ML19289E263
| ML19289E263 | |
| Person / Time | |
|---|---|
| Issue date: | 03/21/1979 |
| From: | Carbon M Advisory Committee on Reactor Safeguards |
| To: | Hendrie J NRC COMMISSION (OCM) |
| References | |
| ACRS-0824, ACRS-824, NUDOCS 7904110014 | |
| Download: ML19289E263 (36) | |
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UNITED STATES p,
NUCLEAR REGULATORY COMMISSIOB'
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ADVISORY COMMITTEE ON REACTOR SAFEGU/.RDS y
WASHINGTON. D. C. 20555 March 21, 1979 Honorable Joseph M. He irie Chairman U. S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
STA'I1JS OF GENERIC ITEMS REIATING TO LIGHT-WA'IER REAC'IORS:
REPORT NO. 7
Dear Dr. Hendrie:
'Ihe Advisory Committee on Reactor Safeguards has previously reported on the " Status of Generic Items Relating to Light-Water Reactors" in its letters of December 18, 1972, February 13, 1974, March 12, 1975, April 16, 1976, February 24, 1977 and November 15, 1977. Since the Committee limits its definition of generic items to those cited specifically in its letters pertaining to projects and related matters, the attached listing is not all-inclusive; the Nuclear Regulatory Commission Staff has additional generic items.
In an effort to simplify referencing, the Committee has revised the number-ing system for its generic items.
(Attachment 4 cross-references this num-bering system with that in Report No. 6.)
Items 1 through 48 in Attachment 1 are a reiteration of the generic items considered resolved at the time the Committee issued Report No. 6, on November 15, 1977.
Items 49 through 52 are those items resolved since November 1977.
Following each resolved item is a brief statement of the specific action that resulted in resolution.
Items S3 through 77 listed in Attachment 2 are those items previously listed for which resolution on a generic be cis is still pending. The ACRS and the NRC Staff will continue to consider the safety significance of these items on a case-by-case basis until generic resolution is reached. Formal actions, such as issuance of Regulations or Regulatory Guides, are anticipated for many of these items.
Owing to questions raised concerning the scope and intent of various generic issues, the Committee has included, in Attachment 3, a brief description for all unresolved items cited in this report.
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Honorable Joseph M. Hendrie March 21, 1979 With regard to the status of generic issues, as they apply to each plant, the NRC Staff addresses the status of the pertinent issues 1 t the appli-cable Safety Evaluation Report. The ACRS identifies those % t it believes relevant in its reports on individual projects.
" Resolved" as used in the Generic Items reports refers to the following:
In some cases an item has been resolved in an administrative sense, recog-nizing that technical evaluation and satisfactory implementation are yet to be completed. Anticipated Transients Without Scram represents an ex-ample of this category.
In other inctances, the resolution has been ac-complished in a narrow or specific sense, recognizing that further steps are desirable, as practical, or that different aspects of the problem re-quire further investigation. Examples are the possibility of improved methods of locating leaks in the primary system, and of improved methods or augmented scope of in-service inspection of reactor pressure vessels.
The ACRS expects to report to you from time to time on the status of generic items.
Sincerely znurs, 7
Max W. Carbon Chairman Attachments:
1.
Resolved Generic Items 2.
Unresolved Generic Items 3.
Descriptions of the Unresolved Generic Items 4.
Cross-reference of Numbering System between the Present and Previous Report.
GENERIC ITEMS Resolved oeneric Items 1.
Net Positive Suction Head for ECCS Pumps: Covered by Regulatory Guide 1.1.
2.
Emergency Power: Covered by Regulatory Guides 1.6,1.9, and 1.32 and portions of ILEE-308 (1971).
3.
Hydrogen Control After a toss-of-Coolant Accident (LOCA): ACRS concurred in proposed Staff p)sition, covered by NRC Standard Review Plan for Nuclear Power Plants.
4.
Instrument Lines Penetrating Containment: Covered by Regulatory Guide 1.11 and Supplement.
5.
Strong Motion Seismic Instrumentation: Covered by Regulatory Guide 1.12.
6.
Fuel Storage P:01 Design Bases: Covered by Regulatory Guide 1.13.
7.
Protection of Primary System and Engineered Safety Features age. inst Pump Flywheel Missiles: Covered by Regulatory Guide 1.14.
8.
Protection Against Industrial Sabotage: Covered by Regulatory Guide 1.17.
9.
Vibration Monitoring of Reactor Internals and Primary System:
Covered by Regulatory Guide 1.20.
10.
In-service Insp3ction of Reactor Coolant Pressure Boundary: Covered by ASME Boiler and Pressure Vessel (BPV) Code,Section XI and Regulatory Guide 1.65.
- 11. Quality Assurance During Design, Construction and Operat. ion:
Covered by 10 CFR 50, Appendix B; ASME BPV Code,Section III; ANSI N-45.2-1971, Regulatory Guides 1.28, 1.33, 1.64, 1.70.6 and Proposed Standard ANS-3.2.
s 12.
Inspection of BWR Steam Lines Beyond Isolation Valves: Cove.ed by ASME BPV Code,Section XI.
13.
Independent Cneck of Primary System Stress Analysis: Covered by L ' BFV Code,Section III.
14.
Operational Stability of Jet Pumps: Test and operating ex wrience at Dresden 2 and 3 and other jet pump BWRs have satisfied the ACRS concerns.
- 15. Pressure Vessel Surveillance of Fluence and NDI' Shift: Covered by 10 CFR 50, Appendix A and Appendix H; and AS'IM Standard E-185.
- 16. Nil Ductility Properties of Pressure Vessel Materials: Covered by 10 CFR 50, Appendix A and Appendix G; ASME BPV Code,Section III;
" Report on the Integrity of Reactor Vessels for Light-Water Power Reactors," (WASH-1285) by the Advisory Committee on Reactor Safe-guards dated January 1974.
- 17. Operation of Reactor With Less W an All Loops In Service: Covered bv ACBS-Regulatory Staff position that manual resetting of several set points on the control room instruments under specific conditions and procedures is acceptable in taking one primary loop out of service.
Wis position is based on the expectation that this mode of operation will be infrequent. Cited in Standard Review Plan Appendix 7-A, Branch Technical Position EICSB 12.
- 18. Criteria for Preoperational Testing: Covered by Regulatory Guide 1.68.
- 19. Diesel Fuel Capacity: Covered by ACRS-Regulatory Staff psition requiring 7 days fuel (Standard Review Plan 9.5.4).
20.
Capability of Biological Shield Withstanding Double-Ended Pipe Break at Safe Ends: Covered by ACRS-Regulatory Staff position cited in several letters that such a failure should have no unacceptable consequences.
21.
Operating One Plat hhile Other(s) is/are Onder Construction:
Specific requirements have been established by ACRS-Regulatory Staff.
Covered in Regulatory Guide 1.17, 1.70 Section 13.6.2; 1.101; ANSI N 18.17 and Standard Review Plan 13.3 Appendix A and 13.6.
22.
Seismic Design of Steam Lines: Covered by Regulatory Guide 1.29.
- 23. Quality Group Classifications for Pressure Retaining Components:
Covered by Regulatory Guide 1.26.
- 24. Ultimate Heat Sink: Covered by Regulatory Guide 1.27.
25.
Instrumentation to Detect Stresses in Containment Walls: Covered by Regulatory Guide 1.18.
- 26. Use of Furnace Sensitized Stainless Steel: Covered by Regulatory Guide 1.44.
27.
Primary System Detection and Incation of Leaks: Covered by Regulatory Guide 1.45.
28.
Protection Against Pipe Whip: Covered by Regulatory Guide 1.46.
- 29. Anticipated Transients Without Scram: Covered by Regulatory Position Document, " Technical Report on Anticipated Transients Without Scram for Water-Cooled Power Reactors," WASH-1270, September 1973.
30.
DCCS Capability of Current and older Plants: Covered by Rulemaking as a general policy decision, although acceptable detailed implementation remains to be developed. Docket RM-50-1, " Acceptance Criteria for Emergency Core Cooling Systems for Light-Water-Cooled-Nuclear Power Reactors," December 28, 1973.
- 31. Positive Moderator Coefficient:
EWRs presently have or expect to have zero or negative coefficients. Where some Technical Specifications allow a slightly positive coefficient, the accident and stability analyses take this into account. Burnable poison provisions have been designed into PWRs to reduce otherwise excessive positive coefficients to allowable values.
32.
Fixed Incore Detectors on High Power PWRs:
Fixed incore detectors are not required for PWRs since reviews of potential power distribution anomalies have not revealed a clear need for continuous incore monitoring.
33.
Performance of Critical Components (pumps, cables, etc.) in post-LOCA Eavironment: Qualification requirement's of critical components are now covered by Regulatory Guides 1.40, 1.63, 1.73 and 1.89 and IEEE Standards 382-1972, 383-1974, 317-1972, 323-1974.
- 34. Vacuum Relief Valves Controlling Bypass Paths on BWR Pressure Suppression Containments:
On designs prior to GE Mark III con-tainment, resolution lies in surveillance and testing of vacuum relief valves. For Mark III containments, an additional require-ment is that the design be capable of accommodating a bypass equivalent to one square foot for a given flow condition.
35.
Emergency Power for Two or More Reactors at the Same Site: Resolved by issue of Regulatory Guide 1.81.
36.
Effluents from Light-Water-Cooled-Nuclear Power Reactors: Resolved by issue of Appendix I to 10 CFR 50.
- 37. Control Rod Ejection Accident: Resolved for PWRs by Regulatory Guide 1.77 38.
Main Steam Isolation Valve Leakage of BWR's: Covered by Regulatory Guide 1.96.
39.
Ebel Densification: Covered by 10 CFR 50 Appendix K plus case-by-case review of vender fuel models.
40.
Rod Sequence Control Systems: Covered by NRC Staff Review and Approval of NEDO-10527 and Presentation to ACRS.
- 41. Seismic Category I Requirements for Auxiliary Systems: Covered by Regulatory Guides 1.26 and 1.29, 42.
Instruments to Detect (limited) Fuel Failures - NRC document, " Fuel Failure Detection in Operating Reactors," B.L. Siegel and H. H. Hagen, June,1976 resolves issue for limited fuel failures, but not for severe failures (See Item 56).
43.
" Instrumentation to Follow the Course of an Accident" Regulatory Guide 1.97 Revision 1 resolves ACRS concerns.
44.
Pressure in Containment Following LOCA - NRC document, " Containment Subcompartment Analysis" September 1976.
45.
Fire Protection. Resolved by Branch Technical Position 9.5.1, and Regulatory Guide 1.120.
- 46. Control Rod Drop Accident (BWRs):
Resolved through NRC review and documentation establishing such an event as not having severe con-sequ.ences (memorandum for M. Bender, Chairman ACRS, from Denwood F. Ross, Jr., Assistant Director for Reactor Safety, DSS, dated February 11, 1977.)
47.
Rupture of High Pressure Lines Outside Containment: Resolved by positions in Stacdard Review Plan 3.6.1 and 3.6.2.
48 Isolation of Iow Pres;ure from High Pressure Systems: Resolved by positions in Standard Review Plan 5.4.7.
49.
M?nitoring For Loose Parts Inside 'Ihe Reactor Pressure Vessel:
Resolved by Staff position to be documented in Regulatory Guide 1.133.
- 50. Qualification of New Fuel Geometries: Resolved by position in Standard Review Plan 4.2, Revision 1.
- 51. Maintenance and Inspection of Plants: Resolved by the amount of Staff attention and industry involvement documented in Memorandum for Larry P. Crocker, Technical Assistant to the Director, DPM, from William E. Kreger, Acting Assistant Director for Site Analysis, IEE subject: Resolution of ACRS Generic Item II C-6 dated February 28, 1979.
52.
Safety Related Interfaces Between Reactor Island and Balance of Plant: Resolved by position in Standard Review Plan 1.8.
Resolution Pending
- 53. Wrbine Missiles: Turbine failures for past 16 years have been evaluated and a statistical probability analysis has been completed.
An ACRS letter (April 18, 1973) discusses the problems. (1) 54.
Effective Operation of Containment Sprays in a LOCA: Extensive documentation in topical reports. Review and evaluation required.
- 55. Possible Failure of Pressure Vessel Post-LOCA By Thermal Shock:
Regulatory Guide 1.2 covers current information. Ultimate position as to significance of thermal shock requires input of fracture mechanics data from the Heavy Section Steel Technology Program.
56.
Instruments to Detect (Severe) Fuel Failures - NRC document, " Fuel Failure Detection in Operating Reactors," B. L. Siegel and H. H.
Hagen.
Item 42 covers limited failures. bbre work is required for the severe failure case to establish instrumentation criteria. (2) 57.
Monitoring for Excessive Vibration Inside the Reactor Pressure Vessel: Neutron Noise Analysis has been successful in detecting vibration of some components; however, additional work may be required concerning systems for detecting vibration in other components within the Reactor Pressure Vessel.
58.
k n-Random Multiple Failures: W is heading covers a multiplicity of diverse components for which requirements should be established.
Due to their diversity, the ACRS feels that specific items should be separated into subsets under the general heading of non-random multiple failures; 58A - Reactor Scram Systems 588 - Alternating Current Sources onsite and offsite 58C - Direct Current Sources h e above items are easily identified, other specific items may be added to this listing in the future.
(1)
Regulatory Guide 1.115, " Protection Against Iow Trajectory Turbine Missles," will be modified to cover both low and high trajectory missiles.
(2)
Identified in the Committee's Report of April 16, 1976 as " Instruments to Detect Fuel Failures."
59.
Behavior of Reactor Fuel Under Abnormal Conditions: his includes:
flow blockage; partial melting of fuel assemblies as it affects reactor safety; and transient effects on fuel integrity. We PBF program will address some of these items.
60.
BVR Recirculation Pump and PdR Primary Coolant Pump Overspeed During LOCA: Decision required by ACRS-NRC Staff. (3)
- 61. We Advisability of Seismic Scram: Further studies required to establish need.
62.
Emergency Core Cooling System Capability for Future Plants:
Partially resolved by amendments to 10 CFR 50 [50.34(a)(4),
50.34 (b) (4), 50.46, and Appendix K).
LOCA evaluation model complete. M RS feels new cooling approaches should be explored.
63.
Ice condenser Containments: Additional analyses are required to establish response during a IOCA, and to establish design margins.
- 54. Steam Generator Tube Leakage: Partially resolved by issuance of Regulatory Guide 1.83 which addresses the concern from a pre-ventative point of view.
- 65. ACRS/NRC Periodic 10-Year Review of all Power Reactors: A more effective, continuous alternative approach to periodic reviews is being proposed. Pending ACRS review, this item is still considered unresolved.
- 66. Computer Reactor Protection System:
Systems should be qualified for reliability, particularly through in situ tests and under various environmental conditions, prior to use in reactor system. (4) 67.
Behavior of BWR Mark II Containments: Various aspects, including vent clearing, vent / coolant interaction, pool swell, pool strati-fication, pressure loads and flow bypass should be resolved. mis is an extension of Item 44.
(3)
Item 60 combines two previous items which dealt with PWR and SVR pump overspeed separately.
(4)
Identified in the ACRS Report of April 16, 1976 as " Hybrid Reactor Protection System."
- 68. Stress Corrosion Cracking in BWR Piping:
Several failures have occurred in operating BWRs. % e ACRS letter of February 8, 1975, discusses possible actions that should lead to generic resolution and extensive programs are underway by industry, ERDA, and NRC.
- 69. Incking Out of ECCS Power Operated Valves: h e Committee suggests that further attention be given to procedures involving locking out electrical sources to specific motor-operated valves required in the engineered safety functions of ECCS.
- 70. Design Features to Control Sabotage: Attention should be given to aspects of design that could improve plant security.
- 71. Decontamination of Reactors: As experience is gained in reactor decontamination it should be factored into future plants to optimize control of radioactivity levels.
- 72. Decommissioning of Reactors:
Specific plans should be developed, including definitive codes and standards to cover the ultimate decommissioning of plants.
- 73. Vessel Support Structures:
Questions that have arisen concerning the loads on pressure vessel support structures due to certain postulated loss-of-coolant accidents should be resolved.
- 74. Water Hammer: Several cases of water slugging or water hammer have occurred in both PdRs and BWRs. Corrective measures should be taken to minimize such events.
- 75. Behavior of BWR Mark I Containments: Various aspects relevant to the EWR Mark I Containment should be resolved.
Included are such items as relief valve restraint, control of local dynamic loads in the torus, vent clearing and establishment of torus water temperature limits during a LOCA. E is is an extension of Item 44.
- 76. Assurance of Continuous tong-term Capability of Hermetic Seals on Instrumentation and Electrical Equipnent: he integrity of seals during post-accide-' conditions may be critical in controlling such an accident. %e ammittee believes appropriate test and maintenance procedures should be developed to assure long-term reliability.
- 77. Soil-Structure Interactions:
Several matters related to soil-structure interaction and the appropriate seismic respnse spectrum for use at foundation levels of nuclear plants are under review and reevaluation.
53 Turbine Missiles Turbine failures for the past 16 years have been evaluated and a statistical probability analysis has been completed. An ACRS letter
' April 19, 1973) discuses the problem.
% rea issues require answers to resolve the turbine missile problem:
(1) '1he first relates to the appropriate failure probability value;
-4 based on historical failures the probability is about 10 / turbine-year.
Industry predicts a much lower failure probability based on improvements in materials and design. To date the ACRS has accepted the more conser-vative value; (2) The second issue is strongly dependent on turbine ori-entation with respect to critical safety structures.
Strike probabilities from high angle missiles are acceptably low for single units and may be acceptable for multi-unit plants, depending on plant layout; however, lower angle missiles with non-optimum (tangential) turbine orientation have unacceptably high strike probabilities; (3) The third issue is one of penetration and damage of structures housed in the containment. We limited experimental data pertaining to penetration of large irregularly shaped missiles are not sufficient to determine structural response to impingement of turbine disc segments. Most missile penetration formulas are not relevant to this case. We EPRI turbine missile impact experiments might resolve this issue, particularly for older plants with non-optimum turbine orientations.
54 - Effective Operation of Containment Sprays in a LOCA Review and evaluation are required of the variety of experiments which have been conducted on the effectiveness of various containment sprays for the removal and re*.ention of airborne radioactive materials anticipated to be present within containment following a LOCA.
Such review should consider adequacy of definition of the physical and chemical forms of the anticipated airborne radionuclides, and quality of evaluative tests of the removal efficiencies of various sprays under the conditions of temperature, pressure, and radiation doses expected to exist under LOCA conditions. A desirable extension might be analyses of the use of sprays containing chemicals (such as NaOH) which have the potential for damaging equiprnent within containment. Studies using other spray additives, such as hydrazine, have been conducted.
If compounds, such as this, have distinct advantages, insofar as minimizing equipnent damage in the event of inadver-tent actuation, action should be taken to encourage their use.
55 - Possible Failure of Pressure Vessel Post-LOCA By Thermal Shock Earlier nuclear reactor pressure vessels subjected to fluences of 19 l-4 x 10 nyt, which are anticipated in the last 20 years of a 40-year life, may suffer severe radiation damage denoted by a pronounced shift in impact transition temperature at the inner surface. tere will be a damage gradient which decreases sharply, so that the properties halfway through the wall are essentially those of the as-fabricated material.
If a LOCA occurs near end-of-life, the injection of cold water on the region of degraded properties may initiate and propagate a crack because of high local stresses near the surface. Analytic procedures indicate the stresses drop rapidly with distance through the wall so the flaw should not propagate beyond some limiting point. We lack of experimental evidence and the relative width of the error band in the analytic results are such that some experiments are required to validate the analytic model. %ese are under way in the HSST program.
t 56 - Instruments To Detect (Severe) Fuel Failures In the event of substantial fuel failure, including the possibility of fuel melt, large amounts of fission products could be rapidly released to the reactor coolant and possibly to the environment.
Instrumentation capable of early warning and timely response may avert an incident be-coming an accident.
Instrumentation related to such diagnostic purposes for limited fuel failure is being used on most power reactors (see Item 42). Further work is required to establish criteria for similar instrumentation for severe fuel failures.
57 - Monitoring For Excessive Vibration Inside The Reactor Pressure Vessel Neutron noise analysis can detect vibration of specific components such as the core barrel. 'Ihe detection of vibration in other reactor pressure vessel components is less well established.
t
58 - Non-Random Multiple Failures (Formerly " Common Mode Failure")
The term " common mode failures" has, in many instances, come to mean multiple failures of identical components exposed to identical or nearly identical conditions or environments, and the use of diversity in components has been proposed or required to avoid such failures. We concern of the ACRS is better expressed by the term "non-random multiple failures," which is intended to include not only the type of "comron mode failure" discussed above but other types of multiple failures for which the consequences and probabilities cannot be predicted by application of the single-failure criterion. Examples include the use of the same sensors or components for both control and protection systems (a resolved matter); sequential multiple failures due to a " domino effect," and simultaneous multiple failures due to a single fault. Since designs usually do not knowingly incorporate features susceptible to such failures, techniques and criteria need to be developed to detect and avoid them in all systems important to safety. te following is a partial listing of systems whose common mode failure has been cited by the ACRS as a matter of safety concern:
58A - Scram Systems 58B - Alternating Current Sources 58C - Direct Current Sources Other items may be added to this listing in the future.
59 - Behavior Of Reactor Fuel Under Abnormal Conditions The behavior of reactor fuel under abnormal conditions is still considered unresolved due to the limited experimental data available.
Partial melting of fuel assemblies due to flow blockage might lead to autocatalytic effects leading to more extensive fuel failure, pressure pulses, etc. Similar behavior might occur in the case of reactivity transients. 'Ihe ACRS encourages analytic modeling but believes appropriate experimental data are necessary. It is anticipated that tests in the Power Burst Facility (PBF) should supply much of the required data.
60 - BWR and PWR Pump Overspeed During A IDCA It is possible for a BWR recirculation pump or a PWR primary coolant pump to overspeed if a large break occurs at the appropriate position in specific piping. Conservative estimates indicate substantial overspeed and possible failure of components, with the generation of missiles. The problem is being approached analytically and experimentally with scaled pumps. The reliability of such protective measures as the use of decouplers between pump and motor is under study for BWRs.
In PWRs the reliability of such protective measures as electrical braking of the pump motor is under study.
e
61 - The Advisability of Seismic Scram he ACRS has recommended that studies be made of techniques for seismic scram and of the potential safety advantages and potential disadvantages of prompt reactor scram in the event of strong seismic motion, say more than one-half the safe shutdown earthquake. Various suitable techniques have been identified and exist, but thus far only limited studies have been reported on the pros and cons of seismic scram. We principal po-tential advantage identified arises from the greatly improved coolability of a core in the unlikely event of a seismically induced LOCA, should scram precede the LOCA by several seconds. A principal reason given in opposition to seismic scram relates to a stated interest in keeping power stations on the line to provide pwer offsite should a severe earthquake occur.
62 - ECCS Capability For Future Plants
'Ihe ACRS has placed considerable emphasis on ECCS safety R&D so that the extent of the conservatism in the ECCS licensing requirements could be made more precise. With more experimental data a realistic and quantitative appraisal of ECC systems would lead to valid judgments on changes in licensing which could be put on a firm basis.
Parallel approaches that seek to improve the reliability of ECC systems, to improve the monitoring of low power peaking, and to improve those fuel assembly designs by achieving lower peaking factors, are encouraged.
- Further, changes in plant design which improve the reflooding of the reactor core should be sought and evaluated.
R&D efforts on analysis of core blowdown and reflood should be pursued and combined with the results of the standard problems and the associated experiments.
Improved analytical methods would provide a basis for optimized ECCS.
63 - Ice Condenser Containments The ice condenser containments have substantially smaller volume on the assumption that the ice will condense the steam during a LOCA, thus pre-venting system overpressurization. The rate of condensation is critical in the initial stages of the blowdown and is influenced by interaction of vapor with the ice.
If the current analyses prove that the condensation model is suitably conservative, the problem may be resolved.
64 - Steam Generator Tube Leakage Normally the steam generator is not a critical compnent during a LOCA-ECCS. Ibwever, a special case exists where the steam generator tubes have been degraded due to corrosion, wastage, etc.
If the shock loads imposed by the LOCA cause a critical number of tubes to fail, say by a double-ended (guillotine) break, the inflow from the secondary side can cause choking of flow during ECC, preventing adequate cooling of the core. We critical number of tubes is relatively small. A position, such as one specifying a statistically significant level of nondestructive examination (NDE), might resolve this issue. We purpose of NDE would be to confirm that damage is not excessive; such examinations should minimize the possibility of catastrophic failure of a significant number of tubes.
65 - Periodic (10-Year) Review Of All Power Reactors In its report of June 14, 1966, the ACRS recommended that periodic comprehensive reviews be conducted of operating licensed power reactors by the NPC Staff. Rese reviews would be preceded by a comprehensive report by the operator which evaluate the past experience and the safety of future operation of the plant.
The NRC Staff has maintained a continuing review of the safety of o erating plants.
In particular, as generic matters of potential rafety significance arise, the appropriate operating reactors are asked to assess the relevance of the matter to each particular reactor. 21s is a necessary but different aspect of the continuing surveillance and review of the safety of operating reactors than was envisaged by the ACRS in its recommendation of June 1% 3.
W e Committee continues to believe both approaches are desirable and awaits the developnent of a crogram of periodic comprehensive reviews.
66 - Computer Reactor Protection ',ystems*
The proposed systems would contain some types of components and subsystems not previously used for reactor protection.
It is necessary that the required system reliability, both during normal operation and under postulated abnormal conditions, be established through an appropriate combination of tests and analyses. Wille the issue originated with the B&W Hybrid concept it is cqually applicable to the proposed CE and W computer reactor protection systems.
- Identified in the ACRS Report of April 16, 1976 as " Hybrid Reactor Protection System."
67 - Behavior Of BWR Mark III Containments The BWR Mark III Containment differs in many respects i am the Mark I and II designs. Various aspects such as vent clearing, vent / coolant interaction, pool swell, pool stratification, pressure loads, and flow bypass must be evaluated and approved; ongoing experimental tests should develop much of the necessary data to confirm the conservatism in design.
68 - Stress Corrosion Cracking la BWR Piping Several failures have occurred in operating BWRs. An ACRS letter of February 8, 1975, discusses possible actions that should lead to generic resolution, and extensive programs are underway by Industry, ERDA and NRC.
The austenitic stainless steels are commonly used as piping material in many BWR lines. A combination of weld sensitization, residual stresses, superposed loads, and oxygen equal to or greater than 0.2 ppm in the BWR coolant can lead to cracking, initiating on the inner sur-face and propagating through the wall.
In most cases there will be a leak well before pipe failure so there is adequate warning; however, or-can postulate a LOCA caused by a guillotine break with minimal prior warning. Current efforts are to minimize stress corrosion by using other materials.
69 - Locking Out Of ECCS Power Operated Valves The physical locking out of electrical sources to specific motor-operated valves required in the engineered safety functions of ECCS has been required, based on the assumption that a spurious electrical signal at an inopportune time could activate the valves to the adverse position; e.g., closed rather than open, or open rather than closed.
While such an event has a finite probability another probability exists that the valves might be adversely positioned due to operator error.
The ACRS believes the matter should be studied using a systems approach, and considering such items as:
(1) the evaluation of the probability of a spurious sigaal; (2) time required to reactivate the valve operator; (3) status of signal lights when the circuit breaker is open; (4) the possibility of locking out in an improper position due to a faulty in-dicator; (5) other designs with improved rel-iability without lock-out; (6) the advantages and disadvantages of cc rective action by an alet.
operator in case of incorrect positioning vis-a-vis a system with power locked out.
70 - Design Features To Control Sabotage Considerable attention has been devoted to control of industrial sabotage of nuclear power plants, particularly with regard to control of unauthorized access, and potential modes of sabotage by individuals or groups external to the operating organization. 'lhe ACRS believes that deliberate attention should be given to aspects of design that could improve plant security. With the emphasis being placed on standardized plant designs, it becomes especially important to in-troduce design measures that could protect against industrial sabotage, or mitigate the consequences thereof.
71 - Decontamination Of Reactors The Committee believes that well developed plans, confirmed by appropriate experiments when necessary, should be available for the decontamination of primary reactor systems. At this time the information on full scale decontamination is '.imited.
Examples of potential problems include such items as handling of decontamination solutions, potential hideout of radioactive products, enhanced corrosion and crud formation following decon-tamination, and de possible incompatibility of the different alloys in the pressure boundary with the decontamination solution.s.
72 - Decorraissioning of Reactors Experience is limited with regard to decommissioning operations, and particularly with rules for dismantling and for mothballing.
Definitive plans and standards should be developed covering such items as adequacy of action, problems in restitution of site, mutual responsibility of State and Federal Govurnment, etc.
73 - Vessel Support Structures A passible consequence of tne instantaneous double-ended pipe break postulated to occur in certain large pipes of PWRs is the asymmetric loading of the reactor pressure vessel support structures. 'Ihe magr.1 -
tude and effects of such loads on the pressure vessel should be de-termined to establish if such loads adversely affect the predicted course of a IDCA.
If analysis indicates that the results are un-acceptable, appropriate corrective action should be taken. A poten-tial effect is pressure vessel movement due to blowdown jet forces at the location of the rupture, transient differential pressures in the annular region between the vessel and the shield, and transient differential pressures across the core barrel within the reactor vessel.
74 - Water Hammer Several instances of water slugging or water hammer have occurred in both BWRs and PWRs due to causes such as the trapping of water be-tween two valves. This slug of water is accelerated by steam or water once the valves are opened. The stored energy is sufficient to damage piping, bend or break pipe restraints, and damage support structures. Water hammer may occur due to flow instabilities in steam generators in conjunction with water flowing into the feedwater inlets, resulting la venparable damage.
Corrective measures should be taken to minimize such occurrences after completion of analytic and experimental studies directed to an under-standing of the causes.
75 - Behavior Of BWR Mark I Containments Recent tests on the BWR Mark I Containment design revealed phenomena not anticipated on the basis of earlier tests where pressure loads were imposed by insertion of air. Specific problems, somewhat comparable to those under review for the Mark III Containment, include relief valve discharge, pipe restraints in the torus, local dynamic loads on the torus, vent clearing, and influence of torus temperature on the LOCA.
Ongoing experiments are expected to develop the necessary data to confirm the adequacy of the existing design or to permit necessary modifications.
76 - Assurance of Continuous Long-Term Capability of Hermetic Seals on Instrumentation and Electrical Equipnent Certain classes of instrumentation incorporate hermetic seals. When safety related components within containment must function during post-LOCA accident conditions, their operability is sensitive to the ingress of steam or water if the hermetic seals are either initially defective or should become defective as a result of damage or agiry. 'Ihe damage processes may fall within Item 33, " Performance of Critical Components in Post-LOCA Environment"; however, a special case requiring evaluation has to do with personnel errors in the maintenance of such equipnent since such errors could lead to the loss of effective hermetic seals.
77 - Soil-Structure Interactions Ongoing sttriies by the NRC and the industry are reviewing and re-evaluating matters related to soil-structure interaction and to the appropriate seismic response spectrum to be used at the foundation level of a nuclear power plant. 'Ihese reviews may lead to a modification of current criteria used in the seismic design of foundation structures.
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Cross-Reference of Numbering System Between Present Report and the Previous Report
- Present Pravious Present Previous Present Previous Present Previous 1
I-l 23 I-23 45 ID-4 65 IIA-4 2
I-2 24 I-24 46 IE-1 66 IIB-1 3
I-3 25 I-25 47 IE-2 50 IIB-2 4
I-4 26 IA-1 48 IE-3 G7 IIB-3 5
I-5 27 IA-2 49 II-5A 68 IIB-4 6
I-6 28 IA-3 50 IIB-2 69 IIC-1 7
I-7 29 IA-4 51 IIC-6 70 IIC-2 8
I-8 30 IA-5 52 IID-1 71 IIC-3A 9
I-9 31 IB-1 53 II-1 72 IIC-3B 10 I-10 32 IB-2 54 II-2 73 IIC-4 11 I-ll 33 IB-3 55 II-3 74 IIC-5 12 I-12 34 IB-4 56 II-4 51 IIC-6 13 I-13 35 IB-5 49 II-5A 75 IIC-7 14 I-14 36 IB-6 57 II-5B 52 IID-1 15 I-15 37 IB-7 58 II-6 76 IID-2 16 I-16 38 IC-1 59 II-7 77 IIE-1 17 I-17 39 IC-2 60 II-8 & IIA-2 18 I-18 40 IC-3 61 II-9 19 I-19 41 IC-4 62 II-10 20 I-20 42 ID-1 63 IIA-1 21 I-21 43 ID-2 60 IIA-2 22 I-22 44 ID-3 64 IIA-3
- Status of Generic Items Relating to Light-Water Reactor Rpt. No. 6, 11/15/77.