ML19308B722
| ML19308B722 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/17/1976 |
| From: | Moeller D Advisory Committee on Reactor Safeguards |
| To: | Gossick L NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| Shared Package | |
| ML19308B699 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-01.114, RTR-REGGD-1.097, TASK-TF, TASK-TMR NUDOCS 8001160792 | |
| Download: ML19308B722 (1) | |
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS NUCLEAR REGULATORY CO.'. T.USSION WASHINGTON. D. C. 20555 oV August l'i, 1976
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Ie. I.ee v. Gossick
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Exec tive Director of Operations U.S. Ecclear Regulatory Cc=nission P =M gten, DC 20555 5~
At l's 196th teeting,, August 12-14, 1976, the ACES approved Revision 1 to Reg'htcry Guide 1.114, " Guidance on Being Operator at the Controls of a Xuclear Power Plant."
With regard to Revision 1 to Regulatory Guide 1.97, " Instrumentation for Ligh -Water-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," the Cornittee believes that a relatively limited n=ber of primary indicators (pressure, temperature, radiation level, etc.) should have instrt=ent ranges t/nich go beyond Class 8 acci-dents, and that these instrt=ents should meet the various environmental gcalification criteria cited, as practical.
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With regard to other instruments providing less essential but significant, or less essential but desirable, information, the Coraittee believes con-sideration shoulti be given to a range of less rigorous qualification criteria.
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Sincerely yours, o
_ce U.~l'oelIer Chairman cc:
B. C. Rusche, NRR R. Minogue, 05D G. Arlotto, OSD S. J. Chilk, SECY i
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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 whether the reactor trip and enginected-safety-A. INTRODUCTION 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 Desien Criteria for Nuclear safety measures in operation:(4) proside 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 radioactisity 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 ensure adequate safety, malfunctions or the plant is not responding effectise-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 w hich actions can be taken to magnitude of the impending tbreat; 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 At the start of an accident. the operator cannot Releases." of Appendix A to l0 CF R Part 50 include's always determine immediately w hat accident has oc-a requirement that means be provided for monitoring curred or is occurring and therefore cannot alwap the reactor containment atmosphere, spaces contain-determine the appropriate response. For this reason.
ing components for recirculation of loss-of-coolant the reactor trip and certain safety actions (e.g..
accident fluid, eftluent discharge paths and the plant environs for radioactivits that may be released from 8.mergency core cooling actuation. containment isola-postulated accidents.
tion. or depressurization) are designed to be per-formed automatically during the imtial stages of an accident. instrumentation is also provided to indicate This guide describes a method acceptable to the information about plant parameters required to NRC staff for comphing with the Commission's re.
enable the operation of manually initiated safety-quirements to provide instrumentation to monitor plant variables and systems during and following an related systems and other appropriate operator ac-tions.
accident in a light-water-cooled nuclear power plant.
The Advisory Committee on Reactor Safeguards has been consulted concernirg this guide and has concur-Examples of serious events that threaten safety if red in the regulatory position.
conditions degrade beyond those assumed in the Fin 1 S fety Analpis Report are loss-of-coolant acci-B. DISCUSSIOil dents (LOC As), reactisity excursions. and radioae-thity releases; Such events require that the operator Monitored variables and systems are used bs the under tand, in a short time period, the state of operator in accident surseillance to (1) assist in deter-readinew f engineered safety features and their minine the nature of an accident: (2) determine potential for being challenged by an accident in
- Lines indicate substanthe changes from previous issue.
progress.
USNRC REGULATORY GUlDES co-.an %io i a.
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. To determine the important variables and the be used for both accident 'and normal operation.
syst;ms.whose salues or status are needed by the Howeser, it is essential that instrumemtion <o up-1
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operator and. therefore. the monitorinc instrumenta-graded does not compromise the accur ;) and sen-l tion needed by _ the operator, a study (Ref.1) was sitisity required for normal operation.
made of a range of postulated accidents. The studs concluded that the following capabilities are most im'.
!* shot!d oe noted that in the safety analysis man) portant to ensuring that the power plant poses no parameters may be identified that will proude {
threat to public safety after an accident: reactor shut.
desirable, but less e3sential, information for the
- dow n. core cooling. containment i>olation. and the -
operator. Any instrumentation used to measure these maintenance of containment pressure control, less chential (i.e.. " backup") parameters i> outside
_ primary system pressure control, and a heat transfer the scope of this guide.
path from the core to a heat sink. These sital capabilities are designed to preserve the integrity of C. REGULATORY POSITION the barriers to radioactivity release (i.e.. the fuel clad-ding. reactor coolant boundary. and containment).
- l. For the postulated accidents listed in Chapter 1.s. of Regulatory Guide 1.70 (Ref. 21. the applicant it is essential that the required instrumentation be should perform detailed safety analyses necessary to capable of sursiving the accident environment in determine the parameters to be measured and the in-w hich 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.c.,
dent.
the instrumentation components in the main control
- b. Determine whether the reactor trip and room), the instrumentation components need no engineered-safety-feature sy stems 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 w
cannot be located in other than non-Seismic barriers to radioactisity 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 effectisely to the safety Parameters selected for accident monitoring can be ustems in operation. and
- f. Allow for earh indication of necessart action selected so as to permit relatisely fea instruments to provide the essential information needed by the to protect the public and for an estimate" of the operator for postaccident monitorinc. Further. it is magnitude of the impending threat.
prudent 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 ensironmental re-guidelines in Reference 3 dealing with monitoring in-quirements and with ranges that extend to the mas-side the power plant. may be used to make such imum salues that the selected parameters can attain analy ses.
under worst-case conditions: for example. a range for the containment pressure monitor extending beyond
- 2. The instrumentation necessary to provide the
.he design pressure of the containment.
information noted in regulatory position I should be 8
specified along with justification to show that the in-Normal power plant instrumentation remaining strumentation is adequate.to proside the operator functional for all accident conditions can p;mide in-with the necessary information. The safety analyses dication. records, and (with certain types oi in-should proside the information necessary to select
.struments) time history responses for man) 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, ensironmental and seismie qualifications. t
_ quired accident monitoring instrumentation from the and insensitisity to variations of energy supply: and normal power plant instrumentation. Since some ac-to specify the method of recording. w hen recording is cidents impose >evere operating requirements on in-deemed necessary.
strumentation components. it may be necenary to j
upgrade some instrumentation components to with-
- 3. ' A limited number of additional accident-s stand the more sesere operating conditions and to momtoring mstruments should hase rances that ex-measure creater variations of monitored sariables tend to the maumum salues that selected parameters that may be associated with the accident if they are to can attam under worst-ease conditions. and the m-1.97 2 o
c.
- O strumentation components should be qualified to
- 8. To the extent practical. accidW monitoring in-oithstand the higher level of environmental condi.
strumentation inputs should be from sensors that tions in w hich they will be required to function. These directly measure the desired variables.
parameters and awociated 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 monitormg as are used necessaril; be !;mited to, the following:
for the normal operations of the plant to enable the operator to use, during accident situations. instru-
- a. Containm' nt pressure:, 3 times design pres-ments with which he is most familiar. Howeser.
e sure for concrete: 4 times design pressure for steel.
where the required rance of accident monitoring m-
- b. Radiation level mside containment: 10' rads
.strumentation results in a loss of instrument:aior$
per hour.
- c. Reactor coolant pressure: 3 times design pres-sensitivity in the normal operatinc rance, 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 pomts: (plant dependent) (range that the operator can easily discern that thev are in-dependent on maumum release rate postulated for a tended for use under accident conditions. '
given release point).
- 4. The accident monitoring instrumentation II Any equipment that is used for both accident m
tormg and nonsafety functions should be clas-should be qualified in accordance with Reculatort Guide 1.89. " Qualification of Class IE Equipment fo'r 5.n ed as part of the accident momtoring instrumenta-
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Nuclear Power Plants.-
tion. The transmission of signals from accident-monitoring 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. " Seismic Design as part of the accident-monitorme instrumentation Classification." should continue to function within and that meet the provisions of the document.
the required accuracy following. but net necessarily
- 12. Means should be provided for checkinc. with a during. a safe shutdow n earthquake.
high degree of confidence, the ope'ational Instrumentation components and their mounts availability of each accident monitoring channel. in-that cannot be located in other than non-Seismic ciuding its input sensor. during reactor operation.
Category I buildings need not meet Seismic Category This may be accomplished in various w ays. for exam-I criteria.
ple:
- a. Bs perturbing the monitored sariable:
- 5. Those parameters selected f, or accident-
- b. B"y 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. Bv cross-checking between channels that bear parameters that proside 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 he designed so that a sincle failure does not accident-monitored instrumentation. For those in-prevent.the operator from accornplishing the objec-struments u here the required interval between testing tives of reculators position 1.
will be less than the normal time interval between cenerating station shutdow ns, a capability for testinc NOTE: " Single failure..., eludes such events as
'durinc power operation should be provided.
m the 3hortmg or opencircuitmg of interconnecting signal or power cables. It also includes single credible EXCEPTION: "One-out-of-tw o" systems are malfunctions or esents that causei number of conse-permitted to violate the single-failure crite'rion during quential component. module. or channel failures. For eh.mnel hypass prosided that acceptable reliabilits of example. the oserheating of an amplifier module operation can be otherwise demonstrated. For eiam-would be a " single failure" esen though seseral tran-ple. the bspass time interval required for a, test.
sistor failures might result. Mechanical damage to a calibration. or maintenance operation could be
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mode switch would be a " single failure" although shown to be so short that the probab"its of failure of seseral channek might become insohed-the active channel would be commensurate with the
' The acciden'...cnitorine instrumentation chan, probability of-failure of the "one-out-of-two" neh that are cedundant shbuld be electrically m-
'}'tems during its normal interval between tests.
dependent. er ergized from station Claw I F pow er.
14 Wheneser means for bypawing channeh are m-
+
l.and physically separated. m accordance with eluded m the design. the design should permit ad-
' Regulatory G uide 1.~5. "I h> uea: Independence of
- minisiratne control of the aecen to such bypaw I.lectrie. S stems '
3 means 1.97 3
'l
- 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 rnethod described herein will calibration adjustments, and test points.
be used in the evaluation of submittals for construc-tion permit applications docketed after September
- 16. The accident monitoring ' instrumentation 30,1977.
design should minimL e the development of condi-tions that would cause meters, annunciators, REFERENCES recorders, alarms, etc., to give anomalous indications confusing to the operator,
- l. Battelle Columbus Laboratories, " Monitoring Post Accident Conditions in Power Reactors,"
- 17. The instrumentation should be designed to facilitate the recognition, location, replacement.
BMI X 647 April 9,1973.
repair, or adjustment of malfunctioning components
- 2. U.S. Nuclear Regulatory Commission," Standard er rnodules.
for Nuclear Power Plants.fI ^"*'
' #E # 8 NUREG 75/094, D. IMPLEMENTATION The purpose of this section is to provide informa.
Regulatory Guide 1.70. Revision 2, September tion to applicants regarding the NRC staffs 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 1972.
proposes an acceptable alternative meth.od for com-f P
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1.97-4
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