ML25196A180

02-24-77 Status of Generic Items Relating to Light-Water Reactors-Report No.5

ML25196A180
Person / Time
Issue date:	02/24/1977
From:	Bender M Advisory Committee on Reactor Safeguards
To:	Rowden M NRC/Chairman
References
Download: ML25196A180 (1)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20566 Honorable Marcus A. Rowden Chairman February 24, 1977 U. s. Nuclear Regulatory Comnission Washington, DC 20555

Subject:

STA'IUS OF GENERIC ITF..MS REIATING 'IO LIGHT-wATER REAC'IORS:

REPORl' NO. 5

Dear Mr. lt1wden:

The Advisory Committee on Reactor Safeguards has previously rep:>rted on the "Status of Generic Items Relating to Light-Water Reactors" in its letters of December 18, 1972, February 13, 1974, March 12, 1975 and April 16, 1976. Since the Conmittee 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 Comnission Staff list has additional generic items.

Group I of the attachment is a reiteration of the generic items considered resolved at the time the Comnittee issued its first report. Group IA in-cludes those items resolved between December 1972 and February 1974; Group IB includes those items resolved between February 1974 and March 1975; Group IC includes those items resolved between March 1975 and April 1976; Group ID includes those items resolved since April 1976.

Following each resolved item-is a brief statement of the specific action that resulted in the re-solution. Group II lists those items included in the original report for which resolution on a generic basis is still pending. Groups IIA, IIB and IIC include generic items that "lere added in the second, third and fourth rep:>rts; Group IID includes those added in the present report. '!he ACBS and the NRC Staff will continue to consider the safety significance of Group II, IIA, IIB, IIC and IID 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 the Group II, IIA, IIB, IIC and IID items.

'!he April 16, 1976, ACBS letter cited nine items that the NRC Staff con-siders to be "resolved", based on s~ific positions taken by the Staff.

However, these *resolutions" did not meet the ACBS criteria for adequate documentation and Conmittee approval and were therefore considered to be pending until formally discussed and accepted by the ACBS.

Items in this category were:

2287

Honorable Marcus A. ~en February 24, 1977 (1)

"Effective Operation of Containment Sprays in a LOCA";

(2)

"Instruments to Detect Fuel Failures";

(3)

"Conroon Mode Failures";

(4)

"'!be Advisability of Seismic Scram";

( 5)

"Pressure in Containment Following UXA";

(6)

"Control Rod Drop Accidents (BWR)";

(7)

"Rupture of High Pressure Lines Outside Containment";

(8)

"Isolation of Low Pressure from High Pressure Systems";

(9)

"Behavior of BWR Mark III Containments."

Since the April 16, 1976 ACRS letter, the NRC Staff has provided docu-ments acceptable to the Cormtittee for items 2 and 5, above.

Two other items which have been resolved since the Report of April 16, 1976 are also listed in Group ID.

Owing to questions raised concerning the scope and intent of various generic issues, the Comnittee has incorporated into the attachments a brief description for all items unresolved now or in earlier reports.

With regard to the status of generic issues, as they apply to each plant, the ACRS understands that the NRC Staff will address the status of the issues in the applicable Safety Evaluation Report.

"Resolved" as used in the Generic Items reports refers to the following:

In some cases an item has been resolved in an administrative sense, re-cognizing that technical evaluation and satisfactory implementation are yet to be conpleted. Anticipated Transients Without Scram represents an example of this category.

In other instances, the resolution has been accomplished 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 to inservice inspection of reactor pressure vessels.

Sincerely y,

'P.11lende Olairman Attachments:

(1)

Group I; (2) Group IA; (3) Group IB; (4) Group IC; (5)

Group ID; (6) Group II; (7) Group IIA; (8) Group IIB; (9)

Group IIC; (10) Group IID 2288

GENERIC ITF.M.5 Group I - Resolved Generic 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 IEEE-308 (1971).

3. Hydrogen Control After a Loss-of-Coolant Accident (LOCA):

ACRS concurred in proposed Staff position, covered by NRC Standard Review Plan for Nuclear Power Plants.

4.

Instrl.Dllent 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 Pool Design Bases: Covered by Regulatory Guide 1.13.

7. Protection of Primary System and Engineered Safety Features Against PlDIIP Fl~eel 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. Inservice Inspection of Reactor Coolant Pressure Boundary:

Covered by ASME Boiler and Pressure Vessel (BPV) Code,Section XI and Regulatory Guide ~.65.

11. Quality Assurance During Design, Construction and Operation:

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.

12. Inspection of BWR Steam Lines Beyond Isolation Valves: Covered by ASME BPV Code,Section XI.

13. Independent Check of Primary System Stress Analysis: Covered by ASME BPV Code,Section III.

14. Operational Stability of Jet P1.1I1Ps:

Test and operating experience at Dresden 2 and 3 and other jet pump BWRs have satisfied the ACRS concerns.

2289

Group I Continued

15. Pressure Vessel Surveillance of Fluence and NDT Shift: Covered by 10 CFR 50, Appendix A and Appendix H; and ASTM, 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 Comnittee on Reactor Safe-guards dated January 1974.

17. Operation of Reactor With Less Than All Loops In Service: Covered by ACRS-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.

This position is based on the expectation that this node 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 position requiring 7 days fuel.

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 Plant While Other(s) is/are Under 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. ()lality Group Classifications for Pressure Retaining Conponents:

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.

2290

Group IA - Generic Items Resolved Since December 18, 1972

1. Use of Furnace Sensitized Stainless Steel: Covered by Regulatory Guide 1. 44.

2. Primary System Detection and Location of Leaks:

Covered by Regulatory Guide 1.45.

3. Protection Against Pipe Whip:

Covered by Regulatory Guide 1.46.

4. 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.

5.

ECCS 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 Einergency Core Cooling Systems for Light-Water-COOled-Nuclear Power Reactors," December 28, 1973.

2291

Group IB - Generic Items Resolved Since February 13, 1974

1.

Positive Moderator Coefficient:

PWRs presently have or expect to have zero or negative coefficients. Nlere 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.

2.

Fixed Incore Detectors on High Power PWRs:

Fixed incore detectors are not required for PWRs since reviews of potential power distribution ananalies have not revealed a clear need for continuous incore monitoring.

3.

Performance of Critical Components (purrps, cables, etc.) in post-ux:::A Environment: Qualification requirements of critical conponents 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.

4.

Vacum Relief Valves Controlling Bypass Paths on BWR Pressure Suppression Containments: 01 designs prior to GE Mark III con-tainment, resolution lies in surveillance and testing of vacum relief valves.

For Mark III containments, an additional require-ment is that the design be capable of accoom:>dating a bypass equivalent to one square foot for a given flow condition.

5.

&nergency Power for Two or More Reactors at the Same Site: Resolved by issue of Regulatory Guide 1. 81.

6.

Effluents from Light-water-Cooled-Nuclear Power Reactors: Resolved by issue of Appendix I to 10 CFR 50.

7.

Control Rod Ejection Accident: Resolved for PWRs by Regulatory Guide 1. 77.

2292

Group IC - Generic items Resolved Since March 12,*1975

1.

Main Steam Isolation Valve Leakage of BWR's:

Covered by Regulatory Guide 1.96.

2.

Fuel Densification: Covered by 10 CFR 50 Appendix K plus case-by-case review of vendor fuel JIDdels.

3.

Rod Sequence Control Systems: Covered by NRC Staff Review and Approval of NEI0-10527 and Presentation to ACRS.

4.

Seisnic Category I Requirements for AllXiliary Systems: Covered by Regulatory Guides 1.26 and 1.29.

2293

Group ID - Generic Items Resolved Since April 16, 1976

1. 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 II-4).

2.

Instrwnentation to Follow the Course of an Accident" Regulatory Guide 1.97 Revision 1 resolves ACRS concerns.

3. Pressure in Containment Following LOCA-NRC document, "Containment Sub-conpartment Analysis September 1976.

4. Fire Protection. Resolved by Branch Technical Position 9.5.1, and Regulatory Guide 1.120.

2294

Group II - Resolution Pending

1. Turbine Missiles: Tllrbine failures for past 16 years have been evaluated and a statistical probability analysis has been conpleted.

An ACRS letter (April 18, 1973) discusses the problems.*

2. Effective Operation of Containment Sprays in a LOCA:

Extensive documentation in topical reports.

Review and evaluation are required.

3. 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 on irradiated steels from the Heavy Section Steel Technology Program.

• • 4. Instn1nents to detect (severe) fuel failures - NRC document, "Fuel Failure Detection in Operating Reactors", B. L. Siegel and H. H.

Hagen.

Item ID covers limited failures.

M:>re work is required for the severe failure case to establish instrumentation criteria.

5. M:>nitoring for Excessive Vibration or Loose Parts Inside the Pressure Vessel: State-of-the-Art results appear promising.

M:>re work may be required prior to decision as to installation of equipnent.

6. Corrmon Mode Failures - Requirements for diverse components should be established.

An example of a partial solution to the problem of non-rando~, multiple failures is WASH 1270, "Technical Report on Anticipated Transients Without Scram" (Item IA-4).

7. Behavior of Reactor Fuel Under Abnormal Conditions: 'Ibis includes:

flCM blockage; partial melting of fuel assemblies as it affects reactor safety; and transient effects on fuel integrity. '!be PBF program will address some of these items.

8. BWR Recirculation Pump Overspeed During LOCA:

Decision required by ACRS-Regulatory Staff.

9. '!be Advisability of Seismic Scram:

Further stooies required to establish need.

10. 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].

L9CA evaluation roodel corrplete.

ACRS feels new cooling approaches should be explored.

• Regulatory Guide 1s in preparation.

• • Identified in the Conmittee's Report of April 16, 1976 as "Instn.111ents to Detect Fuel Failures."

2295

II TURBINE MISSILES Turbine failures for the past 16 years have been evaluated and a statistical probability analysis has been conpleted.

An ACRS letter (April 18, 1973) discuses the problem.

'lhree issues require answers to resolve the turbine missile problem:

(1)

'lhe first relates to the appropriate failure probability value1

-4 based on historical failures the probability is about 10.

Industry predicts a much lower failure probability based on inprovements in materials and design. To date the ACRS has accepted the more conservative value1 (2)

'lhe second issue is strongly dependent on turbine orienta-tion 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 layout1 however, lower angle missiles with non-optimum (tangential) turbine orientation have unacceptably high strike probabilities1 (3)

'lhe third issue is one of penetration and damage of structures housed in the containment. 'lhe limited experimental data pertaining to penetration of large irregularly shaped missiles are not sufficient to determine structural response to inpingement of turbine disc segments. ~st missile penetration formulas are not relevant to this case.

Sane experiments with irregular missiles might resolve this issue, particularly for older plants with non-optimum turbine orientations.

2296

11 EFFECTIVE OPERATION OF C'ONTAINMENT SPRAYS IN A UXA Review and evaluation are required of the variety of experiments which have been conducted on the effectiveness of various containment sprays on the removal and retention 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 ~irborne radionuclides, and quality of evaluative tests of the removal efficiencies of various sprays under the conditions of tenperature, 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 damag-ing equipment within containment. Studies using other spray additives, such as hydrazine, have been conducted. If compounds, such as this, have distinct advr,lntages, insofar as minimizing equipment damage in the event of inadvertent actuation, action should be taken to encourage their use.

2297

II K>SSIBLE FAIWRE OF PRESSURE VESSEL POST-LOCA BY THERMAL SHOCK Earlier nuclear reactor p~essure vessels subjected to fluences of 19 1-4 x 10 nvt, 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. There 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 rcegion of degraded properties may initiate and propagate a crack because of high local stresses near the surface. Analyti~ procedures indicate the stresses drop rapidly with distance through the wall so the flaw should not propagate beyond some limiting point. The 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 roodel.

These are planned under the HSST program.

2298

II INSTRUMENTS '10 DETEX:'r (SEVERE) FUEL FAILURES In the event of substaoti4l 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 envirorurent.

Instrwnentation capable of early warning and timely response may avert an incident be-coming an accident.

Instr1.U11entation related to such diagnostic purposes for limited fuel failure is being used on DDSt power reactors.

(See Item ID-1). Further work is required to establish criteria for similar instrumentation for severe fuel failures.

2299

II K)NI'lORI?<<; FOR EXCFSSIVE VIBRATIOO OR I.OOSE PARJ.'S INSIDE THE PRESSURE VE;SSEL I.oose parts monitoring can provide early warning of potential mechanical problems or failures within the pressure vessel and throughout the primary coolant circuit. Reactor vendors have developed monitoring systems:

however, general requirements remain to be established.

2300

II NON-RANI01 ftlJLTIPLE FAIWRES (FORMERLY "CCJM)N K:>DE FAIUJRE")

'!he term "conm)n 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 conponents has been proposed or required to avoid such failures. '!he concern of the ACRS is better expressed by the term "non-random multiple failures", which is intended to include not only the type of "cOOIIDn 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 carponents 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 then in all systems inportant to safety.

2301

II BEHAVIOR OF REAC'IDR FUEL UNDER ABNORMAL COODITIOOS

'!be 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 mre extensive fuel failure, pressure pulses, etc.

reactivity transients.

Similar behavior might occur in the case of

'!be ACRS encourages analytic mdeling 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.

2302

II BWR PtJitP OVERSPEED DURING A u:x:.A It is possible for a BWR recirculation pllllp to overspeed if a large break occurs at the appropriate position in specific piping. Con-servative estimates indicate substantial overspeed and possible failure of components with the generation of missiles. 'Jbe problem is being approached analytically and experimentally with scaled pllTips.

'lhe reliabiity of such protective measures as the use of decouplers between PllllP and rotor is under stlrly.

2303

II THE ADVISABILITY OF SEISMIC SCRAM

'!be ACRS has recomnended that studies be made of techniques for seismic scram and of the potential safety advantages and potential disadvantages of prorrpt reactor scram in the event of strong seismic mtion, say mre 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. '!be principal potential advantage identified arises from the greatly improved coolability of a core in the unlikel) event of a seismically induced IJX:A, should scram precede the IJX:A 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 power offsite should a severe earthquake occur.

2304

II ECCS CAPABILITY FOR FUTURE PLANTS The 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 rrore precise. With more experimental data a realistic and quantitative appraisal of ECC systems would lead to valid judgments on the 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 ioonitoring of low power peaking, and to improve those fuel assent>ly designs which 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 increased and combined with the results of the standard problems and the associated experiments.

Improved analytical methods would provide a basis for optilni zed ECCS.

2305

Group IIA - Resolution Pending - Items Since December 18, 1972

1. Control Rod Drop Accident (BWRs):

Calculations indicated that the reactivity response differs from earlier values.

New analyses are required including three dimensional effects.

2.

Ice Condenser Containments: Additional analyses are required to establish response during a LOCA, and to establish design margins.

3. Rupture of High Pressure Lines Outside Containment:

The possibility exists that failure of a high pressure line such as a steam pipe can prevent operation of critical safety components.

4.

PWR Punp Overspeed During a IDCA:

Problem arises in similar manner to that of BWRs (Item 8 Group II).

5.

Isolation of Low Pressure Fran High Pressure Systems: Assurance required that low pressure systems cannot inadvertently be inter-connected with a high pressure system leading to failure. There are potential interaction problems between Class 1 and Class 2 or Class 3 pressure connections.

6. Steam Generator Tube Leakage:

Partially resolved by issuance of Regulatory Guide 1.83 which addresses the concern from a pre-ventative point of view.

7. 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.

2306

IIA CCNl'IOL IOD DIOP ACCIDENT (BWRs)

Some uncertainties have arisen in previous calculations of this postulated accident, including the choice of negative reactivity insertion rate due to scram and the potential differences between a two dimensional and a three dimensional calculation. Particularly for the latter point, roore precise theoretical conparisons may be required to resolve the matter although probabilistic considerationss may be relevant.

2307

IIA ICE CDIDENSER CONl'AllfotENTS

'lhe ice condenser con~ainments have substantially smaller volume on the assl.lllption that the ice will condense the steam during a LOCA, thus pre-venting system overpressurization. 'lhe 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.

2308

IIA RUPl'URE OF HIGH PRESSURE LINES CXJTSIDE CCNl'AINMENT The functional or structural integrity of ca1p0nents required for safe shutdown and the maintenance of cold shutdown may be endangered by a failure of high pressure piping at certain locations outside of contain-ment.

Specific design and single failure criteria need to be developed to resolve this problem.

2309

IIA PWR PlJ,U> OVERSPEED DURII<<, A I.OCA It is possible for a PWR primary coolant PlJ1'P to overspeed if a large break occurs at the appropriate po~ition in specific piping. Conservative estimates indicate substantial overspeed and possible failure of conponents such as flywheels with the generation of missiles. 'lhe problem is being approached analytically and experimentally with scaled pumps.

'!he reli-ability of such protective measures as electrical braking of the pump motor is under stooy.

2310

IIA ISOIATION OF IOi PRESSURE FROM HIGH PRF.SSURE SYSTEM.5 Assurance is required that low pressure systems cannot inadvertently be intercomected with a high pressure system leading to a failure of the low pressure systems due to overpressurization. Both Class 2 systems required for ECCS and Class 3 systems used for many auxiliary coolant systems provide vital functions during normal and accident conditions.

Any conditions leading to intercomection could trigger a LOCA.

It is recognized that these system; mst be interconnected because of the functions they perform. Therefore, particular attention is required to valve reliability and reliability of valve actuating circuits to assure that these valves will not open while the primary system is at pressure.

2311

IIA STEAM GENERA'IOR TUBE LEAKAGE Normally the steam generator is not a critical conponent during a ux::A-ECCS.

However, a special case exists where the steam generator tubes have been degraded due to corrosion, wastage, etc. If the shock loads inposed 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 ECCS, preventing adequate cooling of the core. The critical number of tubes is relatively small.

A position such as one specifying a statistically significant level of nondestructive examination (NOE) might resolve this issue. The purpose of NOE would be to confirm that damage is not excessive; such examinations should minimize the possibility of catastrophic failure of a significant nunt>er of tubes.

2312

IIA PERIOOIC (10-YFAR) RE.VIE.W OF AU. PCM:R RFACIDRS In its report of June 14, 1966, the ACRS recomnended that periodic corrprehensive reviews be conducted of operating licensed power reactors by the NRC Staff. 'lhese reviews would be preceded by a conprehensive report by the operator which evaluated the past experience and the safety of future operation of the plant.

'lhe NRC Staff has maintained a continuing review of the safety of operating plants. In particular, as generic matters of potential safety significance arise, the appropriate operating reactors are asked to assess the relevance of the matter to each particular reactor. '!his is a necessary but different as~ct of the continuing surveillance and review of the safety of operating reactors than was envisaged by the ACRS in its recomnendation of June 1966.

'lhe Conmittee continues to believe both approaches are desirable and awaits the developnent of a program of periodic conprehensive reviews.

2313

Group IIB - Resolution Pending - Items Added Since February 13, 1974

• • l. Conputer 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.

2. Qualification of new fuel geometries: The 16xl6 and 17xl7 PWR, and 8x8 BWR fuels should undergo testing to meet Item 2 in Group IC and Item 7 in Group II.

3. Behavior of BWR Mark III Containments: Various aspects, including vent clearing, vent/coolant interaction, pool swell, pool strati-fication, pressure loads and flow bypass should be resolved. This is an extension of Item 3 in Group ID.

4. Stress Corrosion Cracking in BWR Piping: Several failures have occurred in operating BWRs.

The 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.

• • Identified in the ACRS Report of April 16, 1976 as "Hybrid Reactor Protection System."

2314

IIB COMPOl'ER RFAC'IOR PIOTECTION SYSTF.MS**

'lbe 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 al:normal conditions, be established through an appropriate corrt>ination of tests and analyses. While the issue originated with the B&W Hybrid concept it is equally applicable to the proposed CE and~

computer reactor protection systems.

• • Identified in the ACRS Report of April 16, 1976 as "Hybrid Reactor Protection System. "

2315

IIB QUALIFICATICN OF NEW FUEL GEOMETRIES New fuels proposed for both BWRs and PWRs include the 8x8 (BWR), and 16xl6 and 17xl7 PWR fuels.

The Comnittee recognizes that these fuels are intended to operate at power densities lCMer than earlier fuel designs. However, testing programs are considered necessary to establish their densification behavior (IC-2) as well as their behavior under abnormal conditions (II-7).

Appropriate experimental programs should be developed dealing with flow blockage, behavior of fuel after partial melting and fuel response under transient conditions. It is anticipated that the solution of this item will include a synthesis of Power Burst Facility data, experiments on earlier fuel types, behavior of fuel in comrercial reactors, and confirmatory experiments on these fuel designs.

2316

IIB BEHAVIOR OF' BWR MARK III CONTAIR-tENT

'!he BWR Mark III Containment differs in many respects from 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.

2317

IIB STRFSS (X)RRJSIOO CRACKING IN 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.

'l.'tle austenitic stainless steels are corrroonly used as piping material in many of the smaller BWR lines. A combination of weld sensitization, residual stresses, superposed loads, and oxygen equal to or greater than 0.2 ppn in the BWR coolant can lead to cracking, initiating on the inner surface and propagating through the wall. In mst cases there will be a leak well before pipe failure so there is adequate warning: however, one can postulate a LOCA caused by a guillotine break with minimal prior warning. current efforts are to minimize stress corrosion by using other materials.

2318

Group IIC - Resolution Pending - Items Added Since March 12, 1975

1. Locking out of EC:S Power Operated Valves: 'lhe COnmittee suggests that further attention be given to procedures involving locking out electrical sources to specific rootor-operated valves required in the engineered safety fm1ctions of ECCS.

2. Design Features to Control Sabotage: Attention should be given to aspects of design that could improve plant security.

3. Decontamination and Deconmissioning of Reactors: Specific plans should be developed, including definitive codes and standards covering plant decomnissioning. Also experience should be gained in reactor decontamination so that such information is available when needed.

4. 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.

S. Water Hanmer:

Several cases of water slugging or water hanrner have occurred in both PWRs and BWRs.

Corrective measures should be taken to minimize such events.

6. Maintenance and Inspection of Plants: Provisions should be included in the design of future plants which anticipate the maintenance, inspection and operational needs of the plant throughout its service life.

7. Behavior of BWR Mark I Containments: Various aspects relevant to the BWR 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 establishnent of torus water tenperature limits during a LOCA.

'Ibis is an extension of Item 3 in Group ID.

2319

IIC I.DCKING OOT OF OCCS POWER-OPERATED VALVES The physical locking out of electrical sources to specific rotor-operated valves required in the engineered safety functions of ECCS has been required, based on the assunption 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.

'!he 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 signal; (2) time required to reactivate the valve operator; (3) status of signal lights when the ciruit breaker is open; (4) the possibility of locking out in an improper position due to a faulty indicator; (5) other designs with irrproved reliability without lock-out; (6) the advantages and disadvantages of corrective action by an alert operator in case of incorrect positioning vis-a-vis a system with power locked out.

2320

IIC DFSIGN FFATURF.S '10 ~L SABC1l'AGE Considerable attention has been devoted to control of industrial sabotage of nuclear power plants, particularly with regard to control of unauthorized access, and potential roodes of sabotage by individuals or groups external to the operating organization. 'lbe ACRS believes that deliberate attention should be given to aspects of design that could inprove plant security. With the enphasis being placed on standardized plant designs, it becomes especially irrportant to introduce design measures that could protect against industrial sabotage, or mitigate the consequences thereof.

2321

IIC DE~AMINATION AND DECXJo1MISSIONING OF REACTORS

'Ihe Carmittee believes that well developed plans, confirmed by appropriate experiments where necessary, should be available to cover such events as the decontamination of primary reactor systems and the decommissioning of reactors. At this time the information on full scale decontamination is quite limited, particularly for BWRs.

Examples of potential problems include such items as handling of decontamination solutions, potential hideout of radioactive products, enhanced corrosion and croo formation following decontamination, and the possible imcompat-ibility of the different alloys in the pressure boundary to the decontamination solutions.

Experience is limited with regard to decomnissioning operations including rules for dismantling and for nothballing. 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 Goverrunent, etc.

2322

IIC VE$EL SUPPORI' STROCTURES A possible consequence of the instantaneous double-ended pipe break pos-tulated to occur in certain large pipes of PWRs is the asymnetric loading of the reactor pressure vessel support structures. 'lhe magnitooe and effects of such loads on the pressure vessel should be determined to establish if such loads adversely affect the predicted course of a LOCA.

If analysis indicates that the results are lD'lacceptable, appropriate corrective action should be taken. A potential 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.

2323

IIC VESSEL SUPPORI' STRUCTURES A possible consequence of the instantaneous double-ended pipe break pos-tulated to occur in certain large pipes of PWRs is the asymnetric loading of the reactor pressure vessel support structures. '!he magnitude and effects of such loads on the pressure vessel should be determined to establish if such loads adversely affect the predicted oourse of a LOCA.

If analysis indicates that the results are unacceptable, appropriate corrective action should be taken. A potential 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.

2324

IIC WATER HAMMER Several instances of water slugging or water hanmer have occurred in both BWRs and PWRs due to causes such as the trapping of water between two valves.

'Ibis slug of water is accelerated by steam or water once the valves are opened. 'lbe stored energy is sufficient to damage piping, bend or break pipe restraints, and damage support structures. water hanmer may occur due to flow instabilities in steam generators in conjunction with water flowing into the feedwater inlets, resulting in conparable damage.

Corrective measures should be taken to minimize such occurrences after corrpletion of analytic and experimental stooies directed to an understanding of the causes.

2325

IIC MAINTENANCE AND INSPEX::TION OF PLANTS Experience with older plants has verified that appropriate roodifications in piping layout, with respect to walls and structures, type of insulation used, and weld joint design, to cite sone obvious items, lead to improved maintenance, roore reliable inservice inspections, and a better neeting of the operational needs of the plant throughout its service life, including decontamination and eventual decommissioning.

An additional benefit is the reduction in personnel exposures in plants making them roore amenable to maintenance and inspection. Appropriate changes should be considered in future designs to neet these criteria.

2326

IIC BEHAVIOR OF BWR MARK I CXNI'AINMENTS 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 conparable 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 tenperature 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.

2327

Group IID - RESOLUTION PENDING - Items added since April 16, 1976

1. Safety related interfaces between reactor island and balance-of-plant: The nuclear steam suppliers and some architect-engineers have submitted standardized-plant designs. 'lbe Comnittee wishes to be sure that adequate attention is devoted to the interface between the reactor island and balance-of-plant to minimize pro-blems during design and construction. The development and use of interdisciplinary system analyses is an aspect of this problem.

2. Assurance of continuous long-term capability of hermetic seals on instrumentation and electrical equipment: 'lbe integrity of seals during post-accident conditions may be critical in controlling such an accident. The Comnittee believes appropriate test and maintenance procedures should be developed to assure long-term reliability.

2328

IID SAFETY RELATED INTERFACES BE'IWEEN REACTOR ISLAND AND BALANCE-OF-PIANT Questions have been raised concerning both standardized balance-of-plant and nuclear steam supply systems on the one hand and custom-designed site-related structures and components on the other hand. '!he depth of detail required at the stage of Preliminary Design Approval may not be adequate for construction approval. Procedures for instituting quality assrJrance programs covering design, procurement, construction, and startup with emphasis on timely and appropriate interdisciplinary system analyses to assure functional compatibility across the interfaces as well as for other systems are necessary to assure functional compatibility for the postulated design basis accident conditions.

2329

IIIr ASSURANCE OF ~INUOOS I.OG-TERM CAPABILI'IY OF HERMETIC SFALS 00 INSTRUMENTATIOO AND ELECTRICAL BJ{JIPMENI' Certain classes of instr1.111entation incorporate hermetic seals. \\tlen safety related conponents 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 aging. 'lhe damage processes may fall within Item IB-3, "Performance of Critical Conponents in Post-LOCA Environment", however, a special case requiring evaluation has to do with persoMel errors in the maintenance of such equipnent since such errors could lead to the loss of effective hermetic seals.

2330