ML19309C445
| ML19309C445 | |
| Person / Time | |
|---|---|
| Issue date: | 01/09/1980 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| Shared Package | |
| ML19309C446 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 ACRS-1688, NUDOCS 8004080615 | |
| Download: ML19309C445 (26) | |
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DATE ISSUED: 1/9/80 e/i / G
{g(- /b MINtfrES CF THE ACRE SUBCW.I'I' TEE MEETI!G CN REGULATORY ACTIVITIES NCVCiBER 7, 199 WASHINGTON, D.C.
- he AGS Subcommittee on Regulatory Activities held a meeting on Novenber 7, 1979, at 1717 H Street, N.W., Washington, D.C.
'Ihe purpose of this meeting was to review:
1.
proposed Regulatory Guides and Standards, 2.
revisions to existing Regulatory Guides and Standards, and 3.
other matters pertinent to activities that affect the current licensing process or reactor operations.
Notice of this meeting was published on 'Ibesday, October 23, 1979, in the Federal Recister, Volume 44, Number 208; a copy is included as Attachment A.
Mr. Sam D.raiswamy was the Designated Federal Dnployee for the meeting. A list of 'aeeting attendees is included as Attachment B.
GNJCTORY STATEMDfr BY '1HE CHAIRMAN Dr. Siess, the Subcomittee Chairman, convened the meeting at 8:45 a.m.,
reviewed briefly the schedule for the meeting, indicating that the Sub-comittee will hold discussions with the NRC Staff pertinent to the following:
1.
Regulatory Guide 1.141, Revision 1, " Containment Isolation Provisions for Fluid Systems.
2.
Proposed Regulatory Guide, " Qualification and Production Tests for Safety-Related Snubbers".
3.
Proposed Regulatory Guide 1.97, Revision 2, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident".
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Reg Act Mtg November 7, 1979 In addition, the Subccanittee will hear a presentation from the NRC Staff pertinent to the results of several tests conducted on fuel clad swelling in the United States (Oak Ridge and Argonne National Iaboratories) and in overseas (Gemany) and the impact of those results in the licensino process.
Dr. Siess noted that the Subcommittee had received written comments on Regulatory Guide 1.97, Revision 2, from the General Electric Company and the Babcock & Wilcox Company. he Subcommittee had also received requests from ANS 4.5 Working Group, General Electric Company and Westirshouse Owners' Group for time to make oral statements on Regulatory Guide 1.97, Revision 2.
IMPACT CF SCP.E FUEL CLAD TEST RESULTS CN '1HE LICENSING PROCESS Prior to holding discussions on the scheduled Regulatory Guides, the Subcommittee heard a presentation from the NRC Staff with regard to the recent findings frcan some fuel clad test results and the impact of those findings in the licensing process.
Mr. R. Meyer from the Division of Systern Safety of *he NRC stated that several tests have been conducted to gather data for use in the NRC Staff's evaluation of the Nuclear Steam Supply System (NSSS) Vendors Appendix K clad swell and rupture models.
He stat'ed that tests have been conducted to gather data pertinent to:
1.
Nrst temperature vs stress 2.
Nrst strain vs temperature 3.
Assembly flow blockage vs stress
'Ihe results of these tests showed the likelihood of more ruptures (larger rupture strains, and greater flow blockage) than those which were predicted pre-viously. Consequently, the NRC Staff thought that there might be a need to reevaluate all LOCA cladding models to assure that licensing analyses are performed in accordance with Appe-fix K.
Accordingly, the IRC Staff notified the licensees and the NSSS vendors of its, apparent concerns; they also issued a press release on this matter. On November 1-2, 1979, the NRC Staff met with the
Reg Act Mtg November 7, 1979 NSSS vendors to discuss this issue; the NSS vendors provided adequate explanations for the discrepencies between the test data and the pre-viously predicted data and convinced the NRC Staff that adequate con-servatism exists in their models and therefore the impact of these discrepencies is insignificant in terms of peak clad te reratures.
Dr. Sless asked whether the NRC Staff could assure, on the basis of its experience with the codes, that there would be a significant difference in peak clad tenperatures under the circumstances at which some of the test data points were observed to be out of range of tne predicted values of the NSSS vendor models.
1 Mr. Lauben respanded that they could do this for the flow blockage correlation; however, they may not be able to do so for the burst tempera-ture and burst strain correlations.
In response to another question from Dr. Siess with regard to the NRC Staff's ability to perform independent analyses to verify certain NSSS vendor models, Mr. Lauben stated that the NRC Staff has the capability to perform independent analyses. They believe that the models they use for such independent analyses are more conservative than the NSSS vendors models and, because of this addi-
'tional' conservatism, the NRC Staff's models are much more sensitive than the NSSS vendors models to certain parameters such as flow blockage.
Mr. Meyer discussed briefly some of the results of the recent test data (Attachment C).
A comparison of the recent test data with the Westinghouse and B&W data shows that their models are conservative.
In response to a question from Mr. Etherington as to at what stage the NRC Staff issues a press release on certain matters, Mr. Meyer stated that they make public announcements in accordance with certain regulations.
1 hey issued a press release of the apparent fuel clad swelling problem because Emergency Core Cooling System is a subject of public interest.
f Mr. Iauben added that office letter 19 of the ER requires that if the NRC 3taff identifies certain problens, they have to notify the appropriate peop2e.
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1 November 7, 1979
, Reg Act Mtg LATION PRCVISIONS KR_
RILULATORY GUIDE 1.141, REVISION 1, "CONTAIN ENT ISO r
1 was issued for public FLUID SYSTEMS" Mr. Scarbrough stated that Regulatory Guide 1.14the public cerinent i
i comment in April 1978, and all comments receive:! dur ngAt the Ja period of this Guide were resolved.
As a result of ide.
ACRS concurred in the Regulatory Position of this Gu"MI Less:ns Learned g
the tree Mile Island, Unit 2 (SI-2) accident, the imcact of degraded h
Task Force and other Offices of the NRC investigating t eisolation signal deriv l
core cooling on NRC regulations determined that an contaiment from contaiment pressure was not sufficiently reliable l
isolation when necessary.
tion of the public containment should be the primary concern in the protec As a result, they g
health and safety and therefore it should be monitored.
i i level decided that parameters such as containment pressure, rad at on.
tuation inside the containment, and engineered safe'ty feature systen actainm would provide sufficient diversity to ensure con ximum permissible h
prevent the release of radioactive material beyord t e ma dition. Con-limits under any abnormal occurrence or credible accident con 1.141 to provide sequently the NRC Staff has revised Regulatory Guide improved guidance.
i Guide:
Mr. Scarbrough sumarized the changes made to th s that the A new Regulatory Position added to this Guide recernmends l in design of control systems for automatic contaiment iso at o 1.
f containment signal will not result in the automatic reopening oI isolation valves.
ide adequate guidance in tive controls should be included to prov following the the reopening of any containment isolation valve reset of the isolation signal.
d d to Cne of the existing Regulatory Positions has been expan e i
include consideration of the contalment isolation a 2.
a parameters; it recomends that at least the following par -
f initiating noters should be monitored, with each capable o contalment isolation:
High containment pressure High radiation level within the containment a.
b.
o November 7,1979 Reg Act Mtg Any manual, automatic or coincident actuation of an c.
engineered safety feature system or engineered safety feature related subsystem.
A new Regulatory Position has been added to recer. mend that all 3.
non--ssential syste-s should be automatically isolated ty a containment isolation signal.
In response to a question frm Dr. Siess as to whether there are any plants in wnich non-essential systems are not isolated by an isolation signal, Mr. Scarbrough stated that it depends on how the Applicants define the essential and non-essential systems. He pointed out that a footnote has been added to this Guide to indicate that all plants should give care-ful reconsideration to the definition of essential and non-essential systems; should identify each essential and non-essential systems; should describe the basis for selection of each essential system; and should report the results of the reevaluation to the NRC Staff.
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With regard to the implementation section of this Guide, Dr. Siess comented that the implementation procedure of this Guide is not empatible with that outlined in the TiI-2 Lessons Learned Task Force short-term recommendations j
He indicated that NUREE-0578 states that, for operating plants,
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(NUREG-0578).
the recommendation pertinent to the containment isolatien issue should be
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implemented prior to 1980; for all plants in construction permit review and all plants under construction for which an operating license application has not yet been tendered, the recommendation on containment isolation should be incorporated into the plant design as appropriate prior to receipt of an operating license. However,the implementation section of this Guide indicates 4
that the procedures outlined in this Guide shall not be used in the evaluation of applications docketed prior to a certain date; it implies also that it may He believes that not apply to certain applications currently under review.
this Guide should be applied to all the operating plants as well as plants under construction.
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.-c November 7,1979 Reg Act Mtg Mr. Scarbrough responded that the implementation status of this Guide will be discussed at the Regulatorf Requirement Review Committee meeting, and appropriate revisions will be made to make it consistent with the recccrendations delineated in NURH3-0578.
I With regard to Regulatory Position C.2 which indicates that sealed closed isolation valves that are under ' administrative control do not require position indicators for valve status in the control recrn, Dr. Zudans He believes that it ccanented that this may not be an adequate practice.
would be a good practice to provide annunciation in the control rom for Indicating that ell those systems that actually penetrate the containment.
this seems to be a generic problem, Dr. Siess suggested that the ?RC Staff look at this issue in terms of probabilities of cmmon mistakes, contribution l
of human error and the associated consequences.
l Af ter further discussion, the Subccanittee suggested some editorial changes, provided suggestion for clarification in certain areas of this Guide and indicated that it would recorxend this Guide to the full Ccanittee for concurrence with the Regulatory Position of this Guide during the 235th ACRS meeting with the condition that the implementation section of this Guide should be revised censistent with the W.I-2 I.essons Learned Task Force short-term reccanendations.
PROPOSED REGUIATORY GUIDE, " QUALIFICATION AND ACCEPTANCE TESTS FOR S REIATED SNUBBERS" 1his proposed Regulatory Guide delineates construction and test methods acceptable to the NRC Staff for design qualification and testing of safety-related snubbers for use in nuclear power plants.
Dr. Siess commented that this Guide is not well organized; the purpose of this Guide is not clear; certain parameters are defined inconsistently in i
In view of the fact that most of the snubber failures different places.
are not due to deficiencies in the snubber design but due to the occurrence of certain incidents een the snubbers are in use, he wondered how the test g 9 epe me e
November 7, 1979 Reg Act Mtg procedures delineated in this Guide would help alleviate the snubber ld reduce inservice y
. failures. He does not believe also that this Guide wou inspection as stated in paragraph 3.2.2.7 of the Value-Impact Analysis In view of the fact that some of the snubber section of this Guide.
failures are attributable to incroper installation, he believes that h
an inservice inspection program will still be needed to detect suc installation deficiencies.
d Mr. Shaw, NRC Staff, agreed with Dr. Siess that this Guide may not re uce inservice inspection.
i Indicating that the NRC Staff has technical specifications for inserv ce inspection of hydraulic snubbers, Dr. Siess asked for justification of not having such specifications for mechanical snubbers.
justification Mr. Shaw responded that he may not be able to provide adequate
- However, for not having technical specifications for mechanical snubbers.
he believes that technical specifications for hydraulic snubbers were developed to minimize the frequent failures of these snubbers due to the He pointed out that the NRC Staff has revised leakage of hydraulic fluid.
irernents the existing technical specifications to include surveillance requ for mechanical snubbers and these revised specifications will be imp in the near future.
i l Dr. Siess stated that, in view of the fact that the failure of mechan ca snubbers may be due to their exposure to environment, the qualification y help test procedures outlined in this Guide for mechanical snubbers ma reduce the failure of these snubbers due to certain enviror is However, he believes that inservice inspection of mechanical snubbers l
necessary to detect the deterioration of these snubbers due to envirorrne factors and also to help the manufacturers to change their design pro accordingly so as to reduce such problems.
ACRS Dr. Siess pointed out further that the results of a sttdy made for the indicate that the (in connection with the snubbers at Diablo Canyon Plant) during a seismic probability and consequences of a snubber failing to lockup i
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. Novemoer 7, 1979 Reg Act Mtg event are low; however, the probability and consequences of locking up during normal plant operation, were much greater and such failures t.ould result in overicading and damaging the piping systems.
Mr. Shaw stated that he agrees with Dr. Siess that failure of snutbers due to lockuo will result in overstressing and damagina the piping system. Recognizing this problem, the NRC Staff has included a require-ment in this Guide that overload protection should be provided to protect the piping system.
Indicating that seismic analysis of the piping systems have been performed recently using " linear modeling" of snubbers instead of " rigid modelirg",
Dr. Siess asked whether there is some ancillary motive -of performing seismic analysis of the piping systems using "non-linear modeling" to account for the non-linear elements associated with the snubbers.
Mr. Shaw responded that the NRC Staff also has some questions with regard to the feasibility of using " linear modeling" of snubbers in the seismic analysis of the piping system. Wey are issuing some contract work to study this issue using time history analysis to determine how the linear approximation can be improved to account for the non-linear behavior of the snubbers.
Dr. Siess stated that he believed it would be helpful to perform a linear analysis by varying ene spring constants over a considerable range; vary-ing the spring constants by a significant amount might give some fesl for the sensitivity. Dr. Siess cwmented also that he believed that the analyses are getting more and more complex; there seems to be a general tendency among engineers to believe that more complev an analysis is, the more accurate it is, and he thinks that such philosopny is not generally true.
Mr. Shaw pointed out that it is not timir intention to make the analysis more cenplex.
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. November 7, 1979 Reg Act Mtg In response to a question from Dr. Siess as to how many snubbers a
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manufacturer has to test off the production line after conducting a qualification test, Mr. Shaw stated that all snubbers cening off of the production line should be tested. He pointed but that most of the snubber manufacturers have already been following this procedure.
I Dr. Siess ccanented that the NRC Staff's requirenent of 100-percent production testing seems to be inconsistent with the requirement of Regulatory Position C.4 which states that the qualification and acceptance tests loading requirements for those snubbers with a rated capacity greater than 150,000 pounds will be determined on a case-by-case basis. He asked the reasons for not requiring 100-percent produc-150,000 tion testing of those snubbers with a rated capacity greater than pounds.
The NRC Staf f stated that their intention is to require that all snubbers coming off of the production line, irrespective of their capacities, should be tested. However, the test procedures may be determined on a case-by-case basis. W ey agreed with Dr. Siess that there is an inconsistency. Wey indicated that they will make appro-priate changes to preclude the inconsistency.
After further discussion, the Subecmmittee suggested several editorial changes, provided suggestions for clarification and improvtsnent in certain areas of this Guide and indicated that the NRC Staff could issue this Guide for public comment. Dr. Siess reiterated his earlier comments that this Guide is not well organized and it needs a certain ancunt of cleaning up prior to its issuance for public ccament.
PROPOSED REGUIATORY GUIDE 1.97, REVISION 2, "INSTRUMDrTATION FOR LIGIC-MTt.x-COOLED NUCLEAR POWER PINfr TO ASSESS PIANT AND ENVIRCNS CCNDITI DURING AND FOLLOWING AN ACCIDDff" Mr. Hintze stated that the tRC Staff's main intent is to obtain comments c
and suggestions from the Regulatory Activities Subcommittee on the philo-sophy and approach used in Regulatory Guide 1.97, Revision 2 and also to l
obtain the Subconnittee's concurrence in subnitting this Guide for public i
Comunent.
Ex
r 7,1979 w-Reg Act Mtg Mr. Hintze reviewed briefly the history and development of Regulatory Guide 1.97, indicating that the initial development of this Guide which (BMI-X-647) frem Battelle S e initial draft version of this Guide contained f
began in July 1973 was based on a report
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Memorial Institute.
an extensive list of parameters to be considered in accident-monitcring Representatives of several industries raised strong instrumentations.
objections to the requirments delineated in this Guide, indicating that Upon considera-it did not provide specific guidance to the industry.
tion of these objections, the Guide was revised and issued for public I
Subsequent to the public ccmment period, ccament in December 1975.
another revision was made to this Guide to reflect consideration of Ibring this revision, a new Regulatory Position several public comments.
was also added to address a specific ACRS concern that certain accident-g annitoring instruments (Pressure, Temperature, Radiation level) should l
Regulatorf Guide 1.97, have ranges which go beyond Class 8 accidents.
Ibwever, the Revision 1 was finally made effective in August 1977.
1 B e applicants impimentation of this Guide proved to be difficult.
i were reluctant to impleent this Guide because they felt that more de-finitive guidance was needed to define acceptable means of ccepliarce.
The applicants also raised very strong objections to the requirment of Regulatory Position C.3 wttich gave the implication that consideration should be given to Class 9 accidents den selectirq accident-tronitoring instrumentation; the applicants raised this objection on the'hssis that consideration of Class 9 accidents is not required by regulations.
As a result of these objections and other concerns about the impleentation Guide 1.97, the NRC Staff included this issue in its generic of Regulator /
A Task Action Plant (A-34) was initiated to develop guidance items list.
Prior for applicants and the NRC Staff reviewers in implementing this Guide.
to finalizirg the Task Action Plan Report, the need for a revision to this t
l Guide arose to address the lessons learned frcm the "MI-2 accident.
Subsequent to the 7MI-2 accident, the American Nuclear Society (ANS) established Workirq Group 4.5 in July 1979 to prepare a draft Standard L
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Reg Act Mtg November 7,1979 I
on accident-monitoring instrumentation with the understanding that Regulatory Guide 1.97, Revision 2 would endorse this Standard as appropriate.
"'he ANS 4.5 Standard defines four classificati:ns (?/ces A through D) of variables to aid the designer in his selection of accident-monitoring instrumentation and applicable criteria. Se NRC Staff has added a fif th classification (Type E) in Regulatory Guide 1.97, Revision 2 for those variables to be monitored as required to provide defense-in-depth and for other useful purposes. Revision 2 to Regulatory Guide 1.97 also contains a list of variables for both PWRs and BWRs; these variables l
were developed based on the information provided in the Craft Task Action Plan (A-34) report and in accord with the TMI-2 Lessons Learned Task Force recomedations.
i Mr. Hintze pointed out that there are some differing technical opiniens between the NRC Staff and the ANS 4.5 Working Group.
Based on its review i
of the ANS 4.5 Standard, the EC Staff believes that the Standard should include the following information:
I 1.
W e ANS 4.5 Standard should address not only the monitoring concerns of the centrol room operator but also those of the plant operator (licensee); consideration should be given to the additional requirements of variables (e.g., emergency planning) to be monitored by the plant operator during and following an accident (see Regulatory Position C.1).
2.
In conjunction with the parameters that indicate the potential for a breach in the containment, the para-meters that have the potential for causing the breach in the fuel cladding and the reactor coolant pressure boundary should al-:, be included (see Regulatory Posi-tion C.2).
e November 7,1979
. Reg Act Mtg In conjunction with the design basis accident events 3.
delineated in the Standard, those events which are e
expected t.o occur one or more times during the life of a plant should be included (see Regulatory Position C.3).
g Since the NRC Staff believes that Types D and E variables 4.
and instruments are witnin the scope of accident monitor-ing instrumentation, they should be included in the Standard (see Regulatory Position C.4).
In addition to the list of variables for Types A, B and C, 5.
a list to cover Types D and E should also be included in the Standard.
Paragraph 6.1.2 of the ANS 4.5 Standard states that 6.
Ihase II instrumentation shall be qualified to function i
for not less than 100 days unless other times can be g
Ibwever, in light of the 'IMI-2 accident, the justified.
m C Staff believes that Phase II instrumentation should be qualified to function for not less than 200 days miess j
a lesser time can be justified.
Af ter Mr. Hintze's presentation, Mr. Wen:inger stated that the NRC Staff had received several comments from the ACRS and its consultants (
ment D); they have also received some comments through ACRS from the General Electric Company and Babcock and Wilcox Company; since the NRC Staff's main aim is to obtain concurrence of the Regulatory Activities Subconunittee in issuing Regulatory Guide 1.97, Revision 2 for public cocinent, the NRC Staff would prefer to resolve these consnents along with other public ccanents that may be received during the public c:xnment He also indicated that this Guide was written period of the Guide.
originally for forward-fitting purpose only. However, during the public cocinent period, the EC Staf f intends to work with the Applicants to obtain input for backfitting recommendations.
. Novenber 7,1979 Reg Act Mtg Thd Subccanittee did not raise any objection to the NRC Staff's proposal for resolving the ACRS and its consultants' comments along with other public co ments.
The Scoco :mittee heard the following statements from representatives cf ef several industries:
Statement By The General Electric Comany Mr. Tscheche representing the General Electric Ccunpany (GE) expressed concern about the requirenent of incere tamperature monitoring in the BWR core. He stated that GE believes that the BWR incore temoerature monitoring may not provide any use-ful infonnation to the control room operator to identify localized hot areas in the core.
In view of the fact that all BWRs have redundant safety-grade water level indicators along with containment and effluent radiation monitors to provide information to the operator to determine the water level in the core and to assess the core damage, GE believes that the incore temperature monitoring may provide some confusing in-fonnation to the operator.
Indicating that he is not convinced of the argument made by Mr. Tscheche, Mr. Bender asked what kind of information is available to aid the operator in assessing the core damage.
Mr. Tscheche stated that the containment and effluent radiation monitors would provide adequate information to assess the core damage.
Mr. Bender commented that he does not believe that these uenitors wuld provide adequate and reliable information for use in determining the status of the core.
i Dr. Siess commented that, in view of the fact that the operator has to do some sort of interpretation of the information provided by the water level instrumentation and radiation monitors to determine the core danage, the
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Reg Act Mtg Novmber 7, 1979 present arrangement does not seem to resolve the real concern as to wnether there are any other means available to provide information that could be used directly without any further interpretation to assess the core damage.
l Mr. Michelson conmented that it does not seem appropriate to depend only on one variable (such as ater level indication) or one method of measurement; he believes that there should be diverse methods of measure-ments to obtain other appropriate variables for use in assessing the core damage.
In respmse to a question from Dr. Siess as to whether there are still any major objections from the industry to the requirement that considera-tion should be given to Class 9 accidents while selecting accident-monitoring instrumentation, Mr. Benaroya, NRC Staf f, stated that it is his understanding that the industry does not raise any more major objec-tions on this issue.
In response to another question from Dr. Si-ss as to whether the NRC Staff has given consideration to the philosophy that instruments should be provided to indicate whether the accident is progressing in accordance with the prediction, Mr. Wenzinger stated that it has been factored into the developnent of Regulatory Guide 1.97, Revision 2.
Statement By The Westinghouse Owners Group Mr. Stern (Wisconsin Public Service Corporation) representing the Westing-house Owners Group expressed concern about the fact that the NRC Staff did not hold any discussions with the Westinghuse owners Group to obtain their opinions on the selection of post-accident monitorirg instrumentations.
As operators of Westinghouse plants, they have learned several lessons on the type of instruments to be employed for post-accident monitoring. He does not believe that to date either Regulatory Guide 1.97 or the ANS 4.5 Standard has incorporated fully the thoughts and concerns of the Westinghouse owners Group. With their operating experience, they believe that they wrald be able to provide significant contribution to the development of Regulstory Guide 1.97.
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reovemoer 7,1979
-n-Reg Act Mtg Mr. Benaroya responded to the concerns expressed by Mr. Stern, indicating that the )RC Staf f intends to meet with all the Owners Group to ootain their cements af ter getting the concurrence of the ACRS in the philo-sophy of Regulatory Guide 1.97; on several occasions; they have informed the Atonic Industrial Forum, the ANS and appropriate Nuclear Stean Scpply System vendors that the NRC Staff will hold discussions with the nuclear industry on the post-accident monitoring instrumentation issue prior to the implementation of Regulatory Guide 1.97, Revision 2.
Statement By The ANS 4.5 Working Group Mr. Polanski (Comrtenwealth Edison Company) representing ANS 4.5 Working Group stated that ANS 4 established ANS 4.5 Working Group in late July 1979 to prepare a draft Standard on accident-r:enitoring instrtunentation.
ANS 4.5 Standard Provides the following:
1.
Essential instruments necessary to characterize the status of the plant during an accident.
2.
A list of functions to be performed.
3.
A framework to determine those variables to be monitored.
4.
An identification of three time periods of interest.
5.
An identification of four variable types.
6.
A delineation of applicable design criteria for the variables to be monitored.
Mr. Polanski pointed out that an approved ANS 4.5 Standard is expected to be issued in April or May of 1980.
In respsonse to a question from Mr. Michelson, Mr. Polanski stated that the ANS 4.5 Star.dard presents criteria for monitoring the response of the plant to design bas:s events.
It also presents criteria for monitoring the integrity of fission product barriers mder conditions Wich have degraded beyond design bases.
Mr. Polanski stated that the ANS 4.5 Working Group does not believe that Regulatory Guide 1.97, Revision 2, in its present form, does an adequate l-job with regard to accident-monitoring instrtunentation for the following l
reasons:
c
Reg Act Mtg November 7,1979 1.
A systematic approach to design is not followed in Regulatory Guide 1.97, Revision 2.
2.
Se variables (to be monitored during and following an I
accident) listed in Tables 2 and 3 of this Guide do not seem to be derived from accident-monitoring instrumentation criteria.
3.
A comparison of the variables listed in this Guide and in the ANS 4.5 Standard (Attachment E) indicates that Regulatory Guide 1.97 requires several instruments that are not necessary to accomplish the function.
4.
ANS 4.5 Working Group believes that Type D variables and instruments are not considered to be within the scope of accident monitoring instrumentation therefore, these variables should be addressed in the Safety System Standards (IEEE-603).
5.
We definition of Type E variables in this Guide is ambiguous, and the purpose and contribution of these variables to the accident-monitoring instrumentation is not explicit.
6.
While endorsing the ANS 4.5 Standard, this Guide ignores several main issues identified in the Standard and supplanents those with things that are not important.
Mr. Wenzinger pointed out that the question of includity Type D variables and instruments in IEEE-603 Standard was discussed by the IEEE-603 Group and they decided not to include them in the IEEE-603 Standard.
We Subcommittee continued es review of Regulatory Guide 1.97, Revision 2 after hearing the preceding statenents.
Mr. Bender provided the following comments (Attachment D):
1.
te requirements of Regulatory Guide 1.97, Revision 2 are too pervasive and the purpose of the accident-monitoring
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Reg Act Mtg November 7, 1979 instrumentation appears too similar to that of nonnal plant instrumentation. te WI-2 experience, which showed that a few carefully selected instruments were all that was required to respnd to the emergency, should be factored into the development of this Guide.
2.
Bis Guide should differentiate between instnmntation needed promptly at the time of an accident anc that which can be provided later, if needed, on a standby basis.
3.
It is not appropriate to specify accident instrumentation that has performance capability unique to a specific accident miess that accident has a high probability of impasing the performance demands.
4.
It is important that this Guide not become a set of requirements covering the monitoring of every minor acci-dent. te instruments of interest are those to help the operator in energencies een he needs real help in diagnos-ing an accident on a timely basis.
It is advisable to perform some accident analysis to develop a set of bounding accidents to be monitored.
5.
A corscious decision needs to be made as to where to draw the line between instrumentaiton that could routinely monitor an as cident and that which is specifically intended to follow the accident over a specified period of time as a basis for opera-tional guidance and energency response.
6.
It is important that emergency monitoring information be related to symptoms of the accident. Routine monitoring might be im-partant subsequent to the accident to show the condition of equipnent needed durirg accident recovery.
7.
te logic for swirornental qualification ought to be a require-ment in this Guide.
8.
In establishing the qualifications for instrunentation, care should be taken to avoid imposing requirements that result in a highly sophisticated measuring device to satisfy circunstances that have low probability of occurring simultaneously with the type of event to be monitored.
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Reg Act Mtg November 7, 1979 In response to one of Mr. Bender's carmnents (comment 2) that the Guide should differentiate between instruments that ought to be installed per-manently and those which can be installed later, Mr. Hintze stated that this issue has been addressed in ANS 4.5 Standard. Regulatory Guide 1.97, Revision 2 requires that all instruments with the exception of portable ones should be installed permanently.
Mr. Wenzinger stated that he agrees with Mr. Bender's philosophy; however, without knowing the nature of the future accidents it is not pssible to determine the type of instruments that can be installed at a later date.
With regard to the cownt on environmental qualifications (ccrn:nent 7), the NRC Staff stated that adequate information for equipnent qJalification is provided in ANS 4.5 Standard, IEEE-323-1974, IEEE-344-1975, and Regulatory Guide 1.89.
Indicating that neither Regulatory Guide 1.97, Revision 2 nor the ANS 4.5 Standard provides adequate informa'. ion on the equipnent necessary for trend analysis, Mr. Bender asked how tha 14tC Staff will determine the adequacy of the trend-analysis equipment proposed by the Applicants.
Mr. Wen:inger responded that this issue is not within the scope of Regulatory Guide 1.97.
He believes that detailed information on this issue may be addressed in a separate Regulatory Guide or a Standard.
Dr. Catton ecenmented that he believes that a discussion of factors relating to how an operator is going to receive the results of various measurenents would be helpful. Such a discussion should include the process ccanputer and any requirements that should be imposed on it as wil as location of instru-ment displays.
Dr. Zudans commented that appropriate instruments or some sort of a diagnostic system should be provided to assist the operator.
Mr. Michelson ecznmented that the fact that the seismic classifications of equipnent for accident conditions are not necessarily the design requirenents
~-"
p e-November 7,1979 Reg Act Mtg for normal operating conditions is not made clear in this Guide; he suggested that some additional clarification would be helpful.
Be NRC Staff stated that they will provide additional clarification.
'Vith regard to the list of instruments provided under Type D, Mr. Michelson commented that some guidance or additional clarification, to point out that some of the instrtzneatation listed under Type D may also be listed under Type A, would be helpful. Dr. Siess stx3gested that the NRC Staff should think about providing a separate list for Type A instruments to avoid this sort of confusion.
Indicating that certain events that originate outside of the primary containment (auxiliary building, control room, etc.) may eventually lead to core damage, Mr. Michelson suggested that consideration should be given to monitor arx' follow the course of such events that originate outside the contairunent.
Mr. Benaroya responded that the issue of tracking the course of an accident that originates outside the containment has been discussed in Regulatory Guide 1.97, Revision 2.
Mr. Michelson stated that it is not made clear in this Guide and stx3gested that additional clarification would be helpful.
With regard to the flow meter range (-20% to +20%) specified for measuring the reactor coolant loop flow (Table 2 of Regulatory Guide 1.97, Revision 2), Mr. Etherington consnented that a real range of interest would be around 1% or even less. A range as high as specified in this Guide may lead to insufficient accuracy in the natural circulation range.
i Re NRC Staff stated that they will look into this issue.
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Reg Act Mtg Novenber 7,1979 With regard to the variables pertinent to the reactor coolant system (Table 2 of Regulatory Guide 1.97, Revision 2), Mr. Etherington suggested that the NRC Staff should also consider including tesperature as one of the variables to be monitored.
Indicating that some safety valves other than power operated relief valves may not perform their intended function'during certain transients, Mr. Michelson suggested that the NRC Staff should give some thought to this issue and decide whether monitoring of such safety valves are essential.
After further discussion, the Subcommittee and its consultants suggested several other editorial changes, provided suggestions for clarification and improvenent in several areas of this Guide, and indicated that it will recomend to the ACRS full Committee that this Guide be issued for public coment. Ebwever, the NRC Staff should discuss the philosophy and technical contents of this Guide with the full Committee on November 9,1979.
Subject to the concurrence of the ACRS full Comittee, the NRC Staff could issue this Guide for public coment.
We Subcommittee also suggested that a representative of ANS 4.5 Working Group should also give a brief presentation to the full Comittee on November 9, 1979 so as to enable the ACRS to have a clear perspective of the concerns of ANS 4.5 Working Group.
We meeting was adjourned at 12:48 p.m.
1 NOTE: For additional details, a complete transcript of the meeting is available in the EC Public Document Roczn,1717 H St., N.W.,
Washington, D.C. 20555, or from Ace-Federal Reporters, Inc.,
444 North Capital Street, N.W., Washington, D.C.
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sb su2o Federal Resister / Val 44. No. 20s / Tuesday. October :2.1s79 / Notices l
signed c weAnyisen. DC. this tadi day Adyteory Committee on Reactec. / Such commaants shall be based upon a
af October. as*
Safeguards, Sat > committee en V
docmnents on fue and ava2iable for F
lam D. I. ansa.
R*gulatory Activttles; Meeting inspectnan at the NRC Public
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y Admanmrcar. Prnsa and wehre Benept b ACKS Subm-mittee on y a,,
t progrooms, Lohar#anures sat Semcas Regulatory Activities wiD hold an open Fwbr htim ard% to a Adsrausnman Dryanmset s/Idat meeting on November 7.1979. In Room to be discussed. wheth the meeting asom unnerawu n am-s 1157,1717 H SL NW, Washingtan. DC has W nen aUed n M led, b
- cosaareaa 20555. Notice of this meeting was Chammu's mig a ge b 6e published in the Federal Ragiatae opportunity to present oral statements IA and the time allotted therefor can be gg g p,,,g obtained by a prepaid telephone call to NUC1. EAR REOULATM outlined in the Federal Registar on de Dnignskd %ral Employn for COMMIS$ TOM Odober 1.1s79 (44 FR 56408) oral or this meeting. Mr. Sam Duraiswamy.
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"'" tad by (telephone 202/63+-3:S?) btNeen &15 Advicory Cormadttee on Reector
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g Safeguarda, Procedurea and e.m. and 500 pm EDT before, ad T t'
Adsrurdstretloa d-nmittoo;Meeer[ d 8
afte. October OL in e
e se ha ing Dad October te 1sm
% ACRS Procedures and kept, and quescona may be askad only Administraton Subcommittee win hold by members of the Suhm=mittaa,its John C Heyde, an open macong on.Nov==h-7. tes an consultan,ta. and Staff. Persons desiring Adrasury Caavance Momsernet ONee.
Room 1010.1717 H St, NW. Wasinnston. to male cred statamania shodd nonly yaom.mm, raw w aas i t.
DC 20555.
the Designated Federal Employra as far saa. sus oces mes ewe In accordance with the prx.h in advance as practicabla no that appropriate arrangemacis can be made outlined in the Federal Register on October 1.1s 9. [44 FTL 56408). oral or to atow the necessary tune daag ee W MWmmhe wntten statements may be presented by meeting b sah saakments 88W bdnG
& agenda 6 subject meeting M In scoordance with the purposes of members of the public. recordings wiu be as foUows. Wednesday. November 7 Secamens s and 182b. of the Atocuc be permitted only during those portions in & medng wE comune e et H5 Esargy Act (42 U.SC 2039. 2:30 b.);the of the meeting when a transcript is bems Advisery Coeucattee on Reactor am.
kept. and questaans may be asked only h Submnuminee will bear-Safeguards will hold a meetmg on by members of the Subcommittee. Ita presentations frotn the NRC Staff and Newmber 6-10.19 9. in Room 1046.1 17 consultants, and Staff. Persons destring wdl bold d2scussions with thus group H Street. NW. Washington, DC. Nouco to make oral seekments should nody the Designated Federal Employee as far Pertment to the foBowing-of this meeting was published on (1) Proposed Regulatory Guide.
September 20.1979 (44 FR 54558).
in advance as practicable so that "Q:.iificauon and Production Tests for N egende for the eubject meetmg appropnate arrangemens can be made S-fety Related Snabbers, wtB be as foDowe-to allow the necessary tune dunng the nuroday.NeA 4.12m I
meetma for such esatements.
R on s
r The agenda for subsect meeting shall Water-Cooled Noclear Power Piants to a:30 cm.-n:30 pm.: Executive be as fouowe %ednesday !sovember 7 Assen Plant and Environe Cocha Sew (OpenFh htta win 19 9' 115 pan. anni doe onociosion of Dunna and Following an Amd=nt "
heat and discuss the report of the ACRS i
buamaan.
(3) Regulatory Guide 1.141. Reviaim 1.
Chaarman regarding miscellaneces
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& Subcomaattne wiD discuss the Containment Isolation Prov1auma fue matters relatmg to ACRS activites, empe and procedsve for conduct of Fluid Systems."
h f%mmuttae will discuss proposed i
ACRS busmesa.Inc!odmg ACRS Other matters which may be of a ACRS mmments and recommendations
,j involveanent in conswierattac of diffarms predecisional natare relevans to reactor regard ng the NRC regulatory process.
prde**'onal mmma ammg NRC Staf operation of licensing activities may be A30pm.-4Ep au Diob/o Canyort e
S membert discussed following this assaica.
Nuclear PowerStation Units I and2 Further information regarding topics Persons wishing to subaut written (OpenFh Committee will haar and to be discussed, whethat the meetmg statements regardmg Regulatary Gaide discmu reports from representatives of has been canceUed or rescheduled. the 1.141. Revtsion 1. may do so by NRC Staff, and the Pacific Ces and
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Chairs:an s ralms on requesta for the prowling a readily reproducible copy to Electric Company and their consultants /
opportunary to present cral statements the Subcommittee at the begmmns of contractora, as necessary. regardmg 1
i and the t2me allotted therefoe can be tbs meeting. However, to insure that proposed app!! cation of experience obtamed by a prepaid telephone caH to adequate time is available for full gained at the aree Mile Island Nuclear tg the Designated Federal Employee for consideration of thew comments at the Station Unit 2 to the Diablo Canyon j
l this meeting. Mr. Raymond F. Fraley.
necting. It la desirable to send a readily Nuclear Power Station.
(telephone :0:/634-3:55) between tis reproduoble copy of the comments as Portions of this session wiD be closed a m. and $110 pan, FDT bafore, and EF far in advance of the meeting as sa necewary to discusa Proprietary after. October :n.1973.
praeticable to Mr. Sam Duralswamy Information applicable to this matter.
(ACRS), the Designsted Federal 4:Ep.m.-d2p.m.: Weetinghouse Employee for the meeting, to care of Nuclear Steam Supply System with ice.
Dated. N E ism W ** A
ACRS, Nuclenr Regulstacy Commisalaa.
Cwdenser Containment /Sequoyah.
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A tnnry Cunnarar AAwesesens Omasc.
Waahington. D C. 20553 or telempy McGuire Nuclear Planis (Open)-b i
w i
tru oman n-a=== a= =4 them to the Designated Federal Committee wi!! bear and discuss reports
- coes ren e ws Employee (222)-634-33t9) as far hi from representativas of the NRC Staff.
advance d the meetmg as practicable.
and the Applicants and thef!r consultants i
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ArFMcw4E nr R Ot G A M..
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' REGULATORY ACTIVITIES SUBCOMMITTEE MEETING NOVEMBER 7, 1979 WASHINGTON, D.C.
I' i
ATTENDEES LIST _
t NRC g.
AC,RS, D. Powers, NRR/ DSS.CPB C. P. Siess, Chairman G. N. Lauben. NRR/ DSS /AB H. Etherington, Member R. O. Meyer, NRR/ DSS /CPB J. Ray, Member H. K. Shaw, NRR/ DOR /EB M. Bender, Member E. L. Hill, SD/PSB C. Michelson, Consultant R. DiSalvo, RES Z. Zudans, Consultant G. F. Guppy, RSSB I. Catton, Consultant E. C. Wenzinger, RSSB S. Duraiswamy, Staff
- A. S. Hintze, RSSB
- 6. Young, Fellow P. Studdart, DSE J. D. Buchanan, EPSB
- Designated Federal Employee V. Benaroya, NRR/ASB R. Auluck, OSD J. Norberg, OSD AIF_
W. Anderson, OSD W. Morrison, 0S0 P. Higgins BECHTEL POWER CORP VICTOREEN, INC_
C. R. Wienke K. E. Stafford SCIENCE APPLICATIONS, INC ANS 4.5/ COMMONWEALTH EDISOff J. L. Vonherrmann X. Polanski EG&G IDAHO CONSUMERS POWER CO.
M. E. Stewart D. A. Sonsners PRODUCT ASSURANCES SNUPPS (CONSULTANT)
H. Schock H. W. Graves, Jr.
J. C. Mesmeruger T. F. Timons HARVARD GE W - 0FFSHORE POWER R. Wilson
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TtSTEMS A. N. Tschaeche
.E_P RI, P
R. A. Touchton NUTECH R. Leysee WISCONSIN PC9 TIC M. M. Miynczak SERVICE - W Oi!NEPS
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