 4.
A5zE Sec-on III S. Cegulatory Guides 1.20 and 1.58
- 7. StanddReview Plan 3.9.3
- 5.
Basis for Deletion (i.e., related TMI Task, USI or other SEP Topic):
- a. USI A-46, "Seismic Oualification of Equioment in Operating Plants" - NUREG-0606, 0705 The component integrity (both structural integrity and functional operability) for safety related mechanical and electrical equipment for all operating plants including SEP plants will be addressed in this new USI (A-46).
- b. USI A-2, "Asymmetric Blowdown Loads on Reactor Primary Coolant System", NUREG-0649 The assessment of faulted loads for the primary loop are being performed under USI A-2.
Further, the assessment of high energy pipe breaks consider the effect of accident loads with regard to jet impingement, pipe whip and other reaction loads.
-2
- c. SEP Topic 111-6, Seismic Design Considerations The evaluation of equipment structural integrity under seismic loads will be performed under SEP Topic 111-6.
The evaluations required by Tasks USI A-46, A-2, and SEP Topic I11-6 are identical to SEP Topic III-11; therefore, this SEP Topic has been deleted.
DEFINITION TOPIC:
111-12 Envirornmental Qualification of Safety-Related Equipment
- 1. 1,efi hi ti on:
Safety-related electrical and mechanical equipment that is required to survive and function under environmental conditions calculated to result from a loss-cf-coolant accident (LOCA) or a postulated main steam line oreak (MSL3 ) accident inside containment must De environmentally cualified.
1n addition, determine whether environment induced failures of non-safety-related equipment could interfere with the operation of safety equipment. Special attention should De given to the effect of beta radiation on exposed organic surfaces, such as gaskets.
- 2. Safety Objective:
To assure that the mechanical and Class T-electrical equipment of safety systems have been oualified for the most severe environment (temoe-ature, oressure, humidity, chemistry and raciation) of design basis accidents.
- 3. Status:
'4estinchouse is conducting a verification Drogram wnich is expected to De completed Oy the enc of 1977 for those plants cualifiec to IEE -
323 (1971).
The Office of Nuclear Regulatory Research (RES) is sponsoring programs relating to Class :-
eouipment 3ualification, the results of which can De utilizeo to oetermine the adequacy of the equipment pre viously qualified.
- 4.
References:
1..
NUREG 0153, Item 25, "Qualification of Safety-Related Equipment" Decemoer 1976
- 2. OR Technical Activities, Category 3, Item 34, "Envirornmental Qualifications of Safety-elated Equipment (Post LOCA)",
May 1977
- 3. 055 Technical Activities, Category A, Item 33, "Qualification of Class ?E Safety-;elated E;uipment, April 1977
- 4. R. G. 1.39
- 5.
Basis for Deletion (i.e., related TMI Task, USI or other SEP Topic):
USI A-24, Qualification of Safety Related Equioment, NUREG-0371, NUREG-0606 The issue identified in reference 1 (NUREG-0153, Item 25) and the review criteria, i.e., R.G. 1.89, are identical to those speci fied in USI A-24. The Task Action Plan for USI A-24 (NUREG-0371) covers the environmental qual ification of both electrical and mechanical safety related equipment.
The evaluation required by USI A-24 is identical to SEP Topic III-12, therefore, this SEP Topic has been deleted.
DEFINITION TOPIC:
V-3 Overpressurization Protection
- 1. Definition:
inadvertent overOressuriatifon of the primary system at temperatures below the nil ductility transition temperature may result in reactor vessel failure curing heatup and press ure transients are caused by pressure surges when the primary system is water solid.
The most severe transients nave occurred wnen a charging pump starts uo or inadvertent closing of 4 letdown valve with a charging pume running. Pressure temperature limits as a function of neutron fluence of the material at tne reactor vessel beltline are specified in 10 CFR SO, Aopendix G. All PR licensees have been directed to institute interim administrative *rocedures to prevent damaging pressure transients and on a longer time scale to provide permanent protection which will procably include hardware changes such as high capacity safety/relief valves.
- 2.
Safety Objective:
To protect the primary system from potentially damaging Overpressurization transients during plant oressurization and neatup.
- 3.
Status:
Generic review of all PWR licensee suomittals is unde-way.
Criteria for evaluation have Been develooed and refined by NRR/RES.
An effort is teiin made to comelete the review sufficiently early to ensure installation of mitigating systems oy the enc of 1977.
- 4. ;eferences:
- 1.
4UREG 0138
USI A-26, Reactor Vessel Pressure Transient Protection (NUREG-0410)
Under USI A-26 licensees were requested to modify their systems and procedures to protect against low temperature overpressuriza tion.
All ooerating PWRs have made these modifications and Safety Evaluation Reports for the SEP plants have been issued.
The evaluation required by USI A-26 is identical to SEP Topic V-3; therefore, this SEP topic has been deleted.
OEF IN IT ION TOPIC:
V-4 Piping and Safe Ena Integrity
- 1.
Oefinition:
Review the safety aspects that affect BWR and PWR piping and safe end integrity for compliance with 10 CFR Part 50, including fracture tougnness, flaw evaluation, stress corrosion cracking in SWR and PWR piping, and control of materials and welding.
- 2.
Safety Objective:
To assure continued piping integrity and compliance with lU CFR Part 50 ano applicable industry codes and standards.
The Engineering Brancn, DOR, is conducting an ongoing program that includes tne as-needed review of those aspects necessary to ensure tne continuing integrity of piping systems important to safety including stress corrosion cracking of SWR colant pressure coundary piping.
This program will continue for the life of operating reactors.,
- 4.
References:
- 1. Technical Position, Material Selection and Processing wuicelines for S'R Coolant Pressure Boundary Piping
- 2.
- a. USI A-42, Pipe Cracks in Boiling Water Reactors, NUREG-0510 The scope of USI A-42 is the study of stress corrosion cracking in BWR piping.
NTUREC-0313, Rev. 1 "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping" is the resolution of USI A-42 and presents staff positions.
- b. USI A-10, BWR Feedwater Nozzle Cracking and Control Rod Drive Hydraulics Return Line Nozzle Cracking, NUREG-0649.
C. NRR Generic Activitv C-7, PWR System Piping, NUREG 0471 The scope of this activity is the study of stress corrosion cracking in PWR piping. NUREG 0691, "ITnvestigaction and Evaluation of Cracking Incidents in Piping in Pressurized Water Reactors" recommends the same corrective actions (pg. 2-12) proposed for BWRs in UREG 0313, Rev.
1,,USI A-42.
The evaluation required by USI A-42 and Task C-7 is identical to the evaluation required by SEP Topic V-4; therefore, this SEP topic has been deleted.
0 0
DEFINITION TOPIC:
V-8 Steam Generator (SG) Integrity 1, Definition:
Review the safety aspects affecting operation of steam generators including secondary water chemistry, tuoe plugging criteria, inservice inspection, 'ossibly including a dimensional inspection for proper evaluation of denting, steam generator tube leakage, tube denting, flow induced vibration of steam generator tunes, tuoe repair, and tube bundle or steam generator replacement.
- 2. Safety Objective:
To ensure that acceptable levels of integrity of that portion of the reactor coolant pressure boundary made up oy the steam generator are maintained in accordance with current codes, standards, and/or regulatory criteria during normal and postulated accident conditions.
The integrity of the steam generator is needed to ensure that leakage following a postulated design basis accident will not result in doses to the public in excess of 10 CFR Part 100 guidelines and that the emergency core cooling systems will be aole to perform their safety functions.
- 3.
Status:
Review of this topic is being performed oy the Division of Operating Reactors. This effort will continue for the life of operating reactors.
- 4.
References:
- 1. Regulatory Guide 1.83 (Revision 1)
- 3. 10 CFR 50, Appendix A, GDC 30 and 32
- 4. Pink Sook 3-27
- 5.
Basis for Deletion (i.e., related TMI Task, USI or other SEP Tooic):
USI A-3, A-4, A-5, Westinahouse, Combustion Engineering, and Babcock and Wilcox Steam Generator Tube Integrity, NUREG-0649 The definition of this topic and the references cited are covered by USI A-3, 4 and 5.
The evaluation for USI A-3, 4 and 5 is identical to SEP Topic V-8; therefore, this SEP Topic has been deleted.
DEFINITION TOPIC:
V-13 Water Hammer
- 1. Definition:
Water hammer events have occured in light water reactor systems.
Vater hammer events increase the probaoili:y of pipe breaks and could increase the Consequences of certain events such as the loss of coolant accident.
The types of water hammer, the vul neraol e systems ( for examol e, contain ment spray, service water, feedwater and stem) and the safety signifi cance of water hammer have been identified an cefined in a staff report of May 1977.
- 2.
Safety Objective:
To reduce the probability of water hammer events that have the potential to lead to pipe runtures in L R systems which are needed to mitigate the consequences of accidents or that miont increase the consequences of accidents previously analyzed.
- 3.
Status.
Generic review is underway.
On March 10, 1977, an interdivisional DCP./0SS technical review group was formed to investigate the water hammer issue and to develop a program for its appropriate considera tion in Hicensing reviews and for operating reactors. Consultant work has oeen Performned :y CREAM and Livermore Labs.
- 4.
References:
- 1. "Water Hammer in Nuclear Power Plants", N;C Staff ;eport, June 1, 1977
- 2.
"An Evaluation of PWR Steam Generator Water Hammer" by G. 3. Wallis, P. H. Rothe, et. al. of CREARE Inc., draft, February 1977.
- 3.
Lawrence Livermore Laboratory "An Investigation of Pressure Transient Propagation in Pressurized Water Reatcr Feedwater Lines" (Preliminary)
S. 3. Sutton, April 15, 1977.
- 4.
WRR Technical Activities, Category A, Item 1, Water Hammer, Kay 1977.
- 5.
Basis for Deletion (i.e., related TMI Task, USI or other SEP Tooic:
USI A-1, Water Hammer, NUREG-0649 The references cited in this topic were the precursors of USI A-1.
The evaluation required for USI A-1 is identical to SEP Topic V-13; therefore, this SEP topic has been deleted.
DEFINITION TOPIC:
YI-2.A Pressure-Suppression Type SWR Containments
- 1.
Definition:
SWR pressure-suppression type contaiments (e.g.,
Mark I contaiineat) are subjected to hydrodynamic loads during the blowdown phase of a LOCA.
Those loads have the potential for damaging the components and structures (wetwell, internal structures, restraints, supports and connected systems) of the containment. During a relief valve blowdown into the suppression pool the wetwell (torus) shell ano safety/relief valve restraints may be over stressed. The hydrodynamic loads were not explicitly identified and included in the design of the Mark ! pressure-suppression containment.
- 2. Safety Objective:
To assure that the structural integrity of pressure suppression pool containments is maintained under hydrodynamic loading conditions.
It has teen determined that the upward forces during the blowdown phase following a LOCA ootentially cause tne Mark. torus to be lifted, causing failure of connecting systems and supoorts and leading to loss of the containment integrity.
Structural modifications and/or changes in.
the mode of operation might be necessary to assure adequate safety margins.
- 3. Status:
Mark I containments are currently evaluated in a two step generic review program:
The Short-Tern Program (STP),
completed May 1977, has focused on.he determination of the magnitude and significance of hydrodynamic loads. in the Long-7erm Program (LTP),
to be completed by late 1978, the design Dasis loads wiil be finalized and the capability of the containment to withstand the loads within the original design structural margins will be verified. This verification will be based in part on research results from NRC and industry sponsored programs. As a result of the STP, the staff required that Mark I plants be operated with a drywell to etwell differential pressure of at least one psi to reduce the vertical loads.
In addition some licensees have modified the torus support system for additional safety margin.
- 4.
References:
- 1.
Pink Bock - Generic issues (Aoril 1977)
- a. Mark 1 Containment - STP Technical Specifications
- b. Mark I.ontainment Evaluation
- c.
Mark I Containment Evaluation -
- d. Mark I Safety/Relief Valve Line Restraints in Torus
TOPIC YI-2.A
- 2
- 2.
DOR Technical Activities, Category A, April 1977
- a.
Item 2, wMark I Containment STP"
- 0. Item 3, "Mart I Containment LTP"
- c.
Item 23, "Mart II Containment'
- 3. DOR Technical Activities, Category 8, May 1977, Item 12,
'Assessment of Column Buckling Criteria"
- 4. OSS Technical Activities, Category A, April 1977, Item 31, Determination of LOCA and SRY Pool Dynamic Loads for Water Suppression Containments'
USI A-7, Mark I Containment Long-Term Program, NUREG-0649 Under this task, a short term program that evaluated Mark I containment has provided assurance that the Mark I containment system of each operating BWR facility would maintain its integrity and functional capability during a postulated LOCA. A longer term program for BWR facilities, not yet licensed, is planned wherein the NRC staff will evaluate the loads, load combinations, and associated structural acceptance criteria proposed by the Mark'! Owners group prior to the performance of plant-unique structural evaluations.
The Mark I Owners group has initiated a comprehensive testing and evaluation program to define design basis loads for the Mark I containment system and to establish structural acceptance criteria which will assure margins of safety for the containment system which are equivalent to that which is currently specified in the ASME Boiler and Pressure Vessel Code. Also included in their program is an evaluation of the need for structural modifications and/or load mitigation devices to assure adequate Mark I containment system structural safety margins.
The long term program for USI A-7 will assure that all plants with Mark I containments are able to tolerate, without loss of function, the LOCA induced hydrodynamic loads.
The evaluation required by USI A-7 is identical to SEP Tooic VI-2.A; therefore, this SEP topic has been deleted.
.DEFINITION TOPIC:
YI-2.3 Subcompar-ment Analysis
- 1.
Definition:
The rupture of a hih energy line inside a containment subcompartment can cause a pressure differential across the walis of the suocomoartmient.
in the case of a rupture of a ?WR main coolant pipe a Cracent to t,e reactor vessel, the suocooledt blowdown produces 'pressure differentials in the annulus between the reactor vessel and the shield wall and also w thin the reactor vessel across the core barrel.
Thi s asy~etric pressure distribution generates loads on the reactor vessel support and on reactor vesseT internals on other equipment supports and on Suompartment structures whiich have not been analyzed previously for most operating reactors.
- 2. Safety Objectve:
o assure tiat the reactor vessel supports, reactor vessel internals, o the r equipment supports and suocoartment structures are designec dan acequate margin against failure due to these loads.
The failure could result in a loss of EZCS capability.
- 3. Status:
The staff is reviewing t.he NSSS vendor and architect engineer design codes used to calc-ulate tme loads produce,: oy the asymmetric pressure distribution.
Analyses have Deen completed for a limited number of' operating plants.
7 he
-M 0~ code is ao *proved. 3echtel, Gibo ert and United Engilneering have submitted ces for review.
- 4.
References:
- 1. ?ink Book - Generic issue, Item 1-5, Asymmetric LOCA Loads -
PWR",
A2ril 1977
-o Asymmetric L OCA Loads
- 2. 00R Tec. nical Activities, Catecory A, item 32, "Asviretric LOCA Loads
(;eactor Vessel SucortProle)
Apr i 1977 7,ec.jnial Acv a~s, Ca-egorv
- ,Isretrlc Blowcown Loacs on ?eactor Yessel", Acril 7
t 1er o
w
'DP4 7aecnical Activities, Category A, Item 2, "Reactor Yessel Suooorts (Asymmetric LOCA Lo:ads frcm Succen Suocoolea.Bbowcown), Aoril 1977
- 5.
Basis for Deletion (i.e., related TMI Task, USI or other SEP Tooic):
USI A-2, Asymmetric Slowdown Loads on Reactor Priman Coolant System, NUREG-0649 The references cited in this topic were the precursors of USI A-2.
The evaluation required for USI A-2 is identical to SEP Tooic VI-2.8 (see also SEP Topic III-8.0); therefore, this SEP Topic has been deleted.
b OEFINITION TOPIC:
YI-S Combustible Gas Control
- 1. Definition:
Review the combustible gas control system to determine the capability of the system to monitor the combustible gas concentration in the containment; to mix comustible gases within the containment atmosphere; and to maintain combustible gas concentrations below the combustion limits (e.g., by recombination, dilution, or purging).
For facilities which share recombiners (portable) between units or sites, determine that the recembiners can be made available within a suitable time.
For facilities wnich utilize purging as a primnary means of combustible gas control, determine the radiological con sequences of the system operation. Reevaluate hydrogen production and accumulation analysis to consider (1) reduction of Zr/water reaction on the basis of five times the Appendix K calculation amount and (2) potential increases in hydrogen production. from corrosion of metals inside containment.
- 2. Safety Objective:
To prevent the formation of combustible gas explosive concentrations in the containment or in localized recions within containment, following a postulated accident; to assure that the radiological consequences of the system operation are acceptable.
- 3. Status:
Proposed '10 C-R 50.44 would permit a BWR licensee to propose an alternate comoustible cas control system in lieu of inerting. Four such proposals for containment atmosphere dilution (CAD) systems are currently under review, and the COGAP 1T computer code is being revised to perform the system evaluations.
- 4.
References:
- 1. Proposed Rule 10 CFR 50.44
- 2. DOR Technical Activities, Category A, :tem 8, Containment Purge Ouring Normal Operation", April 1977 S. O0R Technical Activities, Category A,Item i4, nerting Recuirements CAO", April 1977
- 4. Branch Technical Position r'2 5-2 E. Standard Review ?!an 6.2.5
-2
- 5.
Basis for Deletion (i.e., related TMI Task, USI or other SEP Tooic):
- a. TMI Task Action Plan -
NUREG-0560 - Task II.B.7, Analysis of Hydrogen Control As a result of TMI II.B.7 short and long term rulemaking to amend 10 CFR 50.44 has been initiated. The short term rulemaking (interim rule) requires that all Mark I and Mark II containments be inerted. It also required that the owners of all plants with other containments perform certain analyses of accident scenarios involving hydrogen releases and furnish the staff with a proposed approach for mitigating these hydrogen releases.
The longer term rulemaking will address both degraded core and melted core issues.
In the area of hydrogen control it will prescribe requirements that are appropriate for opera ting plants as well as for plants under construction.
- b. USI A-48, Hydrocen Control Measures and Effects of Hydrogen Burns on Safety Equipment, NUREG-0705 Under USI A-48 a Task Action Plan has been defined and is being developed that encompasses the concerns in the Definition and the Safety Objective of SEP Topic VI-5.
The evaluation required by TMI II.B.7 and USI A-48 is
.identical to SEP Topic VI-5; therefore, this SEP topic has been deleted.
DEFINITION TOPIC: VI-7 E ECCS Sump Design and Test for Recirculation Mode Effectivenes.s 1
Definition.
Following a LOCA in a PWR an emergency core cooling system (ECCS) automatically injects water into the system to maintain core cooling.
initially, water is drawn from a large supply tank. Water discharging from the break and containment spray collects in.the containment building sump.
When the supply tank has emptied to a predetermined level, the ECCS is switched from the "injection" mode to the "recirculation" mode.
Water is then drawn from the containment building sump.
ECC systems are recuired to operate indefinitely in this mode to provide cecay heat removal.
Certain flow conditions could occur in the sumo, which could cause pump failures.
These include entrained air, prerotation or vortexing and losses leading to deficient NPSH.
- 2. Safety Objective:
To confirm effective operation of ECC systems in the recirculation mode.
- 3. Status:
Confinnation through pre-operational testing is now recuired on all CPs.
Staff has been accepting scaled tests in lieu of pre-oo tests at OL stage. Some plants have required modification to achieve vortex control.
- 4.
References:
- 1.
RE Vortex Technology (RR)
- 2. Reg. Guide 1.79 para. b(2)
USI A-43, Containment Emergency Sumo Reliabilitv, NUREG-30, 0660 The definition of this topic and the references cited are covered by USI A-43. The evaluation for USI A-43 is identical to SEP Topic VI-7E; therefore, this SEP Topic has been deleted.
OEFINITON TOPIC:
VI-8 Control Room Habitability
- 1. Definition:
Control rooms in operating plants may not fully comply with General Design Criterion I9. This review should include, but not be limited to, analysis of the control room air infiltration rate, ventilation system isolaoility and filter efficiency, shielding, emergency breathing apparatus, short distance atmospheric dispersion, operator radiation exposure, and on-site toxic gas storage proximity.
- 2. Safety Objective:
To assure that the plant operators can safely remain in the control room to manipulate the plant controls after an accident.
- 3. Status:
DOR now reviews control room habitability in operating plants when related licensing actions (e.g., assessnent of BWR Containment Air Dilution system post-LOCA radiological impact) recuire it. OSE has a technical assistance contract with the National Bureau of Standards to measure the control room air infiltration rate at a few operating plants. These measurements will be used to gauge the conservatism of the assumed air infiltration rates currently used by NRC.
Some reviews are now in progress for plants we have reason to believe do not meet G. 0. Criterion 19 (S0NGS-1, Vermont Yankee, St. Lucie).
4 eferences:
- 1. SRP 6.4
- 2. 10 C.R.0, Appendix A, GDC 19
- 3.
,uclear Power Plant Control Poom Ventilat
_ystem Design for Meeting General Criterion 19", by K. G. Murphy and Dr. K. M. Camoe, Proceedings of the Thirteentn AEC Air Cleanino Conference
- 1.
G. 1.9E, Rev. I
- 5.
3asis for Deletion (i.e., related TMI Task, USI or other SEP Tooic):
- TMI Task Action Plan, NUREG-0737, Task III 0.3.4, Control Room Habitability Requirements The review criteria required by Task I11.0.3.4 (NUREG-0737, pg. 3-197) is identical to the review criteria specified in the Definition and References of SEP Topic VI-8; therefore, this SEP Tooic has been deleted.
DEFINITION TOPIC:
YII-4 Effects of Failure in Non-Safety Related Systems on Selected Engineered Safety Features
- 1.
Definition:
Pot.etial c=binations of tran-ients and icicants with failures of =a safety-related control systems were not specifically evaluated in the original safety analysis of currently operating reactor plants. Review the effects of control system malfunctions as initiating events for anticipated transients and also as failures concurrent with or subsequent to anticipated events or postulated accidents initiated by a different malfunction (e.g., the effect of the loss of the plant air system on the plan: control and monitoring system). A complete discussion is provided in reference 1.
- 2. Safety Objective:
To assure that any credible combination of a non-safety-related system failure with a postulated transient or accident will not cause unacceptable consequences.
- 3. Status:
A technical assistance contract with ORNL for failure mode analyses of control systems was initiated to determine sensitive areas of thne plant designs. The results of this program in conjunction with the results of the failure made and effects analyses for transients and accidents being performed under contract by INEL should provide a basis for any new review and safety requirements.
- 4.
References:
- 1.
NLREG-0153, item 22, "Systematic Review of Normal Plant Operation and Control System Failures", Decemoer 1975
- 2.
Memorancum from Y. Stello to R. J. Hart; dated 12/22/76, NRR letter No. 46.
- 3.
00R Task Force Report or SEP, Appendix 3 (TFL 113), Novemoer 1975
- a.
Item 33 "Safety Related Control Power"
- b.
item 24 " Safety Related Instrumentation Power'
- c. Item :" ' Effect of Failure in Non-Safety Related Systems CUring Design Basts Events"
- d.
Item 57 "Loss of Plant Air System (Effect on Plant Control and Monitoring)'
- e.
item 77 "Safety Related Control and instrument Power"
- 4.
0OT Recommenced List of SEP Subjects, Spring 1977 C DOT 102, Item iuuz, "Loss of Plant Air System (Effect on Plant Control and Monitoring)
-2
- a. USI A-47, Safety Implications of Control System, NUREG-0705, 0606 The issue defined in reference 1 (NUREG-0153, Item 22) is as follows:
In evaluating plant safety, the effects of control system mal functions should be reviewed as initiating events for anticipated transients and also as failures that could occur concurrently or subsequent to postulated anticipated events (initiated by a different malfunction) or postulated accidents."
The issue defined in USI A-47 is in part as follows:
"This issue concerns the potential for transients or accidents being :aade more severe as a result of the failure or malfunction or control systems.
These failures or malfunctions may occur independently, or as a result of the accident or transient under consideration."
- b. USI A-17, System Interaction in Nuclear Power Plants, NTREG-0649, 0606 The purpose of this task is to develop a method for conducting a disciplined and systematic review of nuclear power plant systems, for both process function couplings of systems and space couplings, to identify the potential sources and types of systems inter actions that are determined to be potentially adverse.
A report has been developed, "Final Report - Phase 1 System Interaction Methodology Application Program," MUTREG/CR-1321, SAND 80-0384 whose obectives are:
- 1. To develop a methodology for conducting a disciplined and systematic review of nuclear power plant systems which facilitates identification and evaluation of systems inter actions that affect the likelihood of core damage.
- 2. To use the methodology to assess the Standard Review Plan to determine the completeness of the plan in identifying and evaluating a limited range of systems interactions.
The work done under USI A-17 may be useful in the development of US: A-47.
The definition of USI A-47 is identical to that of Topic VII-4; therefore, this SEP topic has been deleted.
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TMI Task Action Plan - NURE-0660, 0737 - Task II.F., Instrumentation and Controls There are three subtasks under Task II.F. as follows:
- a. II.F.1 - Additional Accident -Monitoring Instrumentation
- b. II.F.2 -
Identification of and Recovery from Conditions Leading to Inadequate Core Cooling
- c. II.F.3 -
Instruments for Monitoring Accident Conditions Specific postions onthe required instrumentation for II.F.1 and II.F.2 are in NUREG 0737 and Regulatory Guide (R.G.) 1.97, Revision 2 (December 1980).
Instrumentation needed for II.F.3 is also in R.G. 1.97, Revision 2.
The emphasis of MI Task II.F. is the Monitoring of Radiation and Process Variables ; guidance for this relies primarily on R.G. 1.97.
This is identical to the review proposed in Topic VII-3; therefore, this SEP topic has been deleted.
DINITION T C:
IX-2 Overhead Handling Systems -
Cranes
- 1. Definition:
Overhead handling systems (cranes) are used to lift heavy ojects in the vicinity of ?WR and BWR spent fuel storace facilities and inside the reactor building.
f a heavy objec- (e. g., a shielaed cask)
- ere to drop on the soent fuel or on the reactor core during refueling, tnere coulo oe a potential for overexposure of Plant personnel and for release of radioactivity to the environment.
- eview the overnead handling system, including sling and other lifting cevices, anc the potential for the drop of a heavy object on spent fuel including structural effects.
- 2.
Safety Objective:
To assess the safety margins, and imnrove mar:ins where necessary, of the overhead handline systems to assure that the potential for dropping a heavy object on spent fuel is witnin acceptacle limits ano that tne potential raciation case to an individual does not exceed the guicelines of 10 CFR Part 10U.
- 3.
Status:
RegulatCry Guide 1.1U4, "Overnead Crane Handling Systems for Nuclear Power Plants" was issued for comment iii Feoruary 1:76 and reerences various industry standards.
New applications (C-and 3L) are reviewed in accordance with the ACSB Branch 7echnical Position 9-1 wnich is identical to ;egulatory Guice 1.104.
The review of overhead hand'ing systiems of coerating reactor facilities is perfornmed on a generic oasis and has also been identified as a DOR Tecnnical Activity Category A.
- 4.
References:
- r.
A7-S nce *ec7nical osiion 9-l, "Overnesc Handling Systems
.:r Nuclear cwer ?iants"
- 3. Wink icok - Generic :ssue "uel Cas r
na is", April,
- 4.
OR Tecnical Activities, Categcry A :tm so, "Control of
'-eavy Loacs Over Soent :uel", April 1977
UST A-36, Control of Heavy Loads Near Soent Fuel, NUREG-0649 The review criteria required by USI A-36 (SRP 9.1.4 and NUREG-0554) are identical to the review criteria scecified in the References of SEP Topic IX-2 (BTP 9-1 and RG 1.104);
therefore, this SEP Topic has been deleted.
DEFINITION TOPIC:
X Auxiliary Feedwater System
- 1.
Def.1n ition:
Review the auxiliary feedwater system, associated instrumentation, and connection between redundant systems.
7ne review includes the aspects of pump crive and power supply diversity (e.g., electrical and steam-driven sources), and the water supply sources for the auxiliary feedwater system.
- 2. Safety Objective:
To assure that the auxiliary feedwater system can provide in adequate supply of cooling water to the steam generators for decay heat removal in the event of a loss of all main feecwater.
0ODer P'4R plants may not meet the require-ment for pump. drive and power supply diversity.
- 3.
Status:
Reviews for new license applications are performed in accordance with the SR?.
This topic is not under active review for operating plants.
.Refe-ences:
- 1. 5S?,
10.4.9
- 2. A C ST9 10-1, "Design Guioelines for Auxiliary Feedwater System ump Drive and Power Supply Diversity for ? 'R ?lants'
- 5.
Basis for Deletion (i.e., related TMI Task, USI or other SEP Tooic):
TMI Task Action Plan - NUREG-0660--
Task II.E.1.1, Auxiliary Feedwater System Evaluation The TMI-2 accident and subsequent investioations and studies highlighted the importance of the auxiliary feedwater (AFW) system in the mitigation of severe transients and accidents.
Since then, the AFW systems have come under close scrutiny by the NRC and many improvements have been recommended to enhance the reliability of AFW systems for all plants.
The.scooe of the review outlined in the SEP Topic X definition is identical to the scope of NUREG-0737, "Clarification of TMI Action Plan Requirements," Item II.E.l.1(2) which requires that each PWR plant licensee:
"Perform a deterministic review of the AFW system using the acceptance criteria of Standard Review Plan Section 10.4.9 and associated Branch Technical Position ASB 10-1 as pr-incipal guidance."
Tne review crIterIa for the evaluations required by Item TI.E.1.1(b) are identical to SEP Topic X; therefore, this SEP Topic has been deleted.
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- 2 1.R. 1. 1.8 and 1.
- 2.
ANSI N8.1 -
1971Property "ANSI code" (as page type) with input value "ANSI N8.1 -</br></br>1971" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.
- 3. ANS i%18.7 -
1972 e evised
- 4.
Standard Techni:al Specifications,Section VI 10 CFR :0, Appendix~ £
- i.
R. G. 1.,101 ;ev. 1 1977
- 7. SRP 13.3
- 8. Nu;EG 75/111, Guide and Checklist for Develo;ment and Evaluation of State and ocal Goverment Radiolocical me gency Response Plans :n Support of ixed Nuclear Facilities
- 9.
EPA Manual of Protective Action Guides anc ?Protective Action for Nuclear >ncidents',
September 1975
- 10.
Menorandum of Incerstanding, NR; and 02S on State and Local Prepareeness, Marcn 10, 1977 Basis for Deletion (i.e., related TMI Task, USI or other SEP Topic):
- a. TMI Task Action Plan, NUREG-737, Task I.C.6, Guidance on Procedures for Verifying Correct Performance of Operating Activities Under TMI TaskI.C.6 a review-of licensee procedures will be conducted to assure that an effective system of verifying the correct performance of operating activities exist. The purpose of this review is to provide a means of reducing human errors and improving the quality of normal operation. References cited for this review are ANSI Standard N18.7-1972 (ANS 3.2) "Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants," and Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)."
These are the same ra:erences cited for Tooic XIII-1.
- b. TMI Task Action Plan, NUREG-0660 and 0737, Task II.A.1, "Imorovine Licensee 7mercency Preparedness
- Short Term" and Task III.A.2, "mrovins Licensee Emergencv ?reparedness Long Term" Under Task III.A.1 a review of 10 CFR Part 50, Appendix E backfit requirements is being conducted in accordance with N1-UREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power ?ants."
The scope of NUREG-0654 covers SRP 13.3 and NUREG 75/111.
R.G.
1.101 has been deleted and has been superceded by an amended Appendix E co 10 CR Part 50 (45 FR 55410, August 19, 1980).
Under Task III.A.2 a review of licensees emergency prepardness plans ith respect to amended Appendix will be conducted in accordance with NLREG-0654.
- g 9
The evaluations required by TMI Tasks I.C.6, III.A.1 and III.A.2 are identical to SEP Topic XIII-1; therefore, this SEP topic has been deleted.
.0 0
OETINITION COI:
XV-21 Spent Fuel 'ask Drop Accidents 1.I Diiti on:
Review the potential for scent fuel cask drops, the damage which could result from cask droos, and the radiolocical consequences of a cask drop from fuel damaged within the cask under conditions ex:eeding the design basis impact on the cask.
- 2. Safety Oective:
To assure that the damage to fuel within the casks and radio logical consecuences resulting from a cask droo are acceptable or that acceptable measures have been taken to preclude cask croos.
- 3. Status:
Fuel cask droo analysis is a generic item-4hich has been comleted on some 01ants or is presently under review for all other ocerating reactors.
. eferences:
- 1. S.P Section 15.7.5
- 2. R. G. 1.25
- 3. Pink Book
USI A-36, Control of Heavy Loads Near Soent Fuel, NUREG-0649 The review criteria required by USI A-36 (SRP 15.7.5) are identical to the review criteria specified in the References of SEP Topic IX-2; therefore, this SEP Topic has been deleted.
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DOEINITION TCPC: XV-23 Multiple Tube Failures in Steam Gererators
- 1. Definition:
4ssess the effects of multiple steam generator tube failures (rancing from leaks to double ended ructures) as a result of pressure differentials that may occur following a LOCA, steam line break or A7S events.
- 2.
Safety Objective:
Assure that the reflood of the core following a LOCA is possible and that the radiological consequences following these accidents are witnin tne 10 CP Part 100 guidelines.
Status:
The consequences of multiole 'ute failures have not been analyzed for any ;lant at the licensing stage.
'ork has neen done for some operating plants, but ultimate goals have yet to be set.
Re*erenceS:
- 1. Prairie :sliand Docket
- 2.
Turkey -oint Docket
- 3.
Surry -1 and =2 Docket
- 4.
AT 5eoort
- a. USI A-3, 4, 5 "Westinghouse, Combustion Engineering, and Babcock and Wilcox Steam Generator Tube integrity, NUEG 0649.
Two of the tasks of USI A-3, 4, 5 are as follows:
- 1. Analyses of LOCA with Concurrent Steam Generator Tube
- Failures,
- 2. Analyses of Main Steam Line Break.
The analyses required by these two tasks in USI A-3, 4,.5 covers two of the three events specified in the definition.
USI A-9. "Anticipated Transients Without Scram (ATWS)"
NTUIEG-0606
?ressure differentials resulting from ATWS events have been determined to be no greater than those resulting from main steam line break events (JUREG-0
-60, Vol.
2, Appendix V).
The analysis for ATWS event is, therefore, covered under US! A-3, 4,
5.
-2 The evaluation required for USI A-3, 4,5 is identical to SEP Topic XV-23; therefore this SEP Topic has been deleted.
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ENCLOSURE 3 GENERIC SEP TOPICS SEP NRR TOP IC GENERIC NO.
SEP TITLE NO.
GENERIC TASK TITLE lil-7.A (Note 1) Inservice Inspection -
Containments B-49 Inservice Ins pection Criteria for Containments B-38 Tendon Survei llance - Bechtel Containments illI-8.A Loose Parts & Core Barrel Vibration Monitoring B-60 Loose Parts Monitoring System B-73 Monitoring for Excessive Vibration Inside Reactor Vessel C-12 Primary System Vibration Assessment V-1 Compliance with Codes & Standards A-O 10 CFR 50.55a(g) -
ISI A-14 10 CFR 50.55a(g) -
Inservice Testing V-7 (Note 2)
Reactor Coolant Pump Overspeed B-68 Pump Overspeed during a LOCA VI-6 Containment Leak Testing A-04 Appendix J -
Containment Leak Testing A-23 Containment Leak Testing VI-7.1)
Long Term Cooling Passive Failures B-11 Flood of Equipment Important to Safety VI 1-113 Trip Uncertainty & Setpoint NUREG-Ol 38 Ins truinentation Setpoint Analysis Review Issue 13 Drift Vill-1.A Degraded Grid Voltage A-35 Adequacy of Offsite Power Systems B-23 Degraded Grid Voltage IX-6 Fire Protection B-02 Fire Protection Xl Appendix I A-02 Appendix I -
ALARA XI-2 Radiological Monitoring Systems A-02 Appendix I -
ALARA B-67 Effluent & Process Monitoring Instrumentation
TOPIC GENERIC NO.
SEP TITLE NO.
GENERIC TASK TITLE XIII-2 Safeguards/Industrial A-03 Security Reviews XVII Operational QA Program Annual IE Inspections III-4.B (Note 3)
Turbine Missiles B-46 Analysis of Turbine Disc Cracks IV-3 (Note 4)
BWR Jet Pumps Operating BWR Jet Pump Flow Indication Indications B-28 Elimination VI-7,A,2 (Note 5)
Upper Plenum Injection 0-05 Plant UPI Mbdel Problem VI-TA.4 (Note 6)
Core Spray Nozzle Effectiveness D-12 Non-Jet Pump BWR Core Spray Performance VII-6 Frequency Decay A-35 Adequacy of Offsite Power Systems NOTES:
- 1. Applies to Palisades and Ginna only, which have prestressed concrete containments.
Topic is deleted for all other SEP plants.
- 2. Applies to Ginna, Haddam Neck, Palisades, San Onofre, and Yankee Rowe, which are PWRs.
Topic is deleted for all BWR SEP plants.
- 3. Applies to Palisades, Ginna, San Onofre, Yankee Rowe, and Haddam Neck. Topic is deleted for all other SEP plants.
- 4. Applies to Millstone 1 and Dresden 2.
Topic is deleted for all other SEP plants,.
- 5. Applies to Ginna only. Topic is deleted for all other SEP plants.
- 6. Applies to Big Rock Point, Dresden 1, Dresden 2, Millstone 1 and Oyster Creek. Topic is deleted for all other SEP plants.