ML19289E430
| ML19289E430 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/28/1979 |
| From: | Deyoung R Office of Nuclear Reactor Regulation |
| To: | Boyd R, Mattson R, Stello V Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-34, REF-GTECI-IP, TASK-A-34, TASK-OR, TASK-TF, TASK-TMR NUDOCS 7904180060 | |
| Download: ML19289E430 (22) | |
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UNITED STATES jf,igj' 1j NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
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o MAR o 8 mo f
Generic Activity: A-34 MEMORANDUM FOR: Roger J. Mattson, Director Division of Systems Safety Roger S. Boyd, Director Division of Project Management Victor Stello, Jr., Director Division of Operating Reactors Edson G. Case, Chairman Technical Activities Steering Committee FROM:
Richard C. DeYoung, Director Division of Site Safety & Environmental Analysis
SUBJECT:
DRAFT REPORT OF COMPLETION OF GENERIC ACTIVITY A-34 This report sunnarizes the results of staff efforts to develop guidance to facilitate implementation of Regulatory Guide 1.97, Revision 1, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident."
As such this report documents completion of Generic Technical Activity A-34, " Instruments for Monitoring Radiation and Process Variables During Accidents." For reference, a copy of Task Action Plan A-34 is included as Appendix B to the draft report.
I request your comments and/or concurrence to issue this report as a NUREG. Please provide your comments to Fred Hebdon (x27066) by April 13, 1979.
,[i, m.
//',//h Y-Richard C. DeYoung, Director Division of Site Safety and Environmental Analysis
Enclosure:
As stated u-THIS DOCUMENT CONTAINS%'
wPOOR QUALITY PAGES
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1 73041800(,6
2 cc:
D. Muller F. Hebdon G. Chipman W. Russell W. Minners L. Crocker M. Aycock F. Orr F. Eltawila W. Lefevre E. Butcher R. Emch R. Stoddart D. Lasher R. Bursey R. Priebe K. Parczewski J. Slider S. Block A. Hintze C. Moon A. Bournia E. Licitra C. Stahle W. Kreger W. Houston
y Draft Report of Completion of Generic Activity A-34:
" Instruments for Monitoring Radiation and Process Variables During Accidents" Division of Site Safety and Environmental Analysis Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 e
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1.0 INTRODUCTION
In Decenter,1975 the Staff issued for comment Regulatory Guide 1.97,
" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident." After reviewing the coments received the staff issued Revision 1 to this Regulatory Guide in August 1977.
(A copy of Regulatory Guide 1.97, Revision 1 is provided in Appendix A).
The objective of Regulatory Guide 1.97 is to insure that during and following an accident, appropriate parameters and system functions are monitored in order that plantpersonnel will have sufficient information to take appropriate actions to restrict the co~urses and consequences of an accident. At the start of an accident, the operator cannot always determine what accident has occurred and therefore cannot always determine the appropriate response.
For this reason, the reactor trip and certain safety actions (e.g. emergency core cooling actuation) are designed to be performed automatically during the initial stages of an accident. However, instru-mentation is also necessary to provide infonnation about plant parameters and system functioning that alerts the operator to conditions beyond those expected so that appropriate operator actions may be taken. The operator must have sufficient information available to: (1) determine the course of an accident; (2) make intelligent decisions about taking manual action; and (3) assist in detennining what actions, if any, are needed to execute
. the plant emergency plan.
It should be noted that it is not the intent of Regulatory Guide 1.97 that cperators be encouraged to circumvent automatic features prematurely, but rather that they be adequately informed in order that they can take necessary planned and unplanned actions.
In August 1977, the staff issued Task Action Plan A-34, " Instruments for Monitoring Radiation and Process Variables During an Accident" (a copy of the most recent revision of the Task Action Plan is contained in Appendix B). The purpose of the Task Action Plan is to develop guidance for applicants, licensees and staff reviewers concerning imple-mentation of Revision 1 of Regulatory Guide 1.97.
In the course of implementing the initial phase of the Task Action Plan, it became obvious that Regulatory Guide 1.97 included a few provisions which industry claimed to be impractical at the present time, and other provisions for which more definitive guidance was needed to define acceptable means of compliance. The primary issues in controversy are Positions C.1 and C.3 of the Regulatory Guide.
Position C.1 is intended to insure that the station design includes sufficient instrumentation to meet the objectives described in Position C.1 for each of the Design Basis Accidents normally analyzed by an applicant in Chapter 15 of a Safety Analysis Report.
Position C.3 describes specific instrumentation to be used if accident conditions degrade beyond those assumed in the FSAR. Various industry representatives expressed concern about the ranges of the instruments described in Position C.3 and the implication of monitoring for Class 9
. accidents.
This Position is not explicitly intended to monitor Class 9 accidents.
Position C.3 is intended to provide assurance that even under conditions that degrade far beyond those that are assumed in the accident analyses, the ooerator will have usable instrumentation that will provide a basis for decision making.
The operator must not be placed in a position where all his relevant instrumentation is off-scale.
The ranges of the instruments described in Position C.3 are not based directly on accident scenarios but are based on engineering judgments of the admittedly extreme points beyond which the high probability of failure of important fission product barriers (e.g., reactor pressure vessel or containment structure) would make the need for instru-mentation a moot point.
The remaining Positions in the Regulatory Guide describe the details of the design and qualification of the accident monitoring instrumentation and therefore do not pose the same type of implementation problems.
. 2.0 IMPLEMENTATION During the months since issuance of Regulatory Guide 1.97 and Task Action Plan A-34, the staff and representatives of the nuclear industry have attempted to clarify the intent of the Regulatory Guide. Based on this work the staff has reached the following conclusions concern-ing implementation of Regulatory Guide 1.97 Revision 1.
1.
The large amount of experience accumulated to date permits identification of those parameters that should be monitored to satisfy Position C l.
The list of parameters is provided as Appendix C.
The staff will require that these parameters be monitored on all plants for which a construction permit application was docketed after September 30, 1977 (as per section D of Regulatory Guide 1.97 Revision 1). The accident monitoring instrumentation of plants for which a construction permit application was docketed prior to September 30, 1977 has been reviewed as part of the licensing process. Although the parameters monitored at specific plants may be different than those specified in Appendix C, the staff still believes that with the addition of the instruments described in Position C.3, existing accident monitor-ing equipr.ent is acceptable. Therefore, the staff has concluded that the resources that would be required to backfit the instruments required to monitor the parameters listed in Appendix C would not be justified based on the benefits derived from having a standard set of accident monitoring instruments on all plants.
. 2.
The staff concludes that technology currently exists to permit implementation of the instrumentation described in Positions C.3.a through C.3.c.
Prior to issuance of Regulatory Guide 1.97 Revision i the staff did not require that accident monitoring instrumentation be provided with ranges extending beyond the conditions expected to result from Design Basis Accidents. For the reasons discussed in Section 1.0, the staff now believes that such instrumentation should be required on all plants. Therefore, the staff requires that the instrumentation described in Position C.3.a through C.3.c be implemented for reactor plant license applications and all plants licensed for construction or operation.
3.
With respect to Position C.3.d, the staff is not certain that existing release rate monitoring technology is sufficient to permit adequate monitoring of the ranges of radioactivity release rates that might be encountered if, as assumed in Position C.3, conditions degrade beyond those expected to result from the Design Basis Accidents. Therefore, the staff will delay requiring implementation of Position C.3.d until studies of the capabilities of existing re-lease rate monitoring technology can be undertaken.
4.
It has been pointed out that it may not be feasible to qualify instrumentation to extreme conditions consistent with the instru-ment ranges described in Position C.3, particularly radiation 8
levels inside containment of up to 10 rads / hour (Position C.3.b).
The staff agrees that qualification of instrumentation located inside containment to such levels may not currently by possible.
. However, the staff believes that all of the instrumentation de-scribed in Position C.3 can either be shielded or located outside the containment, where a less hostile environment would exist, and appropriately calibrated.
5.
Position C.6 states that accident monitoring instrumentation should be designed so that a single failure does not prevent the operator from accomplishing the objectives of Position C.1.
However, it is the staff's position that redundant instrumentation is not re-quired on each train of a system that has a redundant counterpart.
6.
The staff worked closely with several applicants for construction permits and operating licenses, and with the Atomic Industrial Forum Ad Hoc Committee on Post Accident Monitoring Instrumentation.
All of the concerns raised by the involved industry representatives have not been resolved to the satisfaction of all parties. However, the staff believes that sufficient guidance has been developed so that Task A-34 can be classified as complete. The staff will continue to work with the industry representativ.s in an attempt to resolve any minor issues that remain unresolved.
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Revision 1
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U.S. NUCLEAR REGULATORY COMMISSION August 1977 4
!),- lff) REGULATORY GUIDE 0
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OFFICE OF STANDARDS DEVELOPMENT REGULATORY GUIDE 1.97 INSTRUMENTATION FOR LIGHT-WATER-COOLED NUCLEAR POWER PLANTS TO ASSESS PLANT CONDITIONS DURING AND FOLLOWING AN ACCIDENT A. INTRODUCTION w hether the reactor trip and engineered-safety-feature systems are functioning properly: (3) deter-Criterion 13. " Instrumentation and Control." of mine whether the plant is responding properly to the Appendix A. " General Design Criteria for Nuclear safety measures in operation:( 1) provide information Power Plants," to 10 CFR Part 50, "Licensi g of to the operator that will enable him to determine the Production and Utilization Facilities " includes a re.
potential for breaching the barriers to radioactivity quirement that instrumentation be provided to release: (5) furnish data for deciding on the need to monitor variables and systems for accident condi-take manual action if an engineered safety feature tions as appropriate to e'nsure adequate safety.
malfunctions or the plant is not responding effective-ly to the safety systems in operation:(6) allow for ear-Criterion 19, " Control Room." of Appendix A to ly indication of the need to initiate action necessary 10 CFR Part 50 includes a requirement that a control to protect the public and for an estimate of the room be provided from which actions can be taken to magnitude of the impending threat: and (7) aid in maintain the nuclear power unit in a safe condition determining the cause and consequence of the event under accident conditions, including loss-of-coolant for postaccident investigation.
accidents.
Criterion 64, " Monitoring Radioactivity Releases, of Appendix A to 10 CTR Part 50 meludes At the start of an accident, the operator cannot a requirement that means be provided for monitoring always determine immediately what accident has oc-the reactor containment atmosphere, spaces contain-curred or is occurring and therefore cannot always ing components for recirculation of loss-of-coolant determine the appropriate response. For this reason, accident fluid, effluent discharge paths, and the plant the reactor trip and certain safety actions (e.g.,
environs for radioactivity that may be released from emergency c re coolmg actuation, containment isola-postulated accidents.
tion, or depressurization) are designed to be per-formed automatically during the mitral stages of an This guide describes a method acceptable to the accident. Instrumentation is also provided to indicate NRC staff for comp 1ving with the Commission's re-mf rmation about plant parameters required to quirements to provide instrumentation to monitor enable the operation of manually initiated safety-plant variables and systems during and following an related systems and other appropriate operator ac-accident in a light water-cooled nuclear power plant.
tions.
The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concur-red in the regulatory position.~
Examples of serious events that threaten safety if conditions degrade bevond those assumed in the Final Safety Analysis R' port are loss-of-coolant acci-
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e B. DISCUSSION dents (LOCAs) reactivity excursions, and radioac-Monitored variables and systems are used by the tivity releases. Such events require that the operator operator in accident surveillar' ice to (!) assist in deter-understand in a short time period, the state of mining the nature of an accident: (2) determine re diness of engineered safety features and their potential for bemg challenged by an accident m
- Unes indicate substanuve changes from previous issue.
progress.
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20555. Artemen- 0 rector. 3 v.sen o' Docuraeat Contros APPENDIX A
To determine the important variables and the be used for both accident and normal operation.
systems whose values or status are needed by the However, it is essential that instrumentation so up-operator and, therefore, the monitoring instrumenta-graded does not comprom:se the accuracy and sen.
tion needed by the operator, a study (Ref.1) was sitivity required for normal operation.
made of a range af postulated accidents. The study concluded that the following capabilities are most im'.
It should be noted that in the safety analysis many portant to ensuring that the power plant poses no parameters may be Mentified that will provide threat to public safety after an accident: reactor shut.
desirable, but less essential, information for the down, core cooling, containment isolation, and the operator. Any instrumentation used to measure these maintenance of ' containment pressure control, less essential (i.e., " backup") parame.ers is outside primary system p?cssure control, and a heat transfer the scope of this guide.
path from the core to a heat sink. These vital capabilities are designed to,, reserve the integrity of C. REGUI.ATORY POSITION the barriers to radioactivity release (i.e.. the fuel clad-ding, reactor coolant boundary, and containment).
- 1. For the postulated accidents h.sted.m Chapter 15 of Regulatory Guide 1.70 (Ref. 2), the applicant it is essential that the required instrumentation be should perform detailed safety analyses necessary to capable of surviving the accident environment in determine the parameters to be measured and the in-which it is located for the length of time its function is strument ranges, responses, accuracies, and length of required. It could therefore either be designed to time required to provide the operator with the infor-withstand the accident environment or be protected mation necessary to:
by a local artificial environment. If the environment surrounding an instrument component is the same
- a. Assist in determining the nature of an acci-for accident and normal operating conditions (e.g.,
- dent, the instrumentation components in the main control
- b. Determine whether the reactor trip and room), the instrumentation components need no engineered-safety-feature systems are functioning special environmental capability.
- properly,
- c. Determine whether the plant is responding it is important that accident-monitoring in-properly to the safety measures in operation, strumentation components and their mounts that
- d. Determine the potential for breaching the cannot be located in other than non-Seismic barriers to radioactivity release, Category I buildings be conservatively designed for
- e. Decide on the need to take manual action if the intended service.
an engineered safety feature malfunctions or the plant is not responding effectively to the safety Parameters selected for accident monitoring can be systems in operation, and selected so as to permit relatively few instruments to
- f. Allow for early indication of necessary action provide the essential information needed by the to protect the public and for an estimate of the operator for postaccident monitoring. Further, it is magnitude of the impending threat.
prucent that a limited number of those parameters (e.g., containment pressure) be monitored by instru-The guidelines in Reference 1, along with the ments qualified to more stringent environmental re-guidelines in Reference 3 dealing with monitoring in-quirements and with ranges that extend to the max-side the power plant, may be used to make such imum values that the selected parameters can attain analyses.
under worst-case conditions; for example, a range for the containment pressure monitor extending beyond
- 2. The instrumentation necessary to provide the the design pressure of the containment.
information noted in regulatory position I should be specified along with justification to show that the in-Normal power plant instrumentation remaining strumentation is adequate to provide the operator functional for all accident conditions can provide in-with the necessary information. The safety analyses dication. records, and (with certain types of in-should provide the information necessary to select struments) time-history responses for many the appropriate type of accident-monitoring instru-parameters important to following the course of the ment: to specify the range. accuracy, transient accident. Therefore, it is prudent to select the re-response, environmental and seismic qualifications.
quired accident monitoring instrumentation from the and insensitivity to variations of energy supply; and normal power plant instrumentation. Since some ac-to specify the method of recording, when recording is cidents impose severe operating requirements on in-deemed necessary.
strumentation components, it may be necessary to upgrade some instrumentation components to with-
- 3. A limited number of additional accident-stand the more severe operating conditions and to monitoring instruments should have ranges that ex-measure greater variations of monitored variables tend to the maximum values that selected parameters !
that may be associated with the accident if they are to can attain under worst-case conditions, and the in-I
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1.97-2
strumentation components should be qualified to
- 8. To the extent practical, accident-monitoring in-withstand the higher level of environmental condi-strumentation inputs should be from sensors that tions in which they will be required to function. The:e directly measure the desired variables.
.s parameters and associated maximum values to be
- 9. To the extent practical, the same instruments measured by the instruments should include, but not should be used for accident monitoring as are used necessarily be limited to, the following:
for the normal operations of the plant to enable the
- a. Containment pressure: 3 times design pres-perat r to use, during accident situations, instru-sure for concrete: 4 times design pressure for steel.
ments with which he is most famihar. However,
- b. Radiation level inside containment: 10' rads where the required range of accident-momtonng in-strumentation results in a loss of instrumentation p
- c. Reactor coolant pressure: 3 times design pres-sensitivity in the normal operating range, separate in-struments should be used.
5"'*-
- 10. The accident-monitoring instrumentation
- d. Plant radioactivity release rate through iden-should be specifically identified on control panels so tifiable release points: (plant dependent) (range dependent on maximum release rate postulated for a that the operator can e:sily discern that they are in-tended for use under accident conditions.
given release point).
- 4. The accident-monitoring instrumentation II.Any equipment that is used for both accident should be qualified in accordance with Regulatory momtonng and nonsafety functions should be clas-G uide 1.89. " Qualification of Class IE Equipment for s fied as part of the accident-monitoring instrumenta-Nuclear Power Plants."
tion. The transmission of signals from accident-monitonng equipment for nonsafety system use Instrumentation that is Seismic Category I. as-should be through isolation devices that are classified defined by Regulatory Guide 1.29, "Scismic Design as part of the accident-monitoring instrumentation
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Classification," should continue to function within and that meet the provisions of the document.
the required accuracy following, but not necessarily
- 12. Means should be provided for checking, with a during, a safe shutdown earthquake.
high degree of confidence, the operational Instrumentation components and their mounts availability of each accident-monitoring channel, in-that cannot be located in other than non-Seismic cluding its input sensor, during reactor operation.
Category I buildings need not meet Seismic Category This may be accomplished in various ways, for exam-I criteria.
ple:
I
- a. By perturbing the monitored variable:
- 5. Those parameters selected for accident-
- b. By introducing and varying, as appropriate, a monitoring instrumentation that provide transient or substitute input to the sensor of the same nature as trend information necessary for the operator to per-the measured variable; or form his role should be recorded. Records of
- c. By cross-checking between channels that bear parameters that provide information related to the a known relationship to each other and that have determination of radioactivity release rates and total readouts available.
radioactivity releases should be considered necessary.
- 13. Servicing, testing, and calibration programs
- 6. The accident-monitoring instrumentation should be specified to maintain the capability of the should be designed so that a single failure does not accident-monitored instrumentation. For those in-prevent the operator from accomplishing the objec-struments where the required interval between testing tives of regulatorv position 1.
will be less than the normal time interval between generating station shutdowns. a capcbility for testine NOTE: "S.mgle failure...meludes such events as during power operation should be provided.
the shorting or opencircuiting of interconnecting signal or power cables. It also includes single credible EXCEPTION: "One-out-of-two" systems are
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malfunctions or events that cause a number of conse-permitted to violate the single-failure criterion during quential component module, or channel failures. For channel bypass provided that acceptable reliability of example. the overheating of an amplifier module operation can be otherwise demonstrated. For exam-would be a " single failure" even though several tran-ple. the bypass time interval required for a test.
sistor failures might result. Mechanical damage to a calibration, or maintenance operation could be mode switch would be a " single failure" although shown to be so short that the probability of failure of several channels might become involved.
the active channel vould be commensurate with the
- 7. The accident-monitoring instrumentation chan, probability of failure of the "one-out-of-two" i ne!s that are redundant should be electrically in-systems dunng its normal interval between tests.
l dependent, energized from station Class IE power.
- 14. Whenever means for bypassing channels are in-l and physically separated in accordance with ciuded in the design. the design should permit ad-l Regulatorv Guide 1.75. " Physical Independence of ministrative control of the access to such bypass Electric Systems."
means.
1.97-3
- 15. The design should permit administrative control plying with the specified portions of the Commis-of the access to all setpoint adjustments, module sion's regulations, the method described herein will calibration adjustments, and test points.
be used in the evaluation of submi'tals for construc-tion permit applications docketed after September
- 16. The accident monitoring instrumentation 30,1977.
design should minimize the development of condi-tions that would cause meters, annunciators, recorders, alarms, etc., to give anomalous indications REFERENCES confusing to the operator.
- 1. Battelle-Columbus Laboratories, " Monitoring
- 17. The instrumentation should be designed to Post-Accident Conditions in Power Reactors,"
facilitate the recognition, location, replacement, BMI-X-647, April 9,1973.
repair, or adjustment of malfunctioning components or modules.
- 2. U.S. Nuclear Regulatory Commission, " Standard Format and Content of Safety Analysis Reports D. IMPLEMENTATION for Nuclear Power Plants," NUREG-75/094, The purpose of this section is to provide informa.
Regulatory Guide 1.70, Revision 2, September tion to applicants regarding the NRC stafTs plans for 1975.
using this regulatory guide.
- 3. BNWL-1635, " Technological Considerations in Except in those cases in which the applicant Emergency Instrumentation Preparedness," May proposes an acceptable alternative method for com-1972.
)
e 1.97-4
Task A-34 INSTRUMENTS FOR MONITORING RADIATION AND PROCESS VARIABLES DURING ACCIDENTS Lead NRR Organization:
Division of Site Safety and Environmental Analysis (DSE)
Lead Supervisor:
Richard H. Vollmer A/D for Site Analysis, DSE Task Manager:
Frederick J. Hebdon, Project Manager, Environmental Projects Branch 1, DSE Applicability:
All Reactor Types Projected Completion Date:
November 1978 APPENDIX B
Task A-34 Rev. No. 1 May 1978 j
1.
DESCRIPTION OF PROBLEM To develop criteria and guidelines to be used by applicants, licen-sees and staff reviewers to support implementation of Regulatory Guide 1.97, Revision 1 (Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident).
Such criteria and guidelines would provide specific guidance on functional and operational capabilities required of the various classes of instruments, including inplant and explant instruments.
Where such guidance cannot be provided, the rationale to be applied to derive requirements for specific situations will be provided.
2.
PLAN FOR PROBLEM RESOLUTION A.
Detailed guidance and acceptance criteria concerning implementa-tion of Regulatory Guide 1.97 has not yet been developed.
Therefore, the members of this Task Group will answer questions that arise before and during the development of the required proposals for implementation of Regulatory Guide 1.97 for the lead plants described below.
In.this way, the Task Group will develop the necessary guidance as it is needed by the lead plant applicants.
The Task Group will also be responsible for the review of submittals made by the lead plant applicants.
B.
There are two aspects of the implementation of Regulatory Guide 1.97, Revision 1 (Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident) that must be considered.
(1) Position C.3 of RG 1.97 requires the installation of specific instrumentation to follow the course of an accident (IFCA).
l The staff has determined that this requirement should be satisfied in as timely a manner as possible.
The Task Group established by this Task Action Plan will identify lead plants (at least one BWR and one PWR) for implementa-tion of Position C.3, will answer questions raised by the lead plant applicants, and will assume responsiblity for the review of the proposals for implementation of Position C.3 l
that are submitted.
Based on the experience gained during
[
this review, the Task Group will prepare uniform review procedures and acceptance criteria to be used by the staff for the review of subs,equent implementation proposals.
1 l
A-34/1 i
Task A-34 Rev. No. 1 May 1978 (2); Full implementation of RG 1.97 requires the applicant / licensee to prepare a Safety Analysis which is reviewed by the staff.
Lead plants (at least one BWR and one PWR) for full implementation of RG 1.97 will be designated.
The Task Group established by this Task Action Plan will assist the lead plant applicants in the development of the required Safety Analyses by answering questions from the applicants.
The Task Group will review the Safety Analyses when they are submitted.
Based on the experience gained during the development and review of the Safety Analyses for the lead plants, the Task Group will prepare guidance to assist other applicants / licensees in the development of the required Safety Analysis and acceptance criteria to be used by the staff to review the Safety Analyses submitted.
C.
Description of the End Product of Task Group (1) A letter to all applicants and licensees containing guidance to facilitate the preparation of Safety Analyses required by RG 1.97.
(2) Revision of various Standard Review Plans to provide for the uniform review of required Safety Analyses and Pro-posals for Implementation of Position C.3.
(3) Recommendation for revision of RG 1.70, Standard Format and
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Content of SAR's for Nuclear Power Plants.
(4) Recommendations for confirmatory research as required.
(5) Recommendations for revisions to RG 1.97.
3.
BASIS FOR CONTINUED PLANT OPERATION AND LICENSING PENDING COMPLETION.
OF TASK As described in Sections 1 and 2, the issue addressed by this task is the timely development of criteria and guidelines to support full implementation of Reg. Guide 1.97, Revision 1 for CP's, OL's and Operating Reactors.
Full implementation of Reg. Guide 1.97, Revision 1 requires the applicant / licensee to prepare a Safety Analysis (of instruments to follow the course of an accident which are part of this task as opposed to instruments to prevent an accident which are not) which is to be reviewed by the staff.
This task will provide guidance to applicants for preparation of the Safety Analysis report and criteria and analyses by which the staff will review the report.
The current staff review process assures that the likelihood of serious accidents is extremely low.
Implementation of the defense-in-depth concept and the single failure criterion assure that there is no undue risk to the health and safety to the public.
There A-34/2
Task A-34 Rev. No. 1 May 1978 is, however, a residuum of risk from accidents which are more severe than those evaluated in the applicant's Safety Analysis Report and reported on in the staff's Safety Evaluation Report.
This residuum of risk is small when compared to other risks in society and as such, specific designs to accommodate accident conditions contributing to these risks is not required.
The staff has, however, determined that it is prudent to provide additional capability for plant operators to identify accident conditions which could lead to significant consequences.
Full implementation of the provisions of Regulatory Guide 1.97, Revision I will provide additional assurance that the operator will be able to identify the need for and execute accident mitigation procedures for design basis accidents and be able to-identify and act to rectify accident conditions which have been degraded beyond the design basis.
The low level of the residual risk for current designs presents no undue risk to the health and safety of the public.
3.
NRR TECHNICAL ORGANIZATIONS INVOLVED These branches will carry out their responsibilities through participation on the Task Group.
A.
Accident Analysis Branch (DSE) review the Safety Analyses required by RG 1.97 for the lead plants to ensure that varia-tions in plant variables are adequately defined, from a consequences viewpoint, for the Design Basis Accidents analyzed.
This review will also include evaluation of operator interaction (e.g., procedures, actions, timing) for utilizing instrumentation to follow the course of an accident (IFCA) to assess and minimize risk.
Develop guidance for applicants / licensees and uniform review procedures for the staff to support the implementation of RG 1.97 on other plants.
Review the plans for implementation of Position C.3 for lead plants and develop uniform review procedures for the staff to use to review implementation proposals for other plants.
(Manpower Requirements:
1 reviewer, 2MM per reviewer.)
8.
Reactor Systems Branch (DSS), Containment Systems Branch (DSS),
Auxiliary Systems Branch (OSS), Power Systems Branch (DSS)
Review the Safety Analyses for the lead plants to ensure that significant process variables required to monitor the course of Design Basis Accidents, from a systems performance viewpoint, are identified.
This review will also include evaluation of operator interactions (e.g., procedures, actions, timings) for A-34/3
Task A-34 Rev. No. 1 May 1978 utilizing IFCA to optimize system performance.
Develop guidance for applicants / licensees and uniform review procedures for the staff to use to implement RG 1.97 on other plants.
(Manpower requirements:
1 reviewer per branch, 3MM per reviewer in RSB, IMM per reviewer in CSB, and PSB.)
C.
Radiological Assessment Branch (DSE) and Effluent Treatment Systems Branch (DSE) - develop criteria for application of inplant and explant radioactivity monitoring systems to follow the course of an accident during various accident situations and accident scenarios.
Review the Safety Analyses for the lead plants to ensure that plant radiation sources are adequately defined and that radiation monitoring is adequate from the viewpoint of protection of the health and safety of utility l
staff personnel, of emergency program personnel and of the public outside the immediate plant environs.
(Manpower require-ments:
1 reviewer, 2 MM per reviewer for RAB and 1 reviewer, 1 MM per reviewer for ETSB).
iH!
i D.
Instrumentation and Control Systems Branch (DSS) - review the Safety Analyses for the lead plants to ensure that IFCA is appropriately designed, will remain operable as required, and will accurately represent the information required by the operator.
This review will include consideration of maintenance and test-ing of instrumentation.
Develop guidance for applicants / licensees and review procedures for the staff to use to implement RG 1.97 on other plants.
Review the plans for implementation of Position jj C.3 for lead plants and develop uniform review procedures for the i,
staff to support the review of implementation proposals for j,
other plants.
(Manpower Requirements:
1 reviewer, 2MM per reviewer.)
E.
Operator Licensing Branch (DPM) - assist in evaluating operator interactions and expected operator responses to identify the I
instrumentation required and the procedures to be followed to deal with Design Basis Accidents.
Develop guidance for applicants /
licensees and uniform review procedures for the staff to support implementation of RG 1.97 on other plants.
(Manpower Require-ments:
1 reviewer, IMM per reviewer.)
F.
Emergency Planning Branch (DPM) - review the Safety Analyses for lead plants and the applicant's Emergency Plan to ensure that the operator will be supplied with the information needed to permit him to provide authorities responsible for implementation of Emergency Plan with accurate and timely recommendations concerning implementation of all or part of the plan.
Develop guidance for applicants / licensees and uniform review procedures for the staff to support implementation of RG 1.97 on other A-34/4
Task A-34 Rev. No. 1 May 1978 plants.
Review the plan of Position C.3 for lead plants and develop uniform review procedures for the staff to support the review of implementation proposals for other plants.
(Manpower Requirements:
1 reviewer, 1MM per reviewer.)
G.
Environmental Projects Branch 1 (DSE) - Provide a Task Manager to serve in the principal management function for the project.
(Manpower Requirements:
1 project manager, 3MM manager.)
H.
Operating Technology (DOR) - Review and comment on materials developed by the Task Group.
Adapt the criteria and guidance developed by the Task Group for use by reviewers and licensees of operating reactors.
CManpower Requirements:
1 reviewer per branch (4 branches), 1 MM per reviewer.3 I.
Other Branches in NRR may be called upon to provide technical support to the Task Group as needed on a consultation basis.
(Manpower Requirements:
Total 1 MM.)
5.
TECHNICAL ASSISTANCE FUNDS AND CONFIRMATORY RESEARCH FUNDING REQUIRED It is not presently anticipated that technical assistance funding or confirmatory research funding will be required to directly support this Task Group.
Two projects (described below) may produce data i
that will support the activities of this Task Group.
A.
00R has an existing technical assistance contract with BNL to evaluate certain operating plants to determine the capability of existing effluent radiation monitors to measure radioactivity releases through anticipated release paths from postulated accidents.
The funding level for this program is $25K for FY 1977 and FY 1978.
B.
DSE has an existing technical assistance contract with Allied Chemical Company (INEL) to develop bases for the specification of gaseous effluent accident monitoring instrumentation.
The funding level for this program is $40K for FY 1977.
~
6.
INTERACTION WITH OUTSIDE ORGANIZATIONS The Task Group will maintain close cont &ct with applicants for the lead plants.
7.
ASSISTANCE REQUIREMENTS FROM OTHER NRC 0FFICES Office of Standards Development - Assist in the development of subsequent revisions of RG 1.97 and other Regulatory Guides based on experience gained during the review of the lead plants.
A-34/5
U:tl [
Task A-34 Rev. No. 1 May 1978 8.
F0TENTIAL PROBLEMS Based on preliminary studies, as exemplified in BNWL-1635, it is anticipated that many plant evaluations, particularly those for operating plants, will show the need for monitoring equipment not commercially available and, therefore, a lead time of six months to two years may be necessary for development, procurement, and installation of monitoring equipment.
l l
A-34/6
INSTRUMENTATION TO FOLLOW THE COURSE OF AN ACCIDENT Parameter Containment pressure Hot leg flow (PWR)
Cold leg flow (PWR)
Level in steam generator Main steamline flow rate Pressure of reactor coolant Pressurizer level (PWR)
Radiation level in condenser air ejector Steam-generator pressure (PWR)
Temperature of reactor coolant Position of Valves in Vital Systems Component cooling water system Flow Containment cooling fan flow Containment spray flow Containment sump and suppression pool level Control rod position indicators Emergency cooling water storage tank level Emergency filter train operation Emergency ventilation system (s) damper positions Injection flow Power (Neutron flux)
Residual heat removal flow Appendix C
Parameter Safety infection flow Status of power supplies Ultimate heat sink temperature and level Area radiation levels in auxiliary buildings Boron concentration and/or flow (PWR)
Containment temperature Hydrogen concentration in containment Radiation level in containment Radiation level in main steamline (BWR)
Reactor vessel coolant level Temperature of space in vicinity of vital equipment Activity levels in surface and ground water Activity release rate from principle plant vents and discharge points Wind direction, speed and vertical temperature di fference Environmental Radiation ~ Levels
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