ML19224D548
| ML19224D548 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/03/1979 |
| From: | Taylor J BABCOCK & WILCOX CO. |
| To: | Ross D Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7907120589 | |
| Download: ML19224D548 (24) | |
Text
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Babcock &Wilcox ec.e, cece,,oco c,cm, P.O. Box 1260, Lynchbt,g Va. 24505 Te'cchcr'e: (804) 3844111 July 3, 1979 Dr. D. F. Ross, Jr.
Deputy Director Division of Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555 References :
1.
Letter, Ross to Taylor, " ACRS Recon rendations Relating to TMI-2 Accident."
2.
Letter, Heltemes to Taylor, " Additional ACRS Recommenda-tions Relating to TMI-2 Accident."
Dear Dr. Ross:
In the references, the NRC requested "B&W provide the staff with a concise discussion and position on each of the ACRS recommendations relating to TMI-2."
Attached to this letter is our response to your request. Most of these responses were telecopied to Mr. Heltemes on May 30, 1979.
If you have any questions or desire any acditional information, please contact either myself or Mr. E. R. Kane of my staff.
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/ James H. Taylor Manager, Licensing JHT:dsf Encl.
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Letter, M. Carbon to Chairman Hendrie, Dated Aoril 7. "'79 Recomendation 1 - Parform further analysis of 3 mall break transients and accidents.
BW POSITION As a result of the TMI-2 accident, the staff requested a number of additional analyses. Many of these analyses were completed and provided in our sub-mittal, "Evaluationof Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant".
This submittal is applicable to 177 FA raised loop and lowered loop plants. A partial list of investiga-tions that B&W has underway or completed is given in a letter frcm J. H. Taylor to R. J. Mattson, " Babcock & Wilcox Commitments", dated April 30, 1979.
We believe these analyses adequately answer the safety concerns in the area of small breaks for B&W plants at this time and are responsive to the recommendations in this area. However, we are continuing our analysis of the TMI-2 event to inve tigate conditions beyond 90 minutes.
We also plan to continue to participate in the small break standard problem prog ram.
Recommendation 2 - Provide operator additional information and means to follow the course of an accident; as a minimum, consider expeditiously:
1.
Unambiguous RV level indication 2.
Remotely controlled vent for RCS high point.
B&W POSITION B&W actions since the TMI-2 incident have addressed five separate items relative to this ACRS recommendation:
1.
Reactor Vessel Level 2.
Wide Range T Indication g
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Saturation Condition Moni tor 4
C6re Exit Thermoccuples
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Rcrote Operated Vents
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The B&W position on each item will be summarized secarately, but since all of these items relate to Reactor Coolant System conditions, the pros and cons of each item should be evaluated in the context of information available to the operator and other changes being considered or implemented.
B&W cautions against providing the operator additional non-essential information that could lead to confusion.
Our position at this time is that changes are justified with regard to items 2, 3 and 4. With regard to reactor vessel level and remote operated vents further investigations are needed. B&W also believes that further consideration may be given to additional operator information, that the plant computer be used togreater advantage.
(1) Reactor Vessel Level B&W has been reviewing the merits and methods to measure reactor i
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vessel water level. This effort is comprised of three parallel activities:
(a)
Detennining whether such monitoring is essential to operator actions related to mitigating a small or large LOCA.
(b) Determining whether such monitoring is a desirable input to the operator'sassessmentofplantconditions,consideringtgtthe operator has other information available and considering that there may be negative aspects to such a change.
(c)
Assessing the " state-of-the -art" methods for monitoring reactor vessel level.
As a result of our study to date, we have concluded that addition of a reactor vessel coolant level measurement to provide a direct indication of primary coolant inventory is not required fr use by the operator to safely J.b.'_)
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I Instrumentation to Diagnose and Follow an Accident (Cont'd) mitigate a loss-of-coolant accident and to subsequently place the plant in a stable long term cooling mode.
Other plant parameters along with specific operating procedures and training adequately provide for core protection without the requirement of direct inventory level measurement.
LOCA mitigation is dependent upon the ECCS (HPI, LPI, & passive CFT Systems) to supply adequate cooling water to prevent or limit a thermal excursion in the reactor core. The pumped injection systems are automatically ini-tiated by the ESFAS on low RC pressure or high RB pressure, and the operators have been trained to keep those systems once initiated, in service until proper coolant conditions are regained in the reactor coolant system or until alternate modes of decay heat removal are establishe' Reactor coolant pressure and temperature, show'ng subcooling, are the principal indicators for use Ly thc operosors to establish the need for maximum emergency coolant addition or for securing and/or regulation of the ECCS.
For a small break, B&W has recommended that the operator insure maximum ECCS coolant addition until the reactor coolant is at least 50F subcooled.
The subcooled condition insure that an adequate energy transport mechanism exists to remove decay heat from the core with or without forced reactor coolant flow.
For large breaks the reactor coolant system will remain saturated for long periods of time, and thus ECCS operation must be retained until the LPI system is fully operative for long term cooling.
Detailed information about small break p*ocedures, operator training, the use of existing instrumentation, and plant performance have been provided in Appendix 4 of B&W's submittal to the NRC on May 7,1979. This infomation has been used by our utility customers in developing 'neir detailed small break operating procedures. While not considered necassary, B&W considers
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Instrumentation to Diacnose and Follow an Accident (Cont'd) that reactor vessel level or perhaps reactor coolant system inventory may be desirabl' as a confirmatory parameter. B&W has, therefore, begun investigations to determine availability of a reliable and accurate method to monitor reactor coolant level for the B&W NSS.
In determining availability of such instruments, we will be relying heavily on the develop-ment work previously performed by others such as EG&G in Idaho, and will be meeting with them shortly for this purpose.
Folicwing our review and selection of available methods, we will evaluate feasibility and means of adapting these instruments to the B&W NSS. To date, our investigations have identified the following candidate methods:
(1) Delata-P (2) Ultra-Sonic Probe (3) Conductivity Probe (4)
Impedance Probe j
(5) Transmission Line Reflectometers t
In sumary, we conclude that reactor vessel level measurement is not required for managing RC breaks over the entire spectrum.
However, since the additional confirmatory reading may be desirable, we are continuing our investigation of possible methods and investigation of their adaptability to the B&W NSS.
(2) Wide Range T Indication H
B&W has recommended extendig the range of the reactor coolant hot leg temperature readouts and these changes have been implemented on some plants already.
Prior to the change, the T indicator covered the range from H
520 F to 620 F.
It now has been extended to cover the range from 120 F to 920 F.
This will assist the operator in monitoring the approach to saturation cver a wider range of conditions.
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(3) Saturated Condition Monitor To date, B&W has taken the following steps on this item:
(A) A curve reflecting 50 subcooling over the pressure range has been provided to the operating utilities with the reccanendation that it be posted in the control room if equivalent methods of providing this information to the operators are not already available.
In conjunction with this; recommendations, design requirements, and equipment have been sent to the operating utilities to expand the range of the hot leg temperature indications.
(B) A recommendation and associated details for including subcooling information and associated alanns in the plant computer have been forwarded to the operating plant utilities.
(C) Development of a system to directly indicate subcooled margin is currently underway at B&W.
B&W is continuing discussions with the operating plant customers to detemine the best manner to satisfy operator needs.
(4) Core Exit Tmermoccuoles All B&W plant have incore thermocouples which can be used to measure core outlet fluid temperature. B&W nas recommended to each of its operating utility customers that a minimum number of thermocouples be in service during reactor operation to provide further confirmatory information to the operator during or after a transient.
B&W has also U
recomnended that the range be extended to 2300 F.
(5) Remote Operated Vents B&W's emphasis continues to be on taking steps to prevent the collecticn of non-condensibles in the high points of the reactor coolant system.
However, B&W has performed a preliminary review of the merits of adding
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Ins'.rumentation To Diagnose and Follow an Accident (Cont'd) remotely controlled vent valves at the top of each hot leg.
The primary elements of this review included an evaluation of the following:
(1) Determination of whether additional RCS venting capability is essential for mitigation of small or large LOCA's, (2) Determination of whether additional RCS venting capability is a desirable feature for mitigation of small or large LOCA's, (3) A careful assessment of the less desirable features of designing, installing and using high point vent valves for LOCA mitigation.
B&W has concluded that the addition of remotely controlled vent valves at the top of each hot leg pipe may be a desirable feate e to vent non-condensibles and to aid in depressurization of the reactor coolant system during smal1 break scenarios. However, our assessment indicates that the remotely controlled vent valves are not required to safely miti-gate a loss of coolant accident and place the plant in a stable long term cooling mode.
The benefits of installing remotely controllable vent valves at the highest point in the loop for LOCA mitigation include:
(1) Venting of non-condensibles is possible and would be beneficial for RCS natura.1 circulation.
(2) Any steam produced in the core could be directly vented to the reactor building resulting in a higher heat removal rate.
(3) RCS refilling could be more rapidly achieved and a stable long term cooling made reached earlier.
Our initial assessment is that further evaluation is warranted.
This evaluation should include a detailed analysis to confirm the benefits and define design requirements as well as to understand the potential negative impacts of implementing such a change.
Areas to be considered include:
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(1)
Provisions to assure the vent; cannot inadvertently open or fail to close, thereby initiating a small LOCA.
(2)
Proper sizing of the vents.
(3)
Seismic and LOCA loading desing of a restraint system must be examined.
The high point in the RCS f oop is a region of high seismic acceleration requiring careful selecti;n of valves, operators and supporting equipment.
(4) The valves which would t e used would have to be qualified for operation in a LOCA environment f or either water or steam discharge.
In summary, B&W has concluded that RCS high point venting, while not mandatory, may be desirabla for LOCA mitigation, and merits further investi-gation. However, the neg Ative aspects of implementing such a design change must be carefully evaluated.
Recommendation 3 - Item 4b of Bulletin 79-05A considered unduly prescriptive in view of uncertaintias in predicting course of anomalous small break transients / accidents.
B&W Position B&W has provided revised recommendations in our operating guidelines for small breaks.
Re u mendation 5 - Consideration should be given to additional monitoring of ESF equipment status, and to supporting services, to help assure avail-ability at all times.
B&W Position B&W believes that assurance of system availability is best accomplished by a combinationof periodic surveillance, testing and status monitoring. We believe that additional methods of monitoring the status of the actual ESF devices and their supporting services should be further investigated to determine if such monitoring would increase system availability. The additional monitoring could include such items as the normal status of valves in safety related fluid streams, availability of power to the ESF devices, and the status of other features which could prevent the system from accomplishing its safety-related function. The methods for accomplishing this additional monitoring should not be limited to safety-grade systems, and use of systems such as the computer should be allowed to accom-plish this function.
Plants presently have meens of determining whether the ESF systems are functioning once a trip occurs. There are indicators on the ESF panel which show whether or not the ESF devices have gone to their safe state after a trip.
There are also low flow alarms for HPI, LPI, and containe'it spray; these alarms are activated once the ESFAS trips in order to tell t:e operators if the systems are functioning properly.
In summary, we conclude that the post-actuation ESF system status indication is presently adequate. We do, however, recommend that improvements to the status indication of ESF systems and components and other systems important to safety be further pursued with special emphasis on verification of overall system availability as opposed to individual ccrponent status.
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i We believe that overall system monitoring when used in combination with periodic surveillance and testing best assures system a!ailability.
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Letter, R. Fraley to Commissioners, dated Aoril 18, 1979 Recommencition 1 - Natural Circulation n
B&W Position In addition to the writeup below, please see our position on thermo-couple indications and saturation temperature indications in response Letter A, Recommendation 2.
B&W has considered natural circulation in relation to each of the areas addressed by ACRS recommendations:
- ANALYSES
- PROCEDURES
- VERIFICATION
- PRESSURIZER HEATER REQUIREMENTS B&W has documented its natural circulation methods, has shown by testing that these methods are conservative and has described means by which the operator can verify that natural circulation is occurring.
B&W has alsc evaluated the role of the pressurizer heaters in the natural circulation made of cooling and concluded that while not essential, the cooldown process would be more orderly if the heaters were available.
A survey of the operating plants indicates that power can be supplied by both offsite and onsite sources.
B&W has documented the results of natur al circulttion analysis methods and the results of natural circulation events at the operating 177 FA NSS's in " Report on Analysis Methods for RCS Natural Circulation,"
May 16, 1979, and Appendix 1 of, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant,"
May 7, 1979. The methods which have been used to predict natural 3.'s c2J$
circulation flow rates under steady-state conditions give conservative re s ul ts. Comparisons with test data show that the results under-predict natural circulation flow rates. The differences between anals.a and test results are attributed to the conservative calculation of steam generator heat transfer. The conservative assumptions tend to reduce the natural circulation driving head provided by the cooled fluid in the steam generator tubes.
Although the analysis results are conservative, they adequately predict the trend of increasing system flowrate with increases in power level.
These methods, in conjunction with plant test data as well as data for unplanned transients involving loss of forced circulation in the RCS, show that plant operation with natural circulation provides a safe mode for core cooling and transfer of decay heat to the secondary system.
The analytical methods documented in the May 16, 1979 report also include a description of a modification to the transient code CADD to incorporate the transition to natural circulation.
B&W has documented and reported in the May 7,1979 submittal eleven (11) natural circulation events at B&W operating re3ctors.
The eleve, events consist of four planned natural circulation tests, four planned loss of offsite power tests, and three unplanned events including loss of offsite power events.
In all cases, the plants were adequately cooled in the natural circulation mode and primary system temperatures decreased to approach the prevailing steam generator secondary side saturation terrpe rature.
In regard to procedures for placing the plant on natural circulation and means for verifying that natura
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' Controlled Transition to Natural Circulation.
This guideline defines the system conditions r.ecessary to establish natural circulation.
l The procedure specifies the initial conditions which must be obtained before relying on natural circulation for decay heat removal. The most important condition is sufficient subcooling to present steam oubbles in the hot legs after natural circulation is established. With forced circulation, the temperature increase across the cure is small (approxi-U mately 1-2 F). With natural circulation, the temperature increase will be 40-45 F for high decay heat generation.
This will cause the hot leg temperatures to increase.
The cold leg temperature. vill be controlled by the OTSG steam pressure, but may rise slightly when initially transferring to natural circulation.
Curves have baen provided to show the operator the amount of subcooling necessary in forced circulation to account for the temperature rise after natural circulation is established, such that hot leg water remains subcooled.
The operatnr is instructed to monitor the incore thermocouples located at core outir.t.
If these thermocouples indicate cons' ant subcooled core outlet temperature, natural circulation is rencving decay heat.
A test to assure that natural circulaticn is working, once established, is provided. A Jackup method consists of demonstrating th&t the primary system tempe atures are coupled to the secondary steam pressure.
Reducing steam pressi.re will reduce cold leg temperature if natural circulation is functionin; properly.
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f i If natural circulation is not confir ed or is lost after being
-establishef the operator is instructed to restore forced circulation by restarting a reactor coolant pump.
If no reactor coolant pump can be rescarted, the operator is instructed to initiate HPI cooling to remove decay heat.
In sumary, B&W has documented its natural circulation methods, has shown by testing that these methoc s are conservative and has described means by which the operator can verify natural circulation is occurring.
In considering the role of pressurizer heaters, the preferred mechanism for maintaining reactor coolant system pressure is with the pressurizer hea ters. This is the normal made of RCS pressure control and provides the operator a stable and flexible method for control of pressure.
In the event that power to the pressurizer heaters is lost, the reactor coolant system temperature can be reduced to maintain subcooling until such time as the offsite or ensite power sources are regained.
Even if no action to cool the RCS is taken, the cooldown rate of the pressurizer is low enough to allow ample time to regain pressurizer heaters or begin RCS cooldown to maintai-cubcooling. The ultimate backup to both is use of the makeup or higt. pressure injection system to maintain RCS pressure.
The availability of several alternate mechanisms to maintain RCS sub-cooling does not dictate a requirement for qualified onsite pcwer sources to pressurizer heaters to maintain the reacter in a safe cendition.
B&W recommends, however, that when possible a source of onsite power be availaole.
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i Recorrrendation 2 - Thermoccupies used to measure fuel assembly exit temperatures to determine core performance should be used, where currently available, tr guide operator concerning core status (full range capability).
B&W Position See our Position on Letter A, Recomr.endation 2, "Information to Diagnose and Follow an Accident".
Recorrendation 3 - Operating reactors should oe given priority regarding definition and implementation to diagnose and follow the course of a serious accident, including (a) improved sampling procedures under accident conditions, (b) imoroved techniques to provide guidance to offsite authorities.
B&W Position See also our position on Letter A, Recocrendation 2, "Information to Diagnose and Follow an Accident".
(a)
Improved Sampling Procedures -- Sampling techniques used to sample the radioactive coolants of PWR' iroven to be adequate to m
define the chemis t.ry of such coolants d do not, in our opinion, require a" fitional emphasis.
The design and layout of the sampling facilities, however, should be reviewed from the stand-point of reducing man-rem exposure during an abnormal situation.
Sample point accessibility, shielding, piping locations, venti-system, etc., are some of the areas which should be considered.
This would permit samples to be obtained as frequer.tly as necessary without a high man-rem expene'iture.
9ri (b) Emergency Planning -- B&W agrees that the roles of the NRC and other emergency plan participants should be clearly defined.
The NSS supplier and his subcontractors have access to important background information and expertise which are necessary to timely and efficient decision making.
Availability of up-to-date as-built information is essential.
B&W recomends a systematic assessment be made regarding the infor-mation which proved to be important in the TMI-2 incident and that steps be taken to assure the availability of that information to the principal emergency nlant participants.
B&W plans to give further consideration to both the role it should play and the execution of that role in emergencies.
This will be done in conjunction with our utility customers.
Recomendation 3 - Reiterates previous recomendations that high priority be given to "research to improve reactor safety":
(a) research on behavior of LWR's during anomalous transients, (b) NRC to develop capability to simulate a wide range of postulated transient and accident conditions.
B&W Position B&W agrees that higher oriority be given or at least re-prioritizing is in order for reactor safety programs.
B&W believes that emphasis should be placed in three areas:
(1) TMI-2 follow-up studies, (2) Realistic analysis of system responses to probable occurrerces, (3)
Integrated system and separate effects testing.
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TMI-2 Follow-uo Stu e B&W re. commends that maximum information be obtained from studies of the equipment and systems installed inside the TMI-2 containment.
Equipment and instrumentation degradation should be thoroughly studied. There are multiple identical components which should have experienced very similar environmental conditions. Degradation studies of these components should shed valuable light on qualification information ave !1able to date. This should be especially helpful in assessing the ;ignificance of safety grade versus non-safety grade compone s.
The TMI-2 core a sr san.t wil no doubt produce a vealth of infor mation. A e resui: if the TMI-2 incident, L;W is vitally interested in the cond*t1 ;n of 1."e core with particular emphasis on whether the core wo'ild likely have been damaged to any greater extent if the spacer grids had been Zircaloy rather than Inconel.
B&W has already benchmarked successfully one of its major transient computer codes.
It is recommended that other codes, particularly those used by the NRC, be likewise benchmarked against the TMI-2 transient.
2.
Realistic Systems Behavior Assessment TMI-2 has emphasized the importance of studying best estimate responses to probable occurrences.
It is apparent that a closer link must be established between the system analyst and the operator.
Operating procedures based on very conservative estimates of system response may be misleading.
This presents a challenge in two specific areas; identification of those areas where "best estimate" 3E n9 t J a.i,
c c.O codes are not adequate and the definition of likely combinations of transients and operator actions or misoperations.
Studies needed to bound the combinations should be initiated.
From these studies additional insight will be gained as to whether the single failure criterion should be maintained or replaced.
3.
Integrated System and Seoarate Effects Testing Existing facilities (e.g., semi-scale and LOFT) should be investigated for use in or modification to be suitable for investigation of anomalous transients to aid in benchmarking analysis methods.
In particular, data for benchmarking analytical codes and " bubbly" two-phase and reflux natural circulation modes of cooling would be useful.
The discharge flow rates representative of breaks in high pre: are large diameter piping needs further investigation.
Experiments should be perfomed using realistic conditions.
In summary, B&W supports accelerated research to improve reactor safety. The emphasis should be placed on obtaining the maximum infomation from TMI-2 and on research that will aid in accurately predicting plant response to expected, combined events.
Similar activities as outlined above are being planned or pursued by DOE, EPRI, utilities and other vendors to varying degrees. These concerns directly impact the nuclear industry and as such a coordinated approach must be taken. B&W will participate in these coordination activities as they develop.
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Letter, M. Carbon to Acting Chairman Gilinsky dated April 20, 1979.
Recommendation 1 - initiate immediately a survey of operating procedures for achieving natural circulation, including:
a) event involving loss of offsite power b) consideration of role of pressurizer heaters B&W POSITION See our position on Letter B, Recommendation F 0.
Interim Report No. 3 dated May 16, 1979.
Recommendation 1 - Examine operator qualifications, training and licensing, and requalification training and testing.
Reconmendation 3 - Consider formal review of operating procedures for severe transient by inter-discip in any team, and develop more standardized formats for such procedures.
B&W POSITION As a result of the TMI-2 incident, B&W developed a special supplementary training program for its utility customers. This program was offered to the first plant operators on April 9,1979.
B&W believes that in conjunction with the review of anomalous transients, the need for further changes may become apparent and recommendations will be provided to our utility customers.
As a routine matter, in ccnnection with defining the scope of operator requalification programs, B&W meets wito each customer to determine whether b i.)
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there are special operating experiences which should be emphasized in the requalification training.
i B&W believes there is merit in an interdisciplinary review of selected operating procedures or at ieast of the guideline documents which form the basis for the detailed operating procedures.
We believe that it is important to close the loop between the designer / analyst, trainer, and operator as was done in preparing the recently issued operating guida-lines for handling small treaks.
This would include review of major changes to operating procedures. B&W believes that reviews beyond those currently in place should not be required on routine system and equipment operations.
Recommendation _2 - Establish formal procedures for the use of LER information:
a) in training supervisory and maintenance personnel b) in licensing and requalification of plant operating personnel c) in anticipating safety problems B&W POSITION B&W has initiated the following actions with regard to Licensing Event Reports.
1.
We have requested the f.'RC place B&W on distribution for all PWR LER's issued.
2.
We are conducting a review of in-place systems and procedures to assure they adequately factor in the usage of LER's in evaluating operating plant experience.
3.
B&W is establishing a standing Safety Review Committee which will ga u,
a more formally review abnormal events occurring at our operating plants to assure action is taken when a potential safety concern is identified.
Recommendation 4 - Re-examine comprehensively the adequacy of design testing and maintenance of offsite and onsite AC and DC power supplies with emphasis on:
a) failure modes and effect analyses b) more systematic testing of power system reliability c) improved quality assurance and status monitoring of power supply systems.
BaW POSITION B&W has no comments on this recommendation at this time.
Recommendation 5 - Make a detailed evaluation of current capability to withstand station blackout, including:
a) examination of natural circulation capability under ruch circumstances b) continuing availability of components needed for long term cooling under such circumstan es c) potential for improvement in capability to survive extended blackout.
B&W POSITION BSW has no commcat on this recommendation at tais tice.
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l} Recommendation 6 - Examine a wide range of anomalous transients and degraded accidents which might lead to water hammer with emphasis on: a) controlling of preventing such conditions b) research to provide a better basis for control or prevention of such conditions. B&W POSITION B&W has no comments on this recommendation. Recommendation 7 - Plan and define the NRC role in emergencies, including consideration of: a) assurance that formal emergency plans, procedures, and organizations are in place b) designation of emergency technical advisory teams (names and alternates) c) compilation of an inventory of equipment and materials needed in unusual conditions or situations. B&W POSITION B&W has no comments on this recommendation. Recommendation 8 - Review and revise within three months: a) licensees' bases for obtaining offsite advice and assistance in emergencies from within and outside company, b) licensees' current bases for notifying and providing information to offsite authorities in emergencies. B&W POSITION See our position to Recommendation 3 of R. Fraley letter to Commissioners )bb 2h I dated April 15, 1979. l In addition, B&W has formed a task force to determine the lessons to be learned frem TMI-2. This task was formed in the first half of April. The charter for the task force is:
- 1) Review the technical aspects of the TMI-2 occurrence.
- 2) Develop recommendations for equipment improvements, operator interface, recovery requirements, and incident support.
3. Assess impacc of the TMI-2 occurrence and potential resulting changes in regulatory requirements on Nuclear Power Generation l l Division technical activities. i Recommendation 10 - Expedite resolution of unresolved safety issues by the following means: a) suitable studies on a timely basis by licensees to augment NRC staff efforts. b) use of consultant and contraction support by NRC staff. B/d! POSITION f B&W concurs with the ACRS recommendation that every effort should be made at resolving existing unresolved safety issues in an expeditious fashion, while addressing the new questions that may have arisen from the TMI-2 accident. B&W also supports the recommendation that, where appropriate, the licensees should perform suitable studies on a timely basis and submit proposals for possible implementation of safety improvements; and that the NF.C Staff should rely on consultant and contractor support to evaluate such new studies with minimum impact on the resolution progress b.b f) ((k of the existing unresolved safety items. However, prior to implementing such recommendations, and in order to assess the degree of added support required by the NRC, existing unresolved safety issues must be defined and an assessment must be made of whether or not some or all of the questions which have arisen from the TMI-2 accident are new safety issues or part of previously existing unresolved safety issues. It is B&W position that unresolved safety issues identified prior to the TMI-2 accident are being addressed timely and in an expeditiour fashion consistent with the existing concerns, and that most of the issues which have arisen from the TMI-2 accident are not new safety issues. The TMI-2 accident may have impacted the priority with which the existing unresolved safety issues should be pursued; a re-assessment of the existing priorities by the NRC, the ACRS and the licensees rather than an increase in the required NRC support, may resolve the ACRS concern. Recommendation 11 - Augment expeditiously the NRC Stohf capability to deal with problems in reactor and fuel cycle cFemistry in the following areas: a) behavior of PWR and BWR coolants and other materials under radiation ~ conditions, L! generation, handling and disposal of radiolytic (or other) H 2 at nuclear facilities, c) performance of chemical additives in containment sprays, d) processing and disposal techniques for high and low level radioactive wastes. B&W POSITION /J i; r; B&W has no concent on this recommendation. .'lh >.)
1 i Recommendation 12 - Reconsider whether or not use of the single failure 1 l criterion establishes an appropriate level of reliability for reactor safety systems. B&W POSITION B&W has no comment on this recommendation at this time. Recommendation 13 - With respect to safety research: a) consideration should be given to augmentation of the FY-80 NRC safety research budget, b) consider overtime a larger part of the safety research budget towards exploratory (as opposed to confirmatory) research. B&W POSITION See B&W Position on Recommendati-4; letter R. Fraley to L' nissioners, dated April 18, 1979. Recommendation 14 - Perform design studies of a filtered vented or purgin option for containments for possible use in the event of a serious accident. B&W POSITION B&W has no comment on this recommendation. r' c . } / I.., s, ,}}