ML19289E494
| ML19289E494 | |
| Person / Time | |
|---|---|
| Issue date: | 12/04/1978 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Lawroski S Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML19289E490 | List: |
| References | |
| ACRS-1600, TAC-8162, NUDOCS 7904210131 | |
| Download: ML19289E494 (59) | |
Text
,.
[p arcb UmTto STATES g,'k NUCLEAR REGUL ATORY COMMISSION
.'h WAsHitJGTof4. D. C. 20bL5 DEC 4 b/8 Dr. Stephen Lawroski, Chairman Advisory Committee on Reactor Safeguards S
U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Dr. Lawroski:
This refers to the Committee's letter of November 15, 1977, su bject,
" Status of Generic Items Related to Light-Water Reactors:
Report No. 6".
Enclosed are brief reports on the present status of resolution of each of the " unresolved" items noted in the November 15, 1977 report.
As in our previous status report of May 4,1978, wc have continued to subdivide your item IID-1, " Safety Related Interfaces Between Reactor Island and Balance-of-Plant" into IID-1 A to address the " Safety Related Interfaces Between Reactor Island and Balance-of-Plant" and a IID-1B to address " Systems Interactions".
We consider item IIB-2, " Qualification of New Fuel Geometries", to be rerolved. As you will recall, this issue was determined to be of generic interest at the time of the NSSS vendors were introducing the fuel element designs now in use.
Section 4.2 of the Standard Review Plan has been revised to require prototype testing, lead assembly irradiation, and fuel surveillance.
The specific
. requirements are dependent upon the extent of any future fuel design modifications.
These requirements will be applied to reload fuel for operating reactors as well as to new operating licenses, and are sufficient to assure that any proposed new fuel geometries will be properly qualified.
We also believe that item II-7, " Behavior of Reactor Fuel Under Abnormal Conditions", should be removed from the " unresolved" list.
The overall matter of fuel behavior is the subject of continuing effort both by NRR and RES, with periodic reports on the staff activities presented to the ACRS Fuels Subcommittee.
It is likely that these types of efforts will continue into t-foreseeable future.
The issue is broadly stated and, since
~ specific problems are identified and no specific conclusions are projected, it appears that the issue could safely be removea from the
" unresolved" list and carried as a continuing staff effort with periodic reports to the Committee.
790421 0131
_m
(
a DEC 4 ENO Dr. Stephen Lawroski.
Finally, in the interest of shortening the list of " unresolved" items, it is recommended that item IIA-2, "PWR Pump Overspeed During a LOCA" be incorporated into item 11-3, "BWR Recirculation Pump Overspeed During a LOCA".
The overspeed problem for PWR reactor coolant pumps and for BWR recirculation pumps is being handled by the staff as an integrated effort and we have reported on the status jointly for several reports.
G g
u Harold R. Denton, Director f
' Office of Nuclear Reactor Regulation
Enclosure:
As Stated e
e
.c
~
II-l TURBIflE MISSILES Problem Definition:
The probability of unacceptable damage to essential systems of a nuclear power plant due to turbine missiles can be estimated as the product of the probabilities of (1) ejection of an energetic turbine missile, (2) a missile striking a critical component, and (3) unaccept-able damage occurring to an essential system or component.
The task effort will evaluate models describ-ing the probabilities and analyze the effect
.of (1) turbine orientation, (2) component testing and inspection, (3) toughness of turbine disk materials, and (4) missile trajectory on the reduction of the overall damage probability.
Program:
Regulatory Guide 1.115, " Protection Against Low-Trajectory Turbine Missiles," Revision 1, was issued in July 1977. The guide indicates that turbine orientation and placement relative to essential systems is an acceptable means of protection against low-trajectory turbine missiles. An exception is made for these systems within the low-trajectory tur Diae missile strike zones which subtend relatively small solid angles with respect to potential turbine missiles. The criterion for the exceptionisthatthegtrikeprobability is to be 1.ess than 10- per missile.
An NRC Task Force was formed in March 1976 to evaluate models describing the probability of missile generation, the strike probability considering effects of intermediate barriers, and the probability of unacceptable damage to essential systems. This generic item has been considered during the development of the staff's technical activities program.
It is included in the scope of Task Action Plans A-32 ar.d A-37.
The purpose of this task is (1) to assess the methods currently used to estimate the
~
probability of damage by turbine missiles to the essential systems of a nuclear power plant, (2) to quantify the effect of the various steps that.can be taken to reduce the damage probability, and (3) to recommeid adoption of specific requirements on turbine fabrication and operation which will assure that the overall damage probability is sufficiently low.
II-1 TURBINEMISSILES(CONT'0)
A NUREG report is currently under prepara-tion which will recommend specific requirements on turbine fabrication and operation which can reduce the overall damage probability.
\\
e 9
e e
e 4
e 2-
- ....I...
(
- w.-
11-2 EFFECTIVE OPERATION OF CONTAINMENT SPRAYS IN A LOCA Problem Definition:
Determine the effectiveness of fission product remova' by containment spray systens following i LOCA, and evaluate systems capacities needed based on this information.
Program:
- ANS 56.5 (ANSI N581), "PWR and BWR Containment Spray System Design".
is under development.
Draft 8 has been re-viewed by the staff and approval has been recommended.
It is expected to be approved by ANS-50 in the near future. The Office of Standards Development is p'aparing a Regulatory Guide to endorse - is standard in whole or in part.
Meanwl..le, a docu-ment entitled " Technological Bases for Models of Spray Washout of Airborne Con-taminants in Containment Vessels" (NUREG/CR-0009), has been published by the staff. This document summarizes pertinent theories, principles and experimental re-suits in the design and review of containment spray regarding its fission product removal function.
This generic item has been considered during the development of the staff's technical activities program.
It is in-cluded in the scope of Task Action Plan C-10.
o II-J POSSIBLE FAILURE OF PRESSURE VESSEL POST-LOCA BY THERMAL SH0CK Problem Definition:
Cold wa:er injection from the ECCS operation during a postulated LCCA may result in a thermal shock to, and cracking of, the reactor vessel.
Program:
The HSST program has been the principal source of data generated for the purpose of resolving the reactor pressure vessel thermal shock problem.
Experiments to con-firm the analytical model used in the safety assessment of hot reactor pressure vessels subjected to thermal shock effects due to injection of cold ECCS water following a LOCA have been conducted at the Oak Ridge National Laboratory.
The resulting data have led to the following conclusions.
1.
Crack extension can be induced experimentally by thermal shock.
2.
The observed crack instabilities c.ta be reasonably predicted by fracture mechanics.
3.
Crack arrest, short of catastrophic failure, was observed in tested model pressure vessels (actually open-ended cylindrical sections).
The entire subject of thermal shock has been and is under review by 00R and S0.
Based on review of all available relevant data, comments and recommendations regarding past and future thermal shock experiments have been expressed.
The review and analysis of the experimental program is not yet complete and much of the effort falls within the scope of Task Action Plan A-ll, " Reactor Vessel Materials Toughness".
The thermal shock tests at ORNL have employed cylindrical shells of 21 inch OD and 5.75 in wall.
In their analysis of the thermal shock transients, RES concluded that warm prestressing would occur and limit crack extension.
Experiments conducted at NRL were interpreted to be evidence that
'e~'
'c
(
.:T.
o warm prestressing would increase the re-sistance to fracture under conditions representative of thermal shock in a reactor pressure vessel.
The experiments involving model vesseis have several aspects unresolved relative to the applicability of the results to full-scale reactor vessels.
The model vessel geometry, having a relatively low R/T ratio, was much less flexible than a full-scale vessel.
Thus e observed crack arrest in the model vesseis might be related to the geometry and not necessarily applicable to full size vessels.
Practical limitations of hydraulics and heat transfer preclude reaching sufficiently large strains (elastic stresses) to model the postulated reactor accident cond1tions.
Tests perforned to date have provided soiw confidence that the analytical methods used are satisfactory. Additional ex:3eriments are needed to confirm the analyt il models because of parameters such as Ri
- ratio, crack geometry and warm prestressing.
The NRC concludes that additional experimental data are needed.
Collaboration between the responsible parties is leading toward defini.
tion of a program extension.
One soecific suggestion currently being evaluatea is to increase the model vessel OD to 39 inches.
If a thermal shock facility could provide reasonably severe conditions to such a vessel with a 4 inch wall, the geometry would be much cle;er to typical vessel configurations.
s II-4 INSTRUMErlTSTODETECT(SEVERE)FUELFAli.URES Problem Definition:
Determine whether instruments capable of directly detecting fuel rod failure in additibn to the instruments that monitor the process pai m eters of pressure, tempenture and flow could provide addi-tional protection by giving timely warning of conditions that might, if undetected, result in more serious consequences.
Program:
Sensitive instruments that can detect fuel failures (see NUREG-0401) are present in many plants and their use is required by technical specifications in some operating reactors.
Revision 1 of Section 4.2 of the Standard Review Plan addresses the use of such instruments for all new reactors. When these instruments are used with other plant instruments that sense coolant pressure, temperature, flow, etc., all events or accidents that lead to fuel failures are detectable.
The instruments described in NUT -0401 derive their sensitivity in par'
- .m time delays of a few minutes tha
.ermit background activity to decay.
Tb, are therefore not useful for providing reactor scram signals, but they are not needed for this function because other plant instruments, which do generate scram signals, sense the accidents that cause the failures.
It has been suggested that inlet flow blockage in a BWR is an exception and that fuel failures during that accident could progress without quick detection.
BWR Flow blockage is being evaluated by HRC. The GE topical report NED0-10174, Rev. 1, " Consequences of Postulated Flow Blockage Incident in a Boiling Water Reactor," October 1977, has been received by NRR and is under review. The report presents analytical models which describe 0
&. 9
B o
II-4 INSTRUMENTS TO DETECT (SEVERE) FUEL FAILURES (CONT'D) the effects of flow blockage and bench-type tests to support the models.
Blockages severe enough to cause fuel melting are unlikely because lower plenum velocities are too low to lift large metal objects. The probability of lesser blockage is not well defined.
The probability and consegeunces of these lesser blockages will be studied as part of the staff review.
Flow blockage research programs have been planned by RES. Tests in PBF, NRU, SGHUE and ESSOR are being considered and may begin in 1981.
Improvcd failure detection systems are currently being tested in PBF.
9 s
II-5A MONITORI,,.1G FOR LOOSE PARTS INSIDE THE PRESSURE VESSEL Problem Definition:
Loose parts monitoring can provide early warning of potential mechanical probit or failures within the pressure vessei nd throughout the primary coolant circuit.
However, experience with available detec-tion system equipment is needed to establish the state of the art and to aid in the definition of licensing requirements for these systems.
Program:
All plants currently being licensed are required by the Standard Review Plan to in-stall Loose Part Monitoring systems (LPMS) acceptable to the staff.
At the CP level the applicant supplies general information on sensor location and mounting methods; while at the OL level a detailed system description is required.
However, pending complete definition of an acceptable LPMS, the staff still accepts a commitment by the applicant to install such a system v: hen it becomes available.
A regulatory guide on LPM (Propcsed Regulatory Guide 1.133) has been written, reviewed, and received public comment.
Implementation of the Guide is presently scheduled for June 1,1979.
This regulatory Guide will establish the minimum staff requirements for the accepta-bility of both an LPMS and the licensee's program for utilizing the LPMS.
.In addition, 00R is investigating the effectiveness of LPMS techniques currently being utilized at operating reactor facilities to detect loose parts.
The equipment now in use is generally less sophisticated than that required by Proposed Regulatory
- Guide 1.133, and i' therefore characterized by a lower sensi'a vity and a higher false alarm rate.
Nevertheless, these older systems have been used successfully.
Past experience with loose-part events indicates that the domestic industry will average aporoximately two significant loose-part events per year for the present, and still more '
the future as more reactors.
c' come on line.
In the event that a loose part goes undetected, it could result in mechanical damage, inte.rference with moving parts and coolant flow blockage.
Loose parts massive enough to threaten the pressure boundary are not expected to occur.
Inter-ference with more than one control rod is likewise not anticipated due to exercise and testing programs already in effect.
Fuel failure due to fretting or due to ONB in-duced by coolant flow blockage could result in release of gap activity to the reactor coolant system. This activity would be released through the liquid and gaseous waste treatment systems. Thus, it would not necessarily involve an inadvertant release of activity to the environment.
In the event of an inadvertant release to atmosphere of gap activity due to either fretting or flow blockage, the activity released would be extremely small when compared to the release associated with design basis accidents 5:ch as loss-of-coolant or control-rod-ejection.
Loose parts are therefore more significant for reactor availability and preventing damage, rather than being a sig-nificant contributor to risk.
Additionally, an NRC/RES Research Program leading to the development of more advanced methods for determining the size and location of loose parts. commenced in June 1977 at ORNL.
All reactor '/endors and several independent manufs:+"rers are in the process of develop-ing an'd aproving monitoring systems of this type.
This generic item has been considered during the development of the staff's technical activities program.
It is included in the scope of Task No. B-60..
t,
^
^(
11-5B MONITORING FOR EXCESSIVE VIBRATICH INSIDE THE REACTOR PRESSURE VESSEL Problem Definition:
Neutron noise analysis can detect vibration within specific ccmponents such as the core barrel'. The detection of vibration in otht?
reactor pressure vessel components is less welt established. Development of vibration detec-tion equipment could enable the industry to obtain early warning of potential problems caused oy excessi';a vibration of reactor pressure vessel components.
Program:
Identification of vibration as a special problem occur.aed in the ACRS, " Status of Generic Iteas Relating to Light-Water Reactors:
Report No.
6", issued November 15, 1977.
Until that time, vibration monitoring had been associated with loose parts monitoring.
However, it is apparent that different equipment is required to adequately monitor the two phenomena.
This subject has been approved by the Technicel Activities Steering Committee as a 'ategory B Task. Development of a tr.sk action plan and efforts t; ward rerolution will commence as necessary rcsources become available.
s o
c-i l
II-6 CO? HON MODE FAILURES Problem Definition:
This heading covers a multiplicity of diverse components for which require-ments should be established.
Due to their diversity, the ACRS feels that specific items should be separated into subsets under the general heading of common mode failures.
Program:
Progress on resolution is reported for each subset identified by the ACRS.
9 k
{
L' II-6A REACTOR SCRAV. SYSTEMS Problem Definition:
A need to establish requirements for limiting the consequences of failure of reactor scram system due to common mode failure and improv-ing the reliability of the scram systems with respect to common mode failures.
Program:
The NRC published reports on anticipated transients without scram in December 1975 in which they identified the portions of reactor scram systems that needed modifications to improve the reliability cf scram systems.
In addition, these reports provided guidelines on evaluation model, analysis assumptions, system reliability requirements, and acceptance limits.
EPRI and vendors have recently sub-mitted additional analyses and in some cases further reports on scram system unreliabilities.
The staff has reviewed these studies and has published its findings in a report HUREG-0460, " Anticipated Transients Without Scram for Light Water Reactors" dated April 1978. A supplement to NUREG-0460 uill be issued in December 1978.
This report is currently being reviewed by the ACRS and the Office of Nuclear Reactor Regulation.
After completion of the review, now estimated by January 1979, the Director, NRR will forward its recommendations to the Commission.
This generic item has been considered during the development of the staff's technical activities program.
It is included in the scope of Task Number A-9.
~
/
II-6B N0ft-RANDOM MULTIPLE FAILURES - ALTERNATING CURRENT SOURCES Problem Definition:
Licensing review procedures and requirements need to be developed to assure identifica-tion of and protection against potential "non-random multiple failures" of alternat-ing current power sources.
Examples of potential "non-random multiple failures" include sequential multiple failures (domino effect) due to a single fault and simultaneous multiple failures due to a single fault.
Program:
OFFSITE POWER SYSTEliS The Commission's requirements, as defined in (1) General Design Criterion 17 (GDC-17),
" Electric Power Systems", (2) the NRC's Standard. Review Plan, (3) Regulatory Guide (RG) 1.93, " Availability for Electric Power Sources", and (4) the Technical Specifica-tions for individual nuclear power plants, establish design and operating criteria for electric power sources at nuclear plants.
The objective of the Commission's design and operating requirements is to maximize the probability that offsite power will be available to the reactor facility.
For example, two physically independent circuits from the offsite transmission network to the plant are required.
It is recognized, how-ever, that a loss-of-offsite cower event cannot be precluded, and there"cre, a redundant onsite emergency se:.
2 of power is also required. The onsite
-;er source is considered an engineered sc ety feature and must satisfy the Commission's require-ments for quality, reliability, independence, etc., commensurate with its safety function.
As part of the scope of NRR Technical Activity No. A-35, "Adequasr of Offsig Power Systems", the NRC staff is evaluating the need, iY any, to upgrade the offsite power source and its interface with the onsite power system at licensed nuclear power stations and for license applications.
The results of these tasks will serve as the input and bases for any modificaticos that may be required to our existing licensing r'
criteria relative to:
(1) monitoring grid conditions to identify when a grid would be vulnerable to a subsequent contingency (failure); (2) additional procedural actions or requirements that may be taken within the nuclear plant when the grid is determined to be vulnerable to natural events or to grid system equipment failures, e.g., the licensers and applicants could be required to start their onsite diesel generators: (3) design changes which can provide a dedicated off-site power source m selected nuclear plants; (4) design changes ;o provide additional protection for recuncant safety-related equipn nt from sustained voltage degradation of the offsite source; (5) detennination of the adequacy of existing testing requirements for the onsite power sources; and (6) the reliability of the various designs, consider-ing the effects of transient and degraded grid c.1ditions, for connecting the offsite power from the switchyard to the in-plant emergency busses.
The staff will prepare a report in the form of a flVREG document which will provide com-plete documentation of the details, conclusions and any new or augmented criteria developed as the result of the staff's implementation of this task actior lan re-lating to offsite power.
The NURE report is currently scheduled for completion by July 15,1980.
EMERGENCY ONSITE DIESEL GENERATORS In an effort to improve the reliability of the diesel generato.s, NRC has contracted with an experienced qualified outside con-sultant (University of Dayton) (1) to perform a study of Licensee Event Reports (LERs) related to diesel generator malfunc-tions (2) to make a limited number of visits at operating facilities, (3) obtain the manufacturers' recommendations regarding operations, maintenance and repair of their equipment, and to survey comparable indus-trial experience with standby emergency :
power supplies.
A preliminary finding of the review of the LERs is that some reported failures apoear to indicate the possibility of non-random multiple failures.
The con-tractor has been instructed to be alert and to report such possibilities in the final report containing his findings, evaluation and recommendations.
Typical areas where non-random multiple failures may exist are as follows:
turbo charger failures due to differences in the initial engine design re-quirements and nuclear plant operating practices, inadequate combustion air filters during adverse environmental conditions, multiple electrical relay failures due to the accumulation of airborne particulate matter o.n the relays, exceeding the endurance life of specific components or otheraise cause engine degradation as a result of accumt; lated operation time under certain conditions.
This generic item has been considered during the development of the staff's technical activities program.
It is included in the scope of flRR Technical Activity flo. B-56,
Diesel Reliability".
This activity is scheduled for completion by January 30. 1979.
/d
$4 1W 3 y
F ' y ll c'
Il-6C DIRECTCURRENTSYSTEM Problem Definiticn:
Questions have been raised regarding the adequacy of the current staff position on D.C. p6wer supplies which is summarized in HUREG-0305, " Technical Report on D.C. Power Supplies in Nuclear Power Plants". The staff considers that the questions raised warrant reexamination of the staff's requitements for D.C. power systems.
The minimum acceptable D.C. power system in Program:
accordance with NUREG-0305 is comprised of two physically independent divisions which supply D.C. power for control and actuation of redundant safety-re'.ateo systems. The questions that have been raised concern (1) the dependence on D.C. pc :r of the decay heat removal systems whi-
- re required for long-term heat removal, the fact that failure of one D.C. divir.:n of a two-division redundant system would generaliy result in a reactor scram which then would require removal of decay heat and therefore depends upon the remaining division for D.C. power supply, and (3) questions raised regarding the frequency of reported single D.C. division failures including those resulting from human error.
These considerations warrant reexamination of the staff's design requirements for D.C.
power systems.
A full description of this problem and the plan for resolution are contained in the staff's Task Action Plan _6-30 Under this plan the first four tasks are essentially complete. These tasks have expanded on NUREG-0305 in terms of data base, recalculation of allowable times for manuai actions, and finer definition of the spectrum of concerns that accompany total loss of D.C. power. The next phase of the plan includes quantifying D.C. power syst m reliability in relationship to assuring adequate decay heat removal capability; -These analyses will be based on c'
II-6C DIRECT CURRENT SYSTEMS (CONT'D) reliability data for various systems and components, using Reactor Safety Study methodology.
The end product of this program will be a NUREG report which will provide complete documentation of the analyses performed and develop a staff position regarding the adequacy of the existing acceptance criteria for D.C. power systems. Completion is scheduled in mid-1979.
e e
N
)
II-7 BEHAVIOR OF REACTOR FUEL UilDER ABNORMAL CONDITI0tlS Problem Definition:
A more complete experimental determination Sf fue,1 behavior under abnormal conditions, such as flow blockage and reactivity transients, is desired.
Program:
Licensing activities that address fuel behavior under abnormal conditions focus on fuel failure (i.e., cladding defect) criteria and coolable geometry requirements.
These evaluations, which are usually analytical, are supported by RES experimental programs.
Revision 1 of Sectinn 4.2 of the Standard Review Plan sharpens the definition of the important events, and close cooperation with RES is resulting in corresponding changes in their fuel research programs. That is, both NRR and RES have large and active programs addressing fuel behavior under abnormal con-ditions.
Generic Item II-7 specifically mentions flow blockage and fuel melting.
Flc,: blockage is being addressed by NRC so a low priority basis. RES has plans far in-reactor tests to begin about 198; a.d NRR is reviewing a BWR flow blockage.eport. The status of Item II-4 gives more detail on our flow blockage evaluation.
Generic Item II-7 is so broadly stated, however, that nearly all the fuel-related activities of NRR and RES could be considered to address this item.
No specific problems are icentified and no specified conclusions are projected.
We therefore believe that since the status of the NRC fuel programs are regularly dis-cussed with the Fuels Subcommittee, Item II-7 should no longer be carried as an unresolved generic item.
1
<lv -
a($.-)-
.M 2
II-8 BWR RECIRCULATI0ft PUMP OVERSPEED DURIflG A LOCA Problem Definition:
Investigation of the potential generation of missiles due to the disintegration of the recirculation pump or the reactor coolant pump from overspeed as a result of loss of line pressure in a LOCA.
Program:
Topical reports on pump overspeed have been submitted by General Electric, Combustion Engineering, Babcock & Wilcox and Westinghouse and are being reviewed by the flRC staff.
The initial review of these submittals indicated a need for experimental verifi-cation of analytical calculation EPRI has sponsored two-phase flow cump tests at CE and pump performance modeling at B&W.
Westinghouse is participating in tests being conducted by Framatome. The Phase I and Supplemental Phase II series testing of steam-water steady-state and blowdown condi-tions at CE using a one-fif th scale pump has been comple'cd.
Preliminary Phase I test data have l-
' submitted, and the publishing of Phase I:
.ta is in progress.
Final reporting of the CE L n program is expected during the first quartcr of CY-1979.
CREARE, under contract with EPRI, is continuing one-twentieth scale testing of two model pumps (CE & B&W).
During CY-1979 CREARE plans to complete test-ing and analyze test data, and develop an analytical pump model based on physical phenomena they have observed. MIT has completed its contract with EPRI in investigating two-phase pump performance and analytical modeling of pumps.
B&W has completed an analytical two-phase pump performance model based on their previously conducted steady-state air-water tests using a one-third scale pump.
Westinghouse has completed their steam-water steady-state tests using a one-third scale pump.
Data reduction and empirical pump model dev21opment is underway.
Estimated completion of Westinghouse work is during tN first quarter of CY-1979.
~
(
'd
.t.
r _, _
BWR RECIRCULATION PUMP OVERSpEED DURING A LOCA (CONT'D)
TI-8 GE has submitted a reanalysis of a BWR recirculation pump overspeed during a GE has revised its previous con-LOCA.
clusion and now concludes that a decoupler between the pump and motor is not required since potential missiles resulting from overspeed of both the motor or impeller would be contained within the motor housing or pump casing. The staff review completion is now projected for the second quarter of CY-1979.
The staff has asked each PWR vendor to submit its most recent prediction of pump overspeed during a LOCA in order to reassess the potential for pump flywheel failure and the necessity, practicality, and validity of electrical braking or other means of controlling pump speed.
Westinghouse has stated that their most recent analyses of a RESAR-3S plant show that the pump will not exceed 170 percent of rated speed following a doubla-ended guillotine break with a Moody multiplier of 0.6.
This result is significantly lower than predicted in the topical report. The most recent analyses per-formed by B&W were based on the EPRI contract which showed a small reduction in overspeed when compared to the topical report results.
CE has stated that their more recent analyses use an INEL model two-phase flow homologous head degradation which was r.ot included in their original
~
topical report. Based on the approved ECCS Evaluation Model, CE predicted approximately 400 percent of rated pump speed with a flow discharge coefficient of 1.0.
However, for a more mechanistic calcula-tion which limited the break offset, CE predicted a significant decrease in pump overspeed. _These responses wi.ll be considered with the overall testing program termination and data evaluation.
r q'
s II-8 BWR RECIRCULATION PUMP OVERSPEED DURING'A LOCA (CONT'D)
The staff is performing some independent reactor coolant pump overspeed calcula-tions 'during a LOCA using the RELAP 4/ MOD 5 computer code. The preliminary EPRI sponsored pump test results are used in the evaluation. The study results will be obtained during the CY-1979.
Staff efforts on this problem are included within the scope of Task Action Plan B-68.
0 0
_~
(
II-9 THE ADVISABILITY OF SEISMIC SCRAM Problem Definition:
Whether or not to require an automatic scram system which would be triggered by a preset seismic level.
Program:
The final report by Lawrence Livermore Laboratory, UCRL-52156, " Advisability of Seismic Scram", was reviewed by the staff and on May 19, 1977, a letter from E. G. Case to Myer Bender advised the ACRS of the staff's conclusion that we do not propose to require installation of seismic trip systems on commercial nuclear power plants. The letter further indicated that the staff considered this matter to be adequately resolved.
Staff members met with the Regulatory Activities Subcommittee on June 8,1977 to discuss this matter.
The Subcommittee comments, later documented in a letter from M. Bender to E. G. Case dated June 14, 1977, were that perhaps the selected seismic trip level should be set at about one-half the SSE, which could change the conclusions of UCRL-52156.
Further, the Subcommittee expressed interest it what the Japanese are doing in regard to ' :smic scrams.
Based upon the Commictee's June 14, 1977 letter, the staff has attempted to ascertain the position of the Japanese regarding automatic seismic scram systems.
In July of 1977, the staff requested information from the Japanese regarding their requirements for seismic scram and the bases for these requirements. We also requested the views of the Japanese on the UCRL-52156 study.
Ta date we have received no formal response to this request.
However, we learned during a staff visit to Japan in November 1977, and confirmed during a visit to the U.S. by a Japanese delegation in June 1978, that the Japanese do require the installation of seismic scram systems.
Trip levels are set at what corresponds to 1/2 to 2/3 of the SSE design level.
This generic iten has been cnnsidered during the development of the staff's technical activities program.
It is included in the scope of Task No. 0-1.
11-10 EMERGENCY CORE COOLING SYSTEM CAPABILITY FOR FUTURE PLANTS Program Definition:
Explore new cmergency core cooling approaches for application to future plants, based on diversity rather than just redundancy of systems.
Program:
As part of the Plan to Improve the Safety of Light-Water Nuclear Power Plants, an extensive analytical and experimental research program to investigate alternate emergency core cooling (ECC) concepts will be conducted.
The Semiscale Facility will investi-gate alternate ECC concepts which have the potential to improve the safety of commercial PWR plants.
The purpose of this investigation is to provide greater assurance that the cooling water injected into the primary system will reach and flood the core in a timely and ef fective manner.
This may be accomplished by reducing the present reliance on difficult and complex calculations and thus increasing the accuracy of the analyses of ECC system performance in. postulated loss-of-coolant accidents in light-water reactors.
This experimental and methodology development projemt is scheduled to begin in FY 1980 and requires two to three years to complete.
A program plan has been formulated which includes a three stage approach to evaluate ECC concepts.
The first stage is the collection of ideas and-recommendations of improved ECC concepts and a preliminary evaluation leading to the selection of promising concepts.
A review of current research programs will be incorporated to determine the relative merits of several ECC concepts.
Typical of the concepts to be evaluated for both PWRs and BWRs are:
1.
Current ECC injection systems, 2.
Lower plenum injection, 3.
Variations in upper plenum injection, 4.
Flow limiting nozzles in primary coolant
- piping, 5.
Check values and steam vent valves, 6.
Alternate pumping devices, 7.
In-core spray, 8.
Combinations of ECC injection locations, and 9.
Steam generator as accumulator (PWR only).
g.
- 23
(
L...
/
The second stage involves additional evaluations and analyses and experimental work, where appro-priate, of the identified concepts to minimize the number of concepts to be considered in stage three.
The latter stage represents the majority of the work.
It consists of in-depth analytic and experi-mental investigation to obtain quantitative behavior of the most promising techniques.
The specific tasks required to complete each stage of the program are delineated as follows:
Stage One:
Concept Identification and Selection (SixMonths)
Task I.
Concept Identification:
Survey of nuclear field to obtain concepts and experimental data.
Evaluate code capabilities and applications.
Task II.
Concept Selection:
A.
Risk Reduction Potential:
Identifica-tion and estimation of the benefit of the new concept, B.
Risk Increase Potential:
Identifica-tion and estimation of possible risk
~
increases due to application of the
- concept, C.
Applicability:
Assessment of the potential for use of the concept with new plants, old plants, plants under construction, with respect to different vendors, with respect to simplicity and compatability, of the impact with respect to licensing considerations, and whether the concept has other alternate uses of benefit to the plant, D.
Cost:
Economic considerations, including basic equipment, plant modification, operating costs,
e
.)
E.
Reliability: Assessment of availa-bility and operatien, F.
Evaluation Capability:
Assessment of our capability to quantify the performance of the concepts with respect to analytical and experimental techniques that are in existence or can be reasonably developed, G.
Technology and Effectiveness:
Evaluation of the capability of the concept to be developed into working hardware and the ability of the concept to provide adequate margin of safety, 11. Timeliness:
Estimate of the time required to develop the concept and for its implementation in working power plants.
Stage Two:
Preliminary Performance Appraisal (Six Months)
Task III.
Freliminary performance appraisal and concept selection.
A relatively short investigation with limited experimental and analytical work to further reduce the number of concepts.
Also includes limited additional work similar to Task II where appropriate.
Stage Three:
Final Appraisal (Two to Three Years)
. Task IV:
Code preparation and application:
Prepara-tion of code (s) as necessary for analysis of the ECC concepts and the application of the code (s) to technical evaluation of the concepts during loss-of-coolant events.
Task V:
Experimental Facility Preparation and Application:
Modification of facilities as required for experimental evaluation of ECC concepts and performance of the tests.
e Task VI:
Value Impact:
Evaluation of available information using techniques which are to be developed under a different research project. Additional effort of the type conducted during Task II will be performed as required.
Task VII:
Concept Selection and Recommendations:
Final selection of the prcmising ECC concepts and compilation of all informa-tion pertaining to the project in a final report.
Additional details of this program are contained in the Nuclear Regulatory Commission's Report to Congress entitled, " Plan for Research to Improve the Safety of Light-Water Nuclear Power Plants, A Report to the Congress of the United States of America," NUREG-0438, Office of Nuclear Regulatory Research, hRC, April 12, 1978.
e e
6 9
e
II A-1 ICE CONDEllSER C0!!TAlf!MEtiTS Problem Definition:
The interest is in verifying accident parameters during LOCA and verifying the established design margin for con-tainments using the ice condenser concept.
Program:
Under contract to the ?lRC, the Idaho National Engineering Laboratory (IflEL) has modified the RELAP-4 code for analysis of the short-term phase of ice condenser containment analysis. The modifications are contained in RELAP-4 (MOD 51 which has been used by flRC to evaluate ice condenser contain-ment design capability. The COMPARE-1 code, developed by the Los Alamos Scientific Laboratory (LASL) under ccntract to flRC, also has an ice condenser short term analysis capability.
This code is currently being used by the staff to evaluate ice condenser containment design.
A code for the long term post accident ice e melting analysis (CONTEMPT-4 MOD 2) has been developed at IrlEL.
Verification and maintenance of the code and ice condenser containment design evaluation analyses will be performed in CY-79.
This generic item has been considered during the development of the staff's technical activities program.
It is included in the scope of Task No. B-54.
e 9
(
II A-2 PWR PUMP OVERSpEED DURIflG A LOCA Problem Definition:
It is possible for 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 such as flywheels with the generation of missiles. The problem is being approached analytically and es 3rimentally with scaled pumps. The reliab:
.?.y of such protective measures as electr.tal braking of the pump motor is also part of the review.
Program:
See II-8, BWR Recirculation Pump Overspeed During a LOCA. Staff efforts on this pro-blem are included within the scope of Task Action Plan B-68.
ga 6
O*
4 D
9,
II A-3 STEAM GENERATOR TUBE LEAVAGE Problem Definition:
The integrity of the steam generator is important during a LOCA or steam line break accident.
Special cases exist where the ster generator tubes have been degraded due to c.,rrosion (wastage and cracking), fretting, cavitation or denting.
If the dynamic loads imposed by the LOCA cause a critical number of tubes to fail, the flow can retard refloca-ing of the core during ECCS water injection, preventing adequate cooling of the core.
In the event of a steam line break accident, tube failure would provide the potential for releases of radioactive material to the secondary coolant and leakage to the environm.ent.
Program:
The major portion of the NRC staff efforts involve the review and evaluation of tube degradation in operating facilities and the evaluation of investigations conducted by the NSSS vendors, EPRI supported contractors, and NRC supported technical assistance and confirmatory 2 search programs.
NSSS vendor" are currently conducting research -
grams designed to determine the structur integrity of steam generator tubes which ar_ subjected to various degradation necimn is ns.
In addition the NRC is funding a confirmatory experimental research program at Pacific Northwest Laboratory (PNL) to verify the burst and cyclic strengths c' de-graded steam generator tubes and to obtain leakage rate data. Results of these programs will be used to establish steam generator tube leakage rate limits and tube plugging criteria which will be incorporated into NRC Regulatory Guides and Standard Technical Specifications.
The NRC is currently reviewing and evaluating NSSS vendor's analyses of the probability and consequences of postulated MSLB and LOCA accidents concurrent with steam generator tube failures.
The purpose of these efforts is to determine (1) the maximum number of
Q f
t'
' ~~
" ' ~
.J tube failures which can be tolerated with-out undue risk to the public health and safety and, (2) the probability of degraded tubes failing during normal operam an or during postulated accidents.
Tht-efforts include evaluations of the effect of steam generator tube failures on offsit> doses and on safety related systems.
Several NRC sponsored programs related to these issues are currently in progress.
Idaho National Engineering Laboratory (INEL) is developing a corputer code to aid in the evaluation of the cffects of tube plugging on the predicted peal lad temperatures and on ECCS perform-ance following a postulated LuCA.
Brookhaven National Laboratory (BNL) is in the process of evaluating the impact of steam cenerator tube failures on the consequence of main steam line break accident.
Result:
these programs will be utilized to estab ;
im-proved inservice inspection criteri:
e NRC Regulatory Guides and Standard Technic.
Specifications.
Periodic inservice inspections of a statistically significant number of steam generator tubes in conjunction with a tube leak detection system provides reasonable assurance that a critical number of tubes will not fail during normal operating and postulated accident conditions.
Statistical studies, including an NRC sponsored program at Sandia National Laboratory, are being
'nducted to confirm the adequacy of the existing inservice inspection criteria and to optimize sampling schemes in accordance with results from the above mentioned conse-quences analyses.
Revision of Regulatory Guide 1.83, " Inservice Inspection of PWR Steam Generator Tubes" w'll incorporate re-sults of these programs. Study and develop-ment of eddy current testing t2chniques by NSSS vendors, EPRI, and BNL under NRC con-tract, will be incorporated into the statistical studies and should result in improved confidence levels and in improved techniques for steam generator tube inspection.
The effects of water chemistry and corrosion on steam generator tube degradation are be-ing studied by BNL, as well as by the NSSS vendors.
Improved requirements for secondary water chemistry, which greatly affects steam generator integrity, are being ceveloped, but may be dependent on steam purity require-ments.
The NRC staff is evaluating two utility's plans to replace several degraded steam generators.
NRC review of proposed steam generator replacement and/or retubing may require assistance from outside consultants.
Preliminary discussions have taken place with Westinghouse on an in-situ retubing concept and full scale mock-up test results.
West-inghouse will submit a topical report for NRC review on their retubing program.
This generic item has been considered duri r, the development of the staff's
- tech, il activities program.
It is included in the scope of Task Action Plans A-3, A-4 and A-a which address Westinghouse, Com-bustion Engineering, and Babcock and Wilcox steam generators, respectively.
O
(<
IIA-4 PERIODIC COMPREHENSIVE 10-YEAR REVIEW OF OPERATING POWER REACTORS Problem Definition:
In its report of June 14, 1966, the ACRS recommended that periodic comprehensive reviews be conducted of operating licensed power reactors by the NRC staff. The staff concurs that an improved program to assure th?.t operating plants retain essential sc r'y margins is desirable.
Further, a r.uns to determine how an operating plant corpares, at least in a qualitative sense, stith current safety requirements is desirable.
1he staff has initiated, with Commission cpproval, an alternate approach to periodic reviews which it considers more effective than periodic comprehensive reviews.
Program:
The Systematic Eval *4ation Program for
(:ierating reactors initially will involve
'i of the oldest plants; i.e.,
this will
'.lude all operating plants that have 10
- y. rs or more of operation.
The evaluation is e.>pected to confirm and document staff judp r.t that the safety margins for these plant re adequate.
Each plant also will be ar essed against current criteria on safety issues, and either a r
'onale for acceptance or plant backfittinc.
be developed.
The evaluation will inclue
..nat we understand to be the primary thrust c: the ACRS's recommend-ation, a review of the operating experience for each plant.
After the review of these 11 plants is completed, a decision will be made based on the results on whether the orogram should be extended to other operating reactors.
This generic item was considered during the development of the staff's technical activities program.
However, it.sas determined that it should be treated as a policy matter. requiring Commission approval, rather tnan a *?chnical activi ty.
II B-1 COMPUTER REACTOR PROTECTION SYSTEM Problem Definition:
New reactor protection system designs contain digital.as well as analog components.
The design, development, qualification, and inservice surveillance requirements for these new systems need to be established.
Program:
Three vendors ha"e proposed protectian system designs that incorporate digital computers.
The CE Core Protection Calculator System (CPCS) has been reviewed in conjunction with the Arkansas Nuclear One - Unit 2 operating license application. The B&W Reactor Protection System II (RPS-II) is described in a topical report (BAW-10075-N Rev.1) submitted in June 1974.
This system will be included in the Bellefonte reactor, for which an OL application has been submitted. Our review is scheduled to start in mid 1979. The Westinghouse Integrated Protection' System is being reviewed in conjunc-tion with RESAR-414.
Digital computer based prc tection systems have been considered during
- 11. development of the staff's technical activitiec program. This has been included in the scope of Task Action Plar A-19.
The results of the staff review of the CE CPCS have been discussed with the ACRS Subcommittee on June 30, 1977, March 20, 1978 and with ACRS on April 6, 1978. The staff's Safety Evaluation Report was issued on May 2, 1977, with supplements issued in March and June 1978.
The staff has identified six safety related items which remain outstanding in the applica-tion of the CPCS to AN0-2.
The staff has con-cluded that subject to the satisfactory resolution of the six safety related items, the CPCS is acceptable.
The staff published an evaluation of the acceptability of the design approach described in the B&W topical report on January 8, 1976.
Responses to requests for additional informa-tion were submitted as Revisions 2 and 3 in January and April 1977, respectively. An interim
--v
(
II B-1 COMPUTER REACTOR PROTECTI0fl SYSTEM (C0flT'D) report, describing the experience with a prototyp,e of the calculating module portion
~
of the system that has been operated in parallel with the Rancho Seco rear. tor protec-tion system for a year and a half, das submitted on Septr :er 1,1977. A report
'he reli-ability of the RPS II system har
-e was sub-mitted in March 1978. B&W advisa the staff in January 1978 of revisions to tne calculatini module software and verification tasting.
Some of the changes and test results were provided to the staff in October 1978. The romainder are scheduled to be submitted in December 1978.
Due to the proposed software design changes, our review of the system has been delayed until the revised software program and testing infor-mation has been provided.
The design bases and criteria of the Westing-house IPS will be addressed in the review of Chapter 7 of the RESAR-414 SSAR. The first round of questions were transmitted on September 15, 1977. The second round of questions were transmitted on December 20, 1977. Our Report to the Advisory Committee on Reactor Safeguards containing our evaluation of the IPS was issued in July 1978 and the RESAR-414 SER was published in flovember 1978. We concluded that the safety system design criteria and design base; infor-mation for the RESAR-414 safety syster.' nets the Commis:: ion's requirements. A rept NUREG-0493 "A Defense-In-Death Assessment of a
RESAR-414 Integrated Protection System" vnich addresses the questions of potential interactions between control and safety systems, will be issued in December 1978.
The functional adequacy of the IPS will be evaluated through the review of the series of topical reports WCAP-9163 "IPS Prototype Verification Program", WCAP-8897 " Bypass Logi-
~
for the Westinghouse IPS" and WCAP-8899 " Control System Signal Selection Devices". The review of these reports began in mid 1978 and is expected to require one and half to two years to complete.
t IIB-2 QUALIFICATION OF NEW FUEL GEOMETRIES Problem Definition:
Establish the requirements for new fuel elements design for normal operating conditions, anticipated transients and accident conditions.
Program:
Revisi -; 1 of Section 4.2 of the Standard Review Fian calls for the review of a progrcr., of prototype testing, lead assembly irradiation and fuel surve'11ance.
The SRP makes it clear that tra extent of surveillance will depend on tne extent of any new design featu cc th.. have been intro-duced. These require.. ants are intended to he applied to operating reactors as well as to new operating licenses. We thus consider Item IIB-2 to be resolved.
9
(
II B-3 BEHAVIOR OF BUR MARK III CONTAltlMEtlT_
Prob 1cm Definition:
The BWR Mark III containment differs in many respects from the Mark I and II designs.
Various aspects such as vent cleari.'q, y
- u coolant interaction, pool swell, pooi c:
fication, pressure loads, and flow bypass must be evaluated and approved; ongoing experimental tests should develop much of the necessary data to confirm the conservat:
in design.
The areas of ACRS concern identified above Prograr::
have been included in our safety review for the Mark III containment. We believe that the information which has been developed to date is sufficient to demon-strate the adequacy of the Mark II!
containment design. The acceptant:
criteria we have established wil' rovide adequate safety margins for the III containment. We consider the rei.::,ning Mark III testing to be e 1firmatory in nature and will require..iat these pro-grams be completed p-ior to issuance of an operating license for a Mark III plant.
We are currently preparing a report to th" ACRS to document our position on the Mark III containmcnt design.
This generic item has been considered during the development of the staff's technical activities program.
It is included in the scope of Task No. B-10.
There is no change in this status since the last report.
II B-4 STRESS CORROSION CRACXIt!G IN BHR PIPING Problem Definition:
Cracking in austenitic stai'less steel piping has occurred in seve i BWR reactors in recent years. Some of a:e piping con-tained throughwall cracks which leaked coolant.
Program:
The instances of stainless steel pipe cracking in BUR's have been extensively investigated by the NRC Pipe Cracking Study Group and reported in NUREG 75/067
" Technical Report, Investigation and Evalua-tion of Cracking in Austenitic Stainless Steel Piping of Boiling Water Reactor Plants". The NRC investigation concluded that the cracking was the result of inter-granular stress corrosion cracking.
The study gtoup further concluded that crack-ing of this nature would not present a significant safety hazard provided measures were taken to detect small cracks and that piping in which leaking occurred was re-paired or replaced.
They recommer:'ed that alternate materials having a high
.istanc:
to intergranular stress corrosion
- cking be considered for use in the reace r coolant pressure boundary.
Based on these recommendations, the staff set forth its technical position in the
" Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping" NUREG-0313, pub-lished in July 1977.
Owners of operating BWR plants and the owners of BWR plants that are in the licens-ing review sequence have been requested to provide the details and timing of the imple-mentation of NUREG-0313.
Following recently reported discovery of cracks in larger piping, the Pipe Crack Study Group was reconvened on September 14, 1978.
The Study Group will specifically address the following issues:
sm =
g II B-4 STRESS CORROSION CRACKING IN BWR PIPING (a)
The significance of the cracks discovered in large diameter pipes relative to the conclusiors and recommendations set forth in NUREG 76/067 and its implementation docu-ment NUREG-0313.
(b)
Resolution of concerns raised over the ability to use ultrasonic techniques to detect cracks in austenitic stain-less steels.
(c)
The significance of the cracks found in large diameter sensitized safe ends, and any recommendations regarding the current NRC program for dealing with this matter.
(d)
The potential for stress corrosion cracking in PWRs.
(e)
The significance of safe end cracking at Duane Arnold relative to similar material and design aspects at other facilities.
The study group is.;heduled to
- alete its evaluation and report in January 979.
In addition to the study group effort, the NRC has underway several other generic technical efforts.
Regulatory Guide 1.44, " Control of the Use of Sensitized Stainless Steel", was pre-viously issued to address related problems in this area. The Guide is being revised as necessary to incorporate the improved processes and materials resulting from the current work.
NRC is funding the Primary Coolant Pipe Rupture Study at GE. A promising electro-chemical test for stress corrosion suscepti-bility has been developed under this program.
The test is now being evaluated via a "round robin" test series amongst several e
p II B-4 STRESS CORROSION CRACKING IN BWR P' PING, laboratories.
Another NRC activity includes work at American University on inhibitors.
EPRI and GE are funding more than $8 million for programs to qualify and to establish proper material and fabrication specifica-tions for alternative materials to the currently used 304 stainless steel.
NRC has a member on the EPRI corrosion advisory committee.
In view of the continuing problems with pipe cracking in BWRs, this issue has recently r been designated as Task A-42, " Pipe Cracks at Boiling Water Reactors", by the Technical Activities Steering Committee.
Future pro-gram effort will be carried out under ask A-42.
,_gt c ~~ np ? ?
A O
4 -
IIC-1 LOCKillG OUT OF ECCS POWER-0PERATED VALVES Problem Definition:
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 opened 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 con'-
sidering 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 circuit breaker is open; (4) can the valve be locked out in an improper position due to a faulty indicator; (5) are there other designs improving reliability without lock-out; (6) what are the advantages and disadvantages of corrective action by an alert operator in case of incorrect posi-tioning vis-a-vis a system with power locked out.
Program:
This generic item has been considered during the development of the staff's technical activities program.
It is included in the
. scope of Task tio_. B-8.
A Task Action Plan for this activii.y is currently under development.
s-c N " "
7 m
~<
IIC-2 DESIGN FEATURES TO CONTROL SAB0TAGE Problem Definition:
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 introduce design measures that could protect against sabotage or mitigate the consequences thereof.
Program:
NRC has established a Task Action Plan (TAP A29) to identify and evaluate alternative design concepts which could reduce nuclear power plant vulnerability to industrial sabotage.
Present power reactor designs depend on their inherent protective features and extensive physical security measures to provide the high level of protection recuired by 10 CFR 73.55.
An alternative approach for 1chieving this level of protection would be to further reduce the reactor vulnerability by considering vulnerabilities to sabotage at the conceotual design sta:e. Task Action Plan A-29 is an effort to.2entify and evaluate, ' y an ex-o tensive research program, candidate desian alternatives which would provide such an increase in the inherent sabotage protection of the plant.
Sandia Laboratories has been selected for this research contract and the initial phase of the program, i.e., a standard plant characterization is in progress.
An inter-office working group has met to " iN existing studies and to provide input to research program.
ceport of the workin group's activities. being prepared by i.
Candidate design concepts identified by the group, to be examined for feasibility cost, and effectiveness by Sandia, include:
1.
Plant Layout variations to enhance the protection of vital systems.
The concepts of additional separation of vital systems, as well as co-location of vital components will be examined.
2.
Vital equipment and systems status monitoring to permit the operator to have sufficient knowledge of vital systems integrity to counteract the sabotage attempts with other systems and components.
3.
Damage control concepts, including the potential for rigging portable emergency power and cooling water equipment (kept at another location) to plant systems.
4.
" Hardening" a minimum complement of vital safety systems against external attack and/or insider tampering.
5.
Alternate containment designs which could assure integrity of the containment during all faulted conditions of primary coolant system and reactor core, including a "mel t-down".
6.
Alternate shutdown and decay heat removal concepts, including a completely separate, hardened decay heat removal system.
Additional suggestions and variations of the above-mentioned concepts will be examined as they emerge frcm a detailed examination of these alternatives.
e d
d o
4 II C-3A DECONTAMINATION OF REACTORS Problem Definition:
Experience should be gained in decontamina-tion of actor coolant systems so that such informat;en is available when needed.
Program:
Some reactor system decontamination may be necessary in conjunction with the decommission-ing of power reactors. However, application of decontamination techniques in such instances would not require extensive NRC review since materials compatibility considera-tions normally associated with reactor operation subsequent to decontamination would be minimal.
The decontamination of operating nuclear power plants, on the other hand, requires extensive review to assure that the decontamination method will not adversely effect the integrity of the primary coolant system boundary during the decontamination process and during sub-sequent operation of the facility.
The major purpose of' coolant system decontamina-tion of operating reactors is to reduce radiation levels in order to permit personnel access for inservice inspection, repairs, and modifications that are required to assure plant safety.
To date there has been little experience with primary system aecontamination of operating U.S. commercial power reactors.
The Hanford N reactor primary coolant system has been periodically decontaminated.
In Canada, successful decontaminations have been accom-plished at Gentilly I and Douglas Point.
Also contractors to the Division of Naval Reactors, ERDA, have reported the decontamina-tion of many reactor systems under their cognizance.
More recently the Electric Power Research Institute (EPRI) has initiated research programs on decontamination of operating power reactors and Commonwealth Edison is conducting an extensive test pro-gram in preparation for decontamination of the Dresden Unit 1 primary system.
The staff is evaluating the results of the test program and the effects of the proposed decontamina-tion on system integrity prior to our approval.
The staff is also following EPRI programs. -
(
II C-3 A DECONTAMIrlATION OF REACTORS (C0tlT'D)
This generic item has been considered during the development of the staff's technical activities program.
It is included in the scope of Task Action Plan A-15.
D 4
(
IIC-3B DECOMMISSIONING 0F REACTORS Problem Definition:
Specific plans should be developed, including definitive codes and standards, covering plant decommissioning.
Program:
Access control for radiation areas, exposure control, concentration limits for release of radioactive material, personnel monitoring requirements and radiation survey are among the requirements specified in ln CFR Parts 20 and 50 for all phases of rea or operation, including decommissioning.
Rec ;atory Guide 1.86, published in June of 1974, was developed to provide specific guidance on reactor de-commissioning and includes a discussion of the steps required to assure adequate de-contamination prior to termination of a reactor license.
The Atomic Industrial Forum and Battelle Northwest are now engaged in studies of reactor decommissioning alternatives, including pro-tective storage or mothballing, entombment, dismantlement and combinations of these al-ternatives.
Both studies will evaluate safety, environmental aspects and costs of each de-commissioning alternative. The AIF report was published in November 1976.
Battelle completed their report on PWR decommissioning in June 1978.
An addendum is being prepared that relates the costs of and exposures from deconmissioning facilities of various power levels and addresse, in more detail entombment.
This is due to be published in draft form in
- January 1979.
The draf t BWR report is due to be published in April 1979 with the final due in the fall 1979.
In addition, the NRC Office of Standards Development requested proposals for conducting a study to evaluate the dose commitment for radioactive material released to unrestricted use from decoanissioning of nuclear reactor facilities.
A contract has been issued to ORNL with the final report due to OSD in the Spring of 1978.
Experience has been gained in the deconinissioning of Elk River, Hallam, Fermi 1, Saxton, Peach Bottom 1, and numerous smaller test and research reactors This experience is being factored into the AIF and Battelle N.W. reports. The de-commissioning experience and reports will be used as background information in the modification of existing regulations and guides on reactor decommissioning and in the development of any new standards or guides on reactor decommissioning.
Early this year, the Commission,as briefed by the staff an the overall NRC commissioning policy. On Febru 16, 1978, the Commission approved the sta;i program to initiate a generic rulemaking action and the schedule for disposition of the PIRG petition (a request to require advanced financial arrangements for decommissioning). At the 214th and 215th meetings of the ACRS, the staff provided their program for resolution of this item. ACRS comments were received on March 15, 1978.
This generic item has been considered during the development of the staff's technical activities program.
It is included in the scope of Task Number B-64.
It is anticipated this this program will be completed in about two years..
r IIC-4 REACTOR VESSEL SUPPORTS (ASYMMETRIC LOCA LOADS FROM SUDDE.,' 'USCn0 LED BLOWDOWN Problem Definition: On May 7,1975, the NRC was informed by Virginia Electric & Power Company that the potential for an asymmetric loading condition on the reactor vessel supports resulting from a postulated large, sudden rupture of the reactor coolant pipe at the vessel cold leg nozzle had not been considered by Westinghouse or Stone and Webster in the original design of the reactor vessel support system for North Anna, Units 1 and 2.
This loading condition could occur as a result of forces induced on the vessel due to transient differential pressures across the core barrel and by forces on the vessel due to transient differential pressures in the reactor cavity during a postu-lated LOCA. These differential pressures, although of short duratinn, could place a sig-nif; cant load on the reattor vessel supports; ti. cfore the integrity cf the vessel supports was called into question. In addition, the dif ferential pressures itduced on the reactor internals by transient differential pressures across the core barrel, tccurring during sub-cooled blowdown for LOCA"s resulting from pipe breaks could result in sign 'icant stresses on fuel assemblies, which c:uld affect the ability to maintain a coolable c:re geometry.
Upon closer examination tf this situation, it was determined that post: lated breaks in a reactor coolant pipe at xessel nozzles were not the only area of concern but rather that other pipe breaks in the react:r coolant system could cause internal and extenal transient loads to act upon the reactor vessel and other components.
For the postulated pipe treak in the cold leg, asymmetric pressure chan!es could take place in the annulus between the lore barrel and the vessel.
Decompression cnJld occur on the side of the vessel annulus netrest the pipe break before the pressure on tha opposite side of the vessel changes.
This monentary differential pressure across the core barrel could induce IIC-4 REACTOR VESSEL SUPPORTS (ASYMMETRIC LdCA LOADS FROM SUDDEN SUBC00 LED BLOWDOWN CONT'DJ lateral loads both on the core barrel and on the reactor vessel. Vertical loads could also be applied to the core internals and to the vessel due to the vertical flow resistance through the core and asymmetric axial decom-pression of the vessel.
Simultaneously, for vessel nozzle breaks, the annulus between the reactor and biological shield wall cc. " become asymmetrically pressurized resulting a
differential pressure across the vesst. causing additional horizont?1 and vertical ex ernal loads on the vesse-In addition, the vessel 9 effects of initial ten-could be loaded b) sion release and :
.down thrust at the pipe break. These.loacs could occur simultaneously.
For a reactor vessei outlet break, the same type of loading could occur, but the internal loads would be predominantly vertical due to more rapid decompression of the upper plenum.
Failure of the reactor vessel supports would result in vessel movement which could (1) cause consequent failure of ECCS lines connected to the coolant loops, (2) cause the failure of control rods to function properly, and (3) cause loss of integrity of other reactor coolant system components ' pump and steam generator supports).
Program:
Following disclosure of this problem during the OL review of North Anna Units 1 and 2, a survey of all operating PWR reactors was conducted in October 1975.
That survey showed that none of the above described transier.t differential pressures had been considered in the design of the reactor vessel supports for any operating PWR facility.
In June 1976, we requested licensees of all operating PWR facilities to assess the adequacy of the reactor vessel supports at their facil-ities with respect to sudden subcooled blowdown loads. Most licensees have formed " owners groups" for the purpose of taking concerted action.
Most licensees with Westinghouse plants proposed an augmented inservice inspection program (ISI) of the reacter vessel safe-end-to-end pipe welds.
~
IIC-4 REACTOR VESSEL SUPPORTS (ASYMMETRIC LOCA LOADS FROM SUDDEN SUBC00 LED BLOWDOWN CONT'D) in lieu of providing the detailed analysis we requested.
Licensees with Combustion Engineering plants submitted a probability study rrepared by Science Applications, Inc.
in support of a conolusion that a break at this location has such a low probability of occurrence that no further antlysis is neces-sary.
B&W licensees engaged Science Applications, Inc. for a similar study which is currently under staff teview.
When the W and CE owners group reports were received Tn September 1976, DOR formed a review Task Group consisting of members from DOR, DSS and EDO to evaluate these alternate proposals.
In addition, EGSG Idaho, Inc. was contracted to perform an independent review of the submitted probability study.
The staff review of the proposed alternates has been completed.
The Task Group and EG?G inde-pendently reached the same conclusion:
chat the alternate proposals set forth in. cse reports should not be accepted in lieu of the requested analyses.
The basis is that a suffic-ient data base does not exist within the nuclear industry to provide satisfactory answers to
- 43ny information needs identified by the staff.
The staff has met with the CE Owners Group and will meet with the h[ Owners Group to explain why the probability study /ISI reports could not be accepted. We will also provide them all the questions that have been generated to date as a result of our review of the h[ and CE topical reports.
In addition, letters were sent (Jan.
1978) to each PWR licensee and applicant inform-ing them that an analysis m e: be undertaken to assess the design adequacy the reactor vessel supports and other structur:: at their facilities to withstand the postulated asymmetric LOCA loads.
The application of this problem to BWR facilities has been investigated and a letter will be sent to each BWR licensee and applicant requesting that they assess the potential damage to their primary coolant system from these loads..
f 1
IIC-4 REACTORVESSELSUPPORTS(ASYMMETRICi.0CALOADSFROMSUDDENSUBC00 LED BL6WDOWil C0lT'D)
This generic item has been considered during the development of the staff's technical activities program.
It is included in the scope of Task Action Plan A-2.
t II C-5 RATER RAMMER Problem Definition:
Several cases of water slugging or water hammers have occurred in both PWR's and BWR's.' Corrective actions should be taken to minimize such events.
Program:
The generic consideration of water hammers was incorporated in Task Action Plan A-1,
" Water Hammers." Work on the task is proceeding as described below.
Under Task 1.1 a report of a review of water hammer in nuclear power plants was completed and is currently in the final review stage prior to publication.
Under Task 4.1 work has started at INEL on a review and evaluation of specific water hammer problems identified in the Task 1.1 review. A draft of the final report on this work is scheduled to be issued in February 1979.
Under Task 4.2 work has started on a state-of-the-art review of analytical, experimental and design work pertinent to water hammers in nuclear power plants. A draft of the final report on this work is scheduled to be issued in May 1979.
Under Task 4.3, work is continuing at BUL on potential water hammer problems in preheater type steam generators. The final report is scheduled to be issued in May 1979.
Under Task 4.4 work is continuing at INEL on preparation of calculational methods to be used in the analysis of hydraulic and structural consequences of water hammers in operating plants. Reports on this work are scheduled to be issued in February and May, 1979.
IIC-6 MAltlTErlANCE AfID IllSPECTION OF PLAtlTS Problem Definition:
Provisions should be included in the design of future plants which anticipate the ain-tenance, inspection and operational r 'ds of the plant throughout its service life.
Program:
The Staff and industry have a multifacetted' program which will lead to improved mainte-nance, more reliable inservice inspection.
and a better meeting of the operational needs of the plant throughout its service lif including deccntamination and decommis aning.
The staff has issued Revisir-2 to Regu:atory Guide 1.70, which requires c..alicants to sub-mit details of design features and operating procedures to assure that radiation exposures to plant personnel during operations and maintenance will be as Icw as is reasonably achievable.
The staff will assure that such features are implemented by follow.ifg our Standard Review Plan during our licensed rev'.w activities.
Regulatory Guide 8.8, Revision 3, which was issued in June 1978, provides the basis by which our staff reviewers can assure that radiation exposure during plant maintenance will be as low as is reasonably achievable.
This revised guide was issued in March 1977 for c; rent and is used by the staff in review of at: icants' radiation protection programs.
iff presently has a research cont:
- . c r :uled for completion in 1979 with UNI te
, vest:aate the cost impact and dose reduction 2ffectiveness of the features of Regulatory Guide 8.8. The staff has a technical assistance contract with SAI to determine the radiation exposures associated with safety related activities, such as inservice i
inspection, at operating nuclear power plants.
The industry (AIF and EPRI) has embarked on 2 long term (three to five years) program to investi a.e means to reduce radiation levels and radiatic exposures to personnel during operational anc maintenance activities at nuclear power plants.
The generic item has been considered durino the development of the staff's technical activities program.
It is included in t he scope of task Number B-34
>.T t --
((
(
11C-7 BEHAVIOR OF BWR MARK I CONTAINMENTS Problem Definition:
In' the course of performing large scale testing of an advanced design for pressure-suppression type containments and during inplant testing of Mark I containments, n suppression pool hydrodynamic loads were identified which had not 'xplicitly been included in the original rk I containment design basis.
These new aads_ result from dynamic effects of drywe t i air-and steam being forced into the suceression pool (torus) during a postulated loss of coolant accident (LOCA) and suppression pool response to safety-relief valve operation generally associated with plant transient operating conditions.
Program:
In April 1975 the NRC requested all utilities owning BWR facilities with the Mark I con-tainment to review their plant designs to determine whether the new load information would affect the structural adequacy of their containments.
As a result of the NRC inquiry, all, affected utilities formed a group known as the Mark I Owners Group.
The cbjectives of this group were to determine the magni-tude and significance of these dynamic loads as quickly as possible and to identify courses of action needed to resolve any out-standing concerns. Accordingly, the Mark I Owner.s Group proposed to divide this task into.two programs: a short-term program (STP) to be completed in early 1977 and a long-term program (LTP) presently scheduled for completion in 1979.
The objectives of the STP were (1) to examine the containment system of each operat-ing BWR with the Mark I containment design to verify that it would maintain its integrity and functional capability when subjected to the most probable loads induced by a postu-lated design basis LOCA: and (2) to verify that licensed Mark I BWR facilities may continue to operate safety, without undue risk to the health and safety of the public, while a methodical, comprehensive Long Term Program (LTP) is conducted.
It was determined that, for the STP, " maintenance
(
t of containment integrity and function" would be adequately assured if a safety factor to failure of at least two were demonstrated to exist for the weakest structural or mechnaical component in the Mark I containment system.
The objectives of the LTP are (1) to establish design basis (conservative) loads that ar3 apprdpriate for the anticipated life (40 years) of each
. Mark I BWR facility, and (2) to restore the original intendea design safety margins for each Mark I containment system.
The STP centered around two areas of investigation:
(1) an evaluation of the loads on structures within the torus, and (2) en evaluation of the loads on the torus structure and its supports. The loads on the structures within the torus were based primarily on the data developed from the Mark III containment tests conducted at the General Electric Pressure Suppression Test Facility.
The loads on the torus structure and its supports were based on a series of tests performed on a one-twelfth scale model representing a segment of a Mark I contain-ment torus.
During the STP review, whenever structural safety margins were found to be less than a factor of two at an operating Mark I BWR facility, the safety margins were required to be increased.
One of the methods used to accomplish this has been the use of drywel1 to torus differential pressure control.
In addition, several utilities have iuntarily performed modifications to their
,s support system to provide additi<..;i design safety margin.
The NRC staff has completed its review of the generic Mark I Containment STP conducted by the Mark I Owners Group a-the associated plant-unique information pro. jed by the licensees of operating Mark i dWR facilities and has concluded that licensed Mark I BWR facilities can continue to operate safely, without undue risk to the health and safety
. r.
Ju~.
y
'of the public, during an interim period of approximatei/ two years while a methodical, comprehensive LTP evaluation is conducted.
The results of this review are documented in the staff's " Mark I Containment Short Term Program Safety Evaluation Report", NUREG-0408,
, December 1977.
The LTP, which includes detailed testing and analytical work, commenced in June 1976 and is scheduled for completion in 1979.
Re-vision 3 of the Mark I Containment Program Action Plan (PAP),v,Sich was submitted to the NRC on Febr;ary 15, 1978 by General Electric Company on the behalf of the Mar, Owners Group, describes the objectives of each of the analytical and testing programs being conducted by the Mark I Owners Group in the LTP. The PAP includes both activities to obtain improved load definition for the existing system design and activities to identify and qualify potential load mitigat-ing devices and/or procedures.
The NRC staff has reviewe'd the Mark I Owner's LTP PAP and has found that it is reasonably designed to provide resolution of issues raised during our review of the STP and to satisfy the LTP objectives.
NRC staff will continue to follow the progress of the LTP to assure that appropriate actions are taken in a timely manner.
The NRC has sponsored a testing program at Lawrence Livermore Laboratory (LLL) utiliz-ing a three-dimensional, 1/5th scale, 90 segment of a typical Mark I containment torus.
A final report of the LLL test re-sults was issued in March 1978.
LLL is performing additional analyses of the test results to quantify loading trends.
These analyses are expected to be completed by January 1979.
This program will provide the NRC staff independent confirmatory load information.
I b+
This generic item has been considered during the development of the staff's technical activities program.
It is included in the scope of Task Action Plans A-6 and A-7.
0 9
x t
IID-1)(SAFETYRELATEDINTERFACESBE71JEE i
i Questions have been raised concerning both l
Problem Definition:
standardized balance-of-plant and nuclear l
steam supply systems on the one hand and i
custom-designed site-related structuresThe depth and components on the other hand.
of detail required at the stage of preliminary Design Approval may not be adequcte for con-Proc struction approval.
quality assurance programs covering design, precurement, construction, and startu system analyses to assure functional compati-bility across the interfaces, as well as for other systems, are necessary to assure furtional is compatibility for the postulated design bc accident conditio.s.
The staff has identified the safety related
'rogram:
interfaces of licensing concern for the NSSS and B0P as they relate to each other and The staff to the site and utility owner.
has also developed a format for presentatior, The interfaces and of int'.cfaces in SSARs.
ce presented in NUREG-0102, " Interfaces format Jard Designs", August 1976 which for S-has rc.gived extensive staff review and was made etailable in November 1976 to interested The comments members of the public for comment.the staff's satisfaction.
now have been resolved t'~
A new system for the Stardard Review Plan has been prepared that incorporates NUREG-0102 by reference and provides guidance to the staff for performing the review of interfaces forIt h SSARs.
will be reissued as an appendix to R.G. 1.70 Rev. 2 (Standard Format and Content guide).
The RRRC has recently approved the revised content of NUREG-0102 for publication.
The staff is presently implementing the require-ments for ;aterfaces, as identified in NUREG 0102.
Upon final documentation in its review of SSARs.
as an Appendix to R.G. 1.70, the staff considers Issuance that this issue will have been resolved.
of R.G. 1.70 (Rev. 2) including the new appendix It is not is s.cheduled for December 1978.
included as a part of the technical activities program.
(
- s.
II D-1B SYSTEMS INTERACTIONS IN NUCLEAR POWER PLANTS Problem Definition:
A need has been identified to confinn whether the present methods and procedures for plant design, analyses, and safety re-views acceptably account for potentially adverse systems interact'ons. The problem to be resolved is to datermine whether pre-sent methods and procedures are adequate or whether changes should be made to assure that potentially adverse interactions are accounted for in the design, analysis, and-safety reviews.
Program:
This matter has been incorporated as a part of the staff's technical activities program.
It is included in the scope of Task Action Plan A-17.
The Task Action was approvea on November 15, f
1977 and has been revised to e molish the 0
task with contract assistance.
lis is made Y
b ecessary because of the worLe Ja impact on
@,/{
t chnical branches that ac r:tical to p
ac gmplishment of the ta:
A comoined
./
e, effoqt i volving contract
- sistance, techni-fl[ [y'
[,c-Reactoi * :gulation, and personnel from the cal p'er.
nel from the Ofrice of Nuclear Office of, Standards Development will perform p[/
the tash.\\ The contract assistance gro a will h
4 will devel6p an independent methodology for 4
conducting a\\ review for systems interaction b,6 '
7 and will asse's the Standard Review Plan s
against this met,hodology to identify wheth-any changes are necessary.
The contract assistance group will be guided and assir-by NRR personnel. The perfonnance will evaluated by a selected group within NFr and will be reviewed by,all cognizant
- nni-cal branches within NRR.
The contract effort was initiated in. ay 1978.
The first phase of the task will be com-pleted within sixteen months and will identify any corrective procedures.
Thessecond phase will be completed in an addition'al twelve months and will implement the cor'rsective procedures.
II D-2 ASSURArtCE OF C0flTINUOUS L0ftG-TER'4 CAPABILITY OF HERMETIC SEALS Ort IllSTRUMErlTATI0tl AND ELECTRICAL EOUIPMEtiT Problem Definition:
Certain classes of instrumentation incorporate hermetic seals. When safety related components within containment must function during post LOCA accident condi-tions, their operability is sensitive to the ingress of steam or water.
If the hermetic seals should become defective '=
a result of personnel errors in the mair-tenance of such equipment, such errors could lead to the loss of effective hermetic seals and the resultant loss of equipment operability.
Program:
This generic item has been considered during the development of the staff's technical activities program.
It is included in ti.-
scope of Task tiumber C-1.
We have estab-lished a plan of action which is pending management approval.
The plan includes a schedule for accomplishing the needed inves-tigation into:
a.
field e perience, b.
adequacy of current designs and quality assurance practices, c.
practicability of testable designs, and d.
the need for the development of guidance criteria.
There has been no change in this status since the last report.
.I,
II E-1 S0IL-STRUCTURE INTERACT!0N Problem Definition:
Ongoing studies by the NRC and the industry are reviewing and re-evaluating matters related to soil-strrf.ure interaction and to the appropriate sei ric response spectrum to be used at the founaacion level of a nuclear power plant.
These reviews may lead to a modification of current criteria used in the seismic design of foundation structures.
Program:
This subject has been included in the scope of Task Actinn plan A-40, " Seismic Design Criteria - Short Term Program".
The subject of soil-structure inte - ction analysis is controversial because '- the various techniques preferred by dif#erent applicants.
In order to provide a confirmc-tory basi.s for, or otherwise to rNise the current Standard Review Plan positions, an indepth study which will evaluate from an analytical point of view the various tech-niques, including deconvolution analyses, is being perfomed.
The object:.'e of this investigation is to determine limits and conditions of applica-bility as well as estimates of conservatism in the definition of seismic input and soil-structure interaction procedures currently used in the seismic analysis of nuclear power plants.
Specific attention will be given to the conservatism embodied in the aplication of computer programs such as
..AKE and LUSH employed for deconvolution end soil-structure interaction analysis.
Particular attention will be given to re-quirements concerning variation of soil properties, envelopirg the response spectra at the foundation lewl, and fixing a minimum value of the response spectra at the foundation level.
It is anticipated' thzt work on soil-structure interactics will be completed dur-ing 1978, and a report will be issued during the first quarter of 1979.
Completion of this task then will require incorporation of the results into a Regulatory Guide or the Standard Review Flan.