ML20044B416
| ML20044B416 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 09/13/1990 |
| From: | Murley T Office of Nuclear Reactor Regulation |
| To: | Beck G BWR OWNERS GROUP |
| Shared Package | |
| ML17056C371 | List:
|
| References | |
| CON-IIT07-463A-91, CON-IIT7-461-91, CON-IIT7-463A-91 NUREG-1455, NUDOCS 9009260079 | |
| Download: ML20044B416 (1) | |
Text
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- SEP 1 s 1g93
,, George J. Beck, Chairman
(
)
- (,1 Owners' Group
'rniladelphia Electric Company 955 65 Chesterbrook Boulevarc Wayne, Pennsylvania 19087-5691
Dear Mr. Beck:
I am responding to your letter of August 16, 1990 in which you appeal the NRC staff's position on post. accident neutron flux monitoring systems (NFMS) for boilingwaterreactors(BWRs). This appeal requests reversal of the staff's position that licensees install NFMS instrumentation that meets the Category I criteria of Regulatory Guide 1.97.
I have asked sly staff to evaluate your appeal and present the advantages'and I will notify you of igy decision. To facilitate the disadvantages to me.
rsview, the staff will be contacting you to arrange a meeting to discuss the details of your appeal.
The appeal elso requests that licensees be allowed to defer plant-specific actions to install NFMS pending resolution of the appeal.
I will make a decision on this appeal within the next two months, and therefore, licensees may defer plant. specific actions until that time.
you have any questions regarding this information, please contact Scott berry of my staff on (301) 492 0821.
Sincerely, Original signed by, Thomas I. Marley Thomas E. Murley, Director Office of Nuclear Reactor Regulation DISTRIBUTION:
Central Files SICB RF (2)
R. Jones J. Hannon J. Joyce L. Phillips G. Holahan S. Newberry C. Abbate H. Richings W. Russell A. Thadani B. Marcus PDR T. Murley F. Miraglia NRR Mail Room (YEL 0909185)
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- cor. in 59/11/90 9 00cidb LLgQgg OftlCIAL RECORD COPY Document Name: NEUTRON FLUX APPEAL
I) i b OulNERS' GROUP
"*"" *Wi=3;n clo COMNONLD(ALTH EDtSON COMPANY e Am. 34 FN East e RO. Box 767 e Chky, K 60690 BWROG-8817/BWR1 April 1, 1988 i'
l' U.S. Nuclear Regulatory Commission Division of BWR Licensing Washington, D.C.
20555 Attention:
T.E. Murley, Director NRR BUR OWNERS' CROUP LICENSING TOPICAL REPORT " POSITION ON NRC
SUBJECT:
REGU1ATORY CUIDE 1.97, REVISION 3 REQUIREMENTS FOR POST-ACCIDENT NEUTRON MONITORING SYSTEM" (GENERAL ELECTRIC REPORT NEDO 31558 Centlemen:
The BVR Owners' Croup (BUROC) has completed its analysis of requirements The results are for BVR Post-Accident Neutron Monitoring instrumentation.
QQ used to establish appropriate neutron monitoring post-accident functional Deviations from RG 1.97 requirements are justified.
design criteria.
The BVROG approach and overall strategy has been previously discussed with the NRC (Joseph P.
Joyce and others) at a meeting in Bethesda on January 27, 1988.
At this meeting the NRC indicated their willingness to evaluate alternate approaches te RG 1.97 neutron monitoring system requirements to resolve this issue.
This letter submits 30 copies of the subject Licensing Topical Report for NRC review and approval.
It is requested that this report receive priority review since several BVROG members have near term license actions associated with resolving the RG 1.97 requirements for post-accident neutron monitoring.
In the interim, the BWR Owners' Group requests that all post-accident NMS implementation requirements be deferred until this The total installed report has been carefully evaluated by NRC reviewers.
cost for an upgraded or independent post-accident NMS ranges from $1 to $5 million depending upon options selected (total replacement of SRM/IRM subsystem or addition of equipment specifically for post-accident use).
Due to the high cost, installation dose significance, and cost benefit uncertainty associated with this plant modification, we do not recommend that installation decisions should be required prior to resolution of this Licensing Topical Report.
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BVROG-8817/LUR1 April 1. 1968 Page 2 This Licensing Topical Report has been endorsed by a substantial number of the members of the BVR Owners' Group; however, it should not be interpreted as a commitment of any individual member to a specific course of action.
Each member must formally endorse the BVROG position in order for that position to become the member's position.
Very truly yours,
_ _ _= n _
Robert F. Janeceh, Chairman BVR Ovners' Group
/ta Ot Attachment cc:
BVR Owners' Group Primary Representatives BVR Ovners' Group Executive Overview Committee M.V. Hodges (NRC)
J.P. Joyce (NRC)
R. Evans (NUMARC)
U.S. Green (INPO)
H. Wyckoff (EPRI)
D.N. Grace (BVROG Vice Chairman)
L.S. Gifford (GE, Bethesda)
.. = _ _ _ _. _.. _. _.
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NED0 31558-Class I.
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P POSITION ON NRC REGULATORY GUIDE 1.97, REVISION 3 l
REQUIREMENTS FOR POST ACCIDENT j
NEUTRON MONITORING SYSTEM O
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MARCH 14, 1988 O
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OtBCLAIMER OF RESPONSIBILFTY
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Generet SecMc Company not any of she contr@utors k akin document.
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hiehes any warranty or rey:::
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Position on NRC Regulatory Guide 1.97, Revision 3 Requirements for Post Accident Neutron Monitoring System (NMS)
V Table of Contents f.iat Disclaimer of Responsibility 11 Abstract iv
1.0 INTRODUCTION
1 1.1 Purpose of Licensing Topical Report 1
1.2 Sponsoring Utilities 2
2.0 NHS DESCRIPTION AND DESIGN BASIS 3
2.1 General 3
2.2 Use of Source Range Monitoring System (SRM) 5 2.3 Use of Intemediate Range Monitoring System (IRM) 6 2.4 Use of Power Range Instrument System 7
2.5 BWR Potential for Returning to Criticality 8
2.6 Reliability of NHS 10 3.0 APPLICATION OF REGULATORY GUIDE 1.97 FOR 11 NEUTRON FLUX MONITORING 3.1 General 11 gm 3.2 RG 1.97 Requirements for Reactivity Control 11
}
Instrumentation 4.0 EVENT ANALYSIS TO DETERMINE REQUIRED NMS POST-ACCIDENT 13 MONITORING FUNCTION 4.1 Introduction 13 4.2 Selection of Events 14 4.3 Events Analyzed 16 4.4 Conclusion 36 5.0 FUNCTIONAL DESIGN CRITERIA FOR POST-ACCIDENT 39 NEUTRON MONITORING 5.1 Scope 39 5.2 Requirements, Bases, and Existing Capabilities 39 5.3 Conclusion 47 6.0 ALTERNATE OR SUPPORTING INSTRUMENTATION 48 6.1 Rod Position Information System (RPIS) 48 6.2 Traversing Incore Probe (TIP) 49 6.3 Other Plant Parameters 49 6.4 Summary 50 7.0 BWROG CONCLUSIONS 51
8.0 REFERENCES
52 A
)
9.0 LIST OF ACRONYMS AND ABBREVIATIONS 53 i
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Abstract O
1 Regulatory -Guide 1.97 sets functional design requirements _ for post-accident neutron monitoring instrumentatic,n.- These requirements are generic to both
~
The installed systems at many BWRs do not meet the current Reg. Guide 1.97 requirements.
This report provides a BWR event analysis _ methodology that establishes the-importance of the MS _for' post-accident mitigation.
A wide range of events are considered in keeping _with the intent of Reg.. Guide 1.97. The results of the event analysis are used to set appropriate neutron monitoring post-accident functional design criteria. Deviations from Reg. Guide 1.97 require-ments are justified.
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1.0 INTRODUCTION
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1.1 Puroose of Licensina Tooical Reoort The Regulatory Guide 1.97 (RG 1.97) requirements which deal with design and qualification of the Neutron Monitoring System (NMS) have remained an issue with BWR plants.
In order to resolve this issue the BWROG RG 1.97/ Neutron Monitoring System Connittee was fomed in 1986 to carefully study BWR events and determine the post-accident monitoring function of the BWR Neutron Monitoring System (NMS).
Regulatory Guide 1.97 classifies neutron flux as a key variable for monitoring reactivity control. As such it is required to meet Category 1 design requirements for a specified range of 10-6 percent to 100 percent full power.
Category 1 imposes the most stringent design and qualifica-tion criteria consisting of redundant channels qualified in accordance with Regulatory Guide 1.89, " Qualification of Class IE Equipment for Nuclear Power Plants," and the methodology described in NUREG-0588 n
" Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment".
These requirements reflect a significant depar-ture from the original BWR plant design and licensing basis.
The BWROG believes that the post-acciht system requirements should be evaluated against the increase in overall plant safety and the benefits to plant operation.
The Comittee has examined the HMS requirements considering the operator actions specified by the BWR generic Emergency Procedure Guidelines I
(EPGs).
This approach is in confomance with NUREG-0737 Supplement This inte-requirements for an integrated emergency response program.
grated program has led to reconsidering the category classification of the HMS.
The goal of this report is to establish post-recident design requirements for the HMS which are acceptable alternates to those speci-
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i fled in RG 1.97.
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O This Licensing Topical Report is general'ly applicable for all BWR 2-6s-even though some plant specific differences exist in system and component design.
1.2 Soonsorina utilitici The sponsoring utilities of the RG 1.g7/ Neutron Nonitoring System Commit-tee are identified below:
Boston Edison Company Cleveland Electric Illuminating Company Commonwealth Edison Company Detroit Edison Company Georgia. Power Company GPU-Nuclear Corporation Gulf States Utilities Co.
Iowa Electric Light and Power Company Long Island Lighting Company I
Systems Energy Resources Incorporated Nebraska Public Power District New York Power-Authority Niagara Mohawk Power Company Northeast Utilities Northern States Power Company Tennessee Valley Authority 1
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, O 2.0 NEUTRON MONITORING SYSTEM DESCRIPTION AND DESIGN BASIS b/
2.1 General The purpose of the NMS is to detect neutron flux in the reactor core over a wide span ranging from shutdown conditions to high power conditions reouiring reactor scram.
In addition to the wide range needed, the spatial distribution of the neutron flux is needed to assure that operat-As ing limits are not exceeded at any location within the reactor core.
BWR designs have increased core size, the neutron flux pattern has become more complex such that monitoring local flux conditions becomes necessary to avoid uneven fuel burnup or fuel damage.
To respond to these needs, General Electric developed the neutron moni-toring system (NMS) using detectors located inside the core.
These in-core flux monitors provide detailed spatial flux indication which improves both reactor plant safety and fuel utilization.
The NHS design
[
basis for BWRs never required a post-accident neutron monitoring function h
since there are no design basis accidents that rely on operator action to control reactor power.
To assure that all flux levels expected throughout the range of reactor operation are monitored, three basic types of neutron detectors and signal conditioning equipment are used.
The approximate power level ranges for the three neutron monitoring subsystems which overlap to provide neutron flux information from fully shutdown to greater than rated power are given in Table 2-1.
A brief description of each sub-system is given below, a)
The Source Range Monitoring (SRM) Subsystem is used for monitoring the neutron flux from the fully shut down condition through criti-0 2
cality to a neutron flux of approximately 5 x 10 n/cm /sec (approx-imately 0.0005% power). This system uses' retractable detectors and i
pulse counting electronics coupled with logarithmic readout.
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Table 2-1.
Approximate Power Level Ranges for Neutron Monitoring Subsystems REACTOR POWER
(% RATED)
NEUTRON MONITOR SUBSYSTEMS 125
=
100 LPRMs/APRHs 15
=
1
=
0.0
. --- RG197.rgm 3/14/88
The Intermediate Range Monitoring (IRM) Subsystem overlaps the SRM
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b)
Y system from about 1 x-108,je,t/sec (approximately 0.0001% power) and extends well into the power range (>15% of full power). - The. IRM The uses retractable detectors and voltage variance electronics.
l subsystem consists of ten ranges of one half decade linear steps of n
output proportional to neutron flux.
c)
The power range (1% to full power) is monitored by fixed fission
.g chambers, the Local Power Range Monitoring ~(LPRM) Subsystem, is l
amplified and used for several purposes.
The output ofineutron.
detectors near a control rod selected for motion: are displayed immediately above the reactor control switches, and are used in_ the' Rod Block Monitor (RBM) Subsystem to automatically prevent control rod withdrawal if the local flux change is too great.
In addition, the output of each LPRM is routed to the process. computer for use in power - distribution and local limit determinations, fuel burn-up calculation, etc. The outputs of selected sets.of the chambers' are averaged to provide four to eight channels of core average neutron-y flux and is referred to as the Average Power Range Monitor -(APRM) 1 Subsystem.. The output of this subsystem.is displayed.to the opera-tor, provides an input to the reactor protection system and provides e
rod blocks based on power and flow relationships.
2.2 Use of Source Ranae Monitorina System (SRM)
The SRM subsystem is primarily used for monitoring the neutron flux when-l the plant is fully shutdown (approximately 10-6 percent power) and during In the source range, the neutron flux 'is monitored by four startups.
independent fission counters which are inserted to about the midplane of the core by the drive mechanisms.
l In "STARTUP" mode, the SRM. subsystem provides.the information-needed for reactor startup and low power operations.
It is'used to monitor subcri-tical. multiplication in order to observe the approach to criticality = and
.O deterinine when the reactor is about to go critical. When the' reactor is the critical, the SRM is used to monitor the reactor period :to allowf
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As startup progresses operator to maintain it within specified limits.
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the SRM provides the necessary range to achieve criticality and provides overlap with the intermediate range monitors.
When the reactor reaches the power range, the detectors are moved to a position approximately 2 feet below the core. This places the detectors il in a low neutron flux so that burnup and activation of the detectors are However, even when fully withdrawn they do remain on scale minimized.
with the reactor at moderate or high power.
Therefore, if a significant g
they reactivity control event were to occur with the SRMs withdrawn, would provide some trend indication to the operator.
During controlled plant shutdowns the SRM detectors are inserted by the operator to monitor the complete shutdown.
Such monitoring is not essential if all control rods are inserted by a reactor scram, in which case the operator inserts the detectors as soon as practical.
In the " REFUEL" mode the SRM subsystem is used to monitor neutron count rates during core alteration; the operator monitors the SRM suberitical p
count rate to verify that the reactor is not approaching critical.
The SRM indication at low count rates verifies system operability.
In
" REFUEL" the SRMs are used to provide a scram signal in the non-coincidence mode at some plants if desired, but normally the SRMs cannot cause a plant scram.
The SRM subsystem was not designed by GE to be Class IE since its design use is to monitor flux during controlled plant startups or shutdowns and it does not provide any automatic plant trips during power operation.
Use of Intermediate Ranae Monitor System (IRM) 2.3 The IRM subsystem overlaps the source range and extends into the power range to at least 15% of full power.
It normally employs eight (8) individual fission chambers which are withdrawn itke the SRM detectors during full power operation to maintain their expected life and to reduce A)
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activation.
The IRM drive mechanisms are similar to those used for the
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V SRMs.
l During reactor startup the IRMs provide the required automatic safety protection and operator information required for power ascension through the intermediate range.
In order to control the reactor period during control rod withdrawal in the intermediate power range, the operator keeps the IRMs on scale by changing the IRM range switches. Thereby, the operator avoids short reactor periods and maintains a prescribed startup rate.
If the reactor period is tco high or the operator is unable to keep the IRMs on scale, an automatic plant scram results.
Following plant shutdown or scram, the IRMs are again driven into the i
reactor core to monitor neutron flux and verify a complete shutdown. The operator must keep the IRMs on scale by changing the IRM selector switch-es.
f The IRM subsystem has been designed by GE to be a Class IE system (except f
for the drive mechanism).
This subsystem provides automatic plant trip inputs to the reactor protection system WPS) circuitry during startups.
2.4 Use of Power Ranoe Instrument System The Local Power Range Monitors (LPRMs) overlap the SRMs and measure neutron flux over a range from approximately 1% to 125% of rated power on a linear scale.
LPRM assemblies each contain four fission chambers which are at fixed locations and a calibration guide tube.
The chambers are uniformly spaced throughout the core in an axial direction and lie in four horizontal planes.
Each fission chamber is connected to a d-c amplifier with a linear output.
Internal controls permit adjustment of the amplifier gain to compensate for the reduction of chamber sensitivity caused by burnup of its fissionable saterial.
The LPRMs are used when a control rod or group of control rods is select-ed for movement.
The readings from the detectors adjacent to the rods j
N being moved are displayed on the operator's control benchboard together RGl97.rgm 3/14/88
This allows with a display of the position of the rod or group of rods.
(k for careful ascensions in power and controlled burnup during power After reactor scram, the LPRMs read off-scale low.
operation.
The average power level is measured by four to eight average power range Each monitor measures bulk power in the core by averag-monitors (APP.M).
8-ing signals from as many as 25 LPRM detectors distributed throughout the Actual APRM control room readout is in percent of rated power.
I-core.
The reactor operator uses the APRMs to observe changes in reactor power and to determine the need for rod control or recirculation flow adjust-The output signals from these monitors are also used to initiate ment.
scrams or rod blocks.
If protective actions are taken, the system is used in combination with control rod position indication and other vessel After parameters to verify the reactor has been scrammed or shutdown.
scram the APRM goes downstale.
Most LPRM and APRM equipment has been designed by GE to be Class IE since
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it provides automatic plant trip inputs to the reactor protection system Power is usually supplied from the RPS buses so that a (RPS) circuitry.
in a RPS initiated power failure to the LPRMs or APRMs would result scram.
2.5 BWR Potential for Returnina to Criticality When the scram system automatically inserts all control rods, a BWR is Without deliberate opera-immediately placed in the shutdown condition.
tor action the control rods cannot withdraw after the scram and no c cal (liquid boron) control is required.
Full control rod insertion In results in reactor shutdown with margin for all reactor conditions.
fact, other rod patterns with less than full rod insertion also result in Some BWRs have experienced rod a shutdown reactor for all conditions.
bounce following scram, where a number of rods lock at Notch 02 instead of all the way in (at Notch 00).
However, the plants who have experi-enced this problem have determined that they are shutdown with margin even if all control rods insert and lock at Notch 02.
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Other plants have experienced rod drift, where a single rod which is being withdrawn will fail to lock and is therefore, withdrawn further than intended.
However, this has never happened to' rods that were locked.
It has only happened where an operator has taken-deliberate action to unlatch a control rod and move it to 'a new position (for instance in a plant startup).
t!
Liquid poison (boron injection) is only relied upon under the very rare circumstance of inability to insert a sufficient number-of control rods
[
to achieve cold shutdown.
No BWR worldwide has ever resorted to liquid boron injection to facilitate plant shutdown and the implementation of the ATWS rule (10 CFR 50.62) has further reduced the probability of this type of event.
For these reasons, BWRs have been designed and licensed using neutron flux indication as a requirement only for normal operation.
The NMS can, however, be used as an operator enhancement for abnormal or accident situations.
Pressurized Water Reactors (PWRs), on the other hand, are routinely shut
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down by a combination of control rods and liquid baron in the primary coolant.
For PWRs, even with full-control rod insertion, there are conditions where the plant can be critical if there is insufficient li.. id poison (boron) in the core.
Post-accident neutron monitoring is, therefore, more important for PWRs.
The inherent design of the BWR is very forgiving in hypothetical accident circumstances as demonstrated in Reference 5.
Many pressure and active design features contribute to the capability of BWR's to withstand reactivity-type events. Under normal operating conditions the reactor is in an energized state in terms of system pressure and recirculation flows.
Events which lead to a lowering of the energy state of the system, such as pressure reduction or loss of forced coolant flow, automatically lead to a reduction in the plant fission power level. The basic design of a BWR is such that natural circulation of the coolant is sufficient to provide required cooling to the core in the event that power to recirculation pumps is' lost providing that adequate reactor water level is maintained.
The negative power coefficient and Doppler RGl97.rgm 3/14/88
t absorption automatically and promptly truncate power transients which x
might result from operator error or equipment malfunction.
2.6 Reliability of NMS The reliability of the existing BWR Neutron Monitoring System was deter-mined by analyzing the GE " COMPASS' data base over the period of 1975 through 1985. The percent in unavailability of the subsystems of the HMS are shown below:
Percent Subsystems Unavailable LPRMs 0.26 APRM 0.01 IRMs 0.07 SRMs 0.05 b
Note that Percent Unavailability is the average plant unavailability due j
to forced plant shutdown or critical path maintenance associated with this equipment.
In addition to the GE " COMPASS" data base, the INPO Licensing Event Report Data Base was researched to determine whether any events had been No reported which resulted in the loss of neutron monitoring capability.
events which cause the total loss of monitoring capability have been reported.
From the operating experience during normal, startup and trip (scram) conditions the existing neutron monitoring instrumentation provides highly reliable monitoring and trip functions.
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3.0 AP.f11 CATION OF REGULATORY GUIDE 1.97 FOR NEUTRON FLUX MONITORING 3.1 General l
l Regulatory Guide 1.97 describes design requirements for monitoring instrumentation used during and following accidents in terms of "cate-gory" and " type".
Type designation is based upon requirements for directing operator actions for which no automatic action is provided under design basis accident events (type A), verifying accomplishment of safety functions (type B), verifying fission product barrier integrity (type C), verifying system operation (type D) and assessing radioactivity release (type E).
Category designation is detemined by importance of function.
Key variables for monitoring safety functions are assigned to the most stringent category (category 1); system operating status is assigned to a less stringent category, though they must have a highly reliable power source (category 2); backup and diagnostic instruments or instruments where the state of the art will not support a higher class are assigned to the lowest category (category 3).
D The determination of design requirements for accident monitoring instru-mentation considers a spectrum of events such as loss-of-coolant acci-
- dents, anticipated operational occurrences that include Anticipated Transient Without Scram (ATWS), and reactivity excursions that result in release of radioactive materials.
Key variable instrumentation must be capable of surviving the most severe accident environment in which it is required to operate for the length of time its function is required.
3.2 RG 1.97 Reouirements for Reactivity Control Instrumentation Regulatory Guide 1.97 requires that instrumentation be provided to monitor reactivity control following an accident.
It identifies neutron flux over control rod position and boron concentration as the key vari-able for determining the accomplishment of reactivity control.
The guide has specified neutron flux monitoring as Category I which represents the highest design requirement.
Category 1 design requires RGl97.rgm
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redundant, seismically and environmentally qualified channels powered by The monitors must provide unambiguous indication
[;
Class 1E power sources.
which is recorded and displayed in a manner consistent with good human factors practices.
for
- l RG 1.97 specifies neutron flux monitoring as a -Type 8 variable determining whr".ner f.*nt safety functions are being accomplished for reactivity core,rol. To assure that safety functions are being performed p
for key Type 8 variables, the instrumentation must be qualified for its expected accident environment in which it is located and over a suffi-cient time period into the accident.
RG 1.97 does not classify neutron flux as any other variable type.
Reactivity is controlled automatically in design basis events by the RPS No reactivity control actions must be taken by reactor scram system.
operators for design basis events, thus neutron flux is not a. type A variable.
Neutron flux gives no indication of fuel clad integrity, thus s
it is not a type C variable.
Similarly, since neutron flux does not verify system operation or measure radioactive releases it is neither a -
type D or E variable. Therefore, the classification of neutron flux as a type B variable is appropriate and neutron flux monitoring instrumenta-tion to meet RG 1.97 requirements must be available to ensure that safety functions are being performed.
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4.0 EVENT ANALYSIS TO DETERMINE REOUIRED WS POST-ACCIDENT MONITORING FUNCTION 4.1 Introduction The purpose of the event analysis is to assess the importance of neutron flux indication by examining the consequences of post-accident HMS failures in order to specify appropriate design requirements for post-accident NHS operation.
This section evaluates a range of postulated events where the operator may be required to use the NHS for post-accident monitoring and determines the effect of HMS failure on the outcome.
The top-level instructions for the operator's response to significant transient and accident events are contained in each plant's emergency operating procedures (EOPs).
Supplemental plant procedures provide more detailed system operating instructions, but these instructions must not k(,J) conflict with the top-level E0P instructions.
Each plant has based their plant unique E0Ps upon the generic BWR Owners' Group Emergency Procedure
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Guidelines (EPGs).
The EPGs contain the fundamental actions based upon symptomatic conditions that plant operators must take in response to postulated events.
The latest EPG, Revision 4, was submitted to the NRC for approval in early 1987.
However, the analysis is not effected by differences in the operator actions using NMS indication since Revision 2 of the EPGs. Therefore, EPG Revision 4 is used as the basis for operator actions in this study.
The EPGs address conditions both less severe and more severe than design basis accident conditions.
For example, the scope of EPG development included instructions to mitigate events when the reactor is not shut-down, when power is still high, and when the operator cannot determine shutdown status or power level.
The EPGs do not specify the methods or instrumentation that the operator is to use to determine values and trends of specific parameters.
If the reactor is not shutdown, the operator would prefer to use the APRMs if RG197.rgm 3/14/88
available to determine reactor power level.
IRMs could also be used to s
SRMs and determine power level if they had been inserted into the core.
x IRMs could also indicate current shutdown status when they are driven N t the MS instrumentation will not however, guarantcv into the core.
reactor will remain shutdown as it is cooled down and rwactor u 9: ions For example, the HMS may show that the reactor is shutdown now, change.
but as the reactor is cooled down moderator reactivity coefficients change and if control rods are not sufficiently inserted the reactor can Therefore, current shutdown information from the HMS return to power.
does not mean that the reactor cannot return to power later.
Other information may be used by the operator to determine reactor shutdown status or power level.
This is discussed in more detail in Section 6 of this report.
The scope of EPG development includes the ability to safely mitigate events when the HMS is not available.
4.2 Selection of Events A broad spectrum of events have been considered in establishing the events which are to be analyzed.
These include all FSAR transient and accident events as well as ATWS and other events beyond the plant design The evaluated event basis to be consistent with the intent of RG 1.97.
categories include:
o Transients with scram o
Accidents with scram o
Transients without scram Other occurrences without scram o
In general, these are events which occur with the reactor operating at full power.
" Transients without scram" includes both events where no control rods are ever driven into the core and those with some or delayed "Other occurrences without scram" assume that the control rod insertion.
Reactivity events operator is eventally able to insert control rods.
such as rod withdrawal errors and control rod drop accidents have been
(
Other events such as i~
considered in the " Accidents with Scram" category. RG197.rgm 3/14/88
f i
i a LOCA with a scram failure have not been considered credible events for D
this analysis since they are of very low probability and are_ outside the-scope of ATWS requirements.
Leaks or other occurrences with scram are
[
within the scope of events considered, but they are bounded by the other event categories.
-f The events Events within each category have been selected for analysis.
selected are bounding for the post-accident MS evaluation in that J
together they meet the following criteria:
The neutron flux information provided by MS would be most useful to 1.
the operator.
The spectrum of operator actions related to post-accident neutron 2.
flux monitoring are exercised.
The spectrum of conditions the operator must evaluate to determine 3.
appropriate actions if the MS were to fail are exercised.
e The impact on plant parameters and operator actions if the MS were j
4.
to fail are maximized.
For these evaluations, the postulated post-accident MS failure is Since defined as a failure of all APM, LPM, IM and SM indication; the failure is postulated to occur after the accident has initiated, automatic trip functions which occur prior to the presence of a hostile environment are not effected by the failure.
Similarly, s. MS failure during normal plant conditions would be governed by technical specifica -
In addition, automatic trip functions are outside the tion requirements.
scope of a post accident instrumentation requirements specification.
Consequently, event initiation after a MS failure is-not considered.-
l e
- RGl97.rge
'3/14/88 1
--- ~--
l 4.3 Events Analyzed The events analyzed are summarized in Table 4-1.
A detailed description j
of 'each event including operator actions, the environmental conditions various IMS components would experience, and impact of a WIS failure are
[
provided below.
4.3.1 Irnisients with Scram a
1)
Event:
Feedwater controller failure - maximum demand.
==
Description:==
A feedwater controller failure increases feedwater flow to the maximum the system can deliver.
With excess feedwater flow, core j
l inlet temperature decreases and water level rises to the high level main turbine and turbine driven feedwater pump trip setpoint.
The turbine trip causes a reactor scram signal.
The high water level trip occurs before the temperature decrease causes an increase in neutron flux to
.f reach the high flux scram setpoint.
t The operator ' enters the EPGs for RPV control (level, i
Operator Actions:
l pressure, and power) following turbine trip on the high RPV' pressure signal (above the high RPV pressure scram setpoint).
The EPG actions are:
confirm automatic actions, establish high pressure injection systems for. long-term maintenance of RPV water level, control reactor pressure with the turbine bypass valves, and monitor and control reactor l
The EPG specified operator actions related to powe. control are power.
complete as soon as it is determined that the reactor-is shutdown.
Control rods "all in" indication ~ would immediately confire reactor shutdown. The APRMs would trip downscale and the operator could ~not use the NMS to confirm reactor shutdown until the SRMs or IRMs had been e
driven into the core region..
Environmental Impact: The environment near NMS equipment in the reactor,.
j drywell, and reactor building would not be effected by this event because the reactor is not isolated from the main condenser and normal heat i RG197.rge 3/14/88
/m Table 4-1 Summar3 tent Analysis Operator Use of HMS Imoact of NMS Failure Event Classification Event No impact from a NMS 4.3.1 Transients With Scram o Feedwater controller o Monitor shutdown o
failure - maximum after initial event failure alone l
demand has been mitigated o With additional RPIS (no isolation from and SRMs or IRMs failure some routine main condenser) have been inserted actions required, but no boron injection o Turbine trip with o Monitor shutdown o No impact from a MMS bypass failure after initial event failure alone (isolation from has been mitigated o With additional RPIS main condenser) and SRMs or IRMs failure some routine have been inserted actions required, but no baron injection 4.3.2 Accidents With Scram o Large Break LOCA o Not used by the o No adverse impact from a NMS failure (rapid blowdown operator and ECCS injection) o Small Break LOCA o Monitor shutdown o No impact from a NMS (operator control after initial event failure alone of RPV pressure and has been mitigated o With additional RPIS waterlevel) and SRMs or IRMs have failure, some ATWS been inserted.
actions including boron injection are possible o Control Rod Drop o Monitor shutdown o No impact from a MMS Accident following scram failure alone (reactivity insertion)
(IRMs or SRNs are o With additional RPIS already inserted as failure some routine event initiates at actions possible, but low power) no boron injection RG197.rgm 3/14/88 L____-__-__-____.
Tablo 4-1 i
/ of Event' Analysis i
ntinu:d)
Event Classification Event operator Use of mms Irr_t of lets Failure-4.3.3 Transients Without Scram o MSIV closure o Detemine power level o No impact from a 1845 with complete o Monitor power during failure; obvious that-scram failure boron injection all ATWS mitigation actions are required (isolation from o Long term boron i
main condenser) concentration monitored i
by sampling:
o Stuck open relief o Detemine power level o Boron injection and valve with partial o Monitor power level other ATUS mitigating-scram failure during water level actions more likely;.
(no. isolation reduction and control.
could lead to same-from main condenser) rod insertion to actions as taken for l
potentially avoid boron MSIV closure with' injection complete scram failure.
4.3.4 Other Occurrences o Recirculation pump o Monitor power as o No adverse impact from Without Scram seal leak control rods are a 1895 failure (leak inside inserted.
containment) o Monitor shutdown when power is reduced sufficiently for SIUts and IRMs to be inserted.
o. Scram discharge o Monitor power as o No adverse. impact from volume leak control-rods are a ISIS failure (outside containment inserted except for Mark III
-o Monitor shutdown when plants) power is reduced sufficiently for SRMs and IRMs to be inserted.
- RG197.rge 3/14/88 m
.. ~.
Summat((heed)-
N
, Event Analysis Table 4-1 (co Doerator Use of MS Impact of MS Failure Event Classification -
Event i
4.3.4 Other Occurrences o toss of drywell o Monitor power as o-No impact from a MS coolers control rods are failure Without' Scram inserted (continued) o Monitor shutdown j
when power is reduced sufficiently for SMs and IMs to be inserted M
1 f
~
i
- 19..
RG197.rge 3/14/88
removal systems continue to function. Therefore, the environment is not
[O expected to degrade significantly from normal operation conditions.
,\\
Impact of MS Failure: The Rod Position Indication Systen (RPIS) is u3ed to confirm control rod position and reactor shutdown as discussed in Without the E S, the operator cannot use neutron flux infor-Section 6.0.
nation to confirm reactor shutdown.
If RPIS is also not available to I~
confirm reactor shutdown, the operator would follow EPG instructions to place the reactor mode switch in " SHUTDOWN" (which provides an automatic I
reactor scram signal) and run back recirculation pumps if they had not already been runback or tripped.
These are routine actions for turbine The trip type events which would occur even if MS and RPIS were working.
operator would also initiate the alternate rod insertion (ARI) system and enter the Level / Power control contingency. The operator would use alter-nate indications to determine reactor power as also discussed in Section If however, the operator could not use alternate information to 6.0.
determine that reactor power is below approximately 3% power, then the EPG An instruction to specified actions are to trip the recirculation pumps.
inject liquid boron or lower RPV water level to reduce reactor power would A
not be generated because the suppression pool would not heat up suffi-ciently to cause these actions.
Therefore, the Level / Power control con-tingency would not specify any actions different than normal level control for this event.
Thus the actions the operator would take for this event with a loss of the MS even coupled with a loss of RPIS and inability to determine power is below approximately 3% power do not significantly affect the plant response.
2)
Event: Turbine trip with bypass failure.
This A variety of malfunctions will cause a turbine trip.
==
Description:==
trip will cause the turbine stop valves to close and initiate a reactor With a turbine bypass failure, the reactor will pressurize until scram.
the SRVs open to relieve pressure and discharge energy to the suppression pool.
ik RG197.rgm 3/14/88 c
Operator Actions:
The nperator enters the EPGs for RPV control on the L
high RPV pressure signal. The EPG specified actions are:
confirm auto-matic actions, manually open SRVs to terminate SRV cycling (or confirm low-low set SRV operation), establish reactor high pressure injection for long-term maintenance of RPV water level, and monitor and control reactor As discussed above, the operator completes EPG specified power power.
control actions as soon as it is determined that the reactor is shut down (control rods are sufficiently inserted or neutron flux indication).
Environmental Impact:
The environment near NMS equipment in the drywell and reactor building would not be effected by this event because the minimal heat addition is confined to the suppression pool and not signi-Therefore, ficantly propogated to the areas that contain NMS equipment.
the environment is not expected to degrade significantly from normal operation conditions.
Impact of NHS Failure:
If RPIS can confirm reactor shutdown, the operator
[
enters the scram procedure and there is no impact from the NMS failure.
[V If RPIS fails and the operator cannot use HMS to confirm reactor shutdown the operator would continue to follow the routine EPG instructions for turbine trip type events as outlined above.
Instructions to initiate boron injection or to lower RPV water level would not be generated for this case due to the very small suppression pool heat up.
Thus the actions the operator would take for this event with a loss of NHS even coupled with a loss of RPIS and inability to determine power is below approximately 3% power do not significantly affect the plant response.
4.3.2 Accidents with Scram 1)
Event:
Large Break LOCA with Failure of One Division of Low Pressure ECCS The break causes immediate high 'drywell pressure and low RPV
==
Description:==
water level LOCA signals. The plant scrams and begins a rapid depressur-ization through the break.
The low pressure ECCS injection restores RPV
,b water level.
- However, the initial water level drop may cause a RGl97.rgm 3/14/88
f3 Core reflood will be with a highly voided
'l
)
significant core uncovery.
mixture inside the shroud which swells water level above the top-of-active As the core is subcooled by the large amount of water injected, fuel.
water level will settle out at just above the top of the jet pumps with the injection rate equal to the rate at which water is pouring out the break (BWRs without jet pumps rely on core spray to maintain core cool-1.
ing).
The initial operator actions for this event are rela-
[.
Operator Actions:
tively limited.
The event occurs rapidly and the automatic systems are designed such that the operator does not have to take manual actions until the reactor is depressurized and reflooded with the low pressure ECCS.
The operator cannot restore water level above top-of-active fuel for this the Therefore, when he confirms that the reactor is shutdown, event.
actions in the primary containment flooding contingency are taken to flood containment until water level can be restored above the top-of-active fuel.
Environmental Impact:
This event will product a harsh environment for equipment in the reactor and drywell.
NHS equipment in those locations would not be expected to survive this event long enough to either verify the APRM downscale trip or to drive SRMs or IRMs into the core.
Impact of NMS Failure:
If RPIS can confirm reactor shutdown, the operator continues with the containment flooding actions and there is no impact If RPIS and NHS both fail it is likely that the from a MS failure.
operator will know the plant is shutdown by virtue of the excessive automatic low pressure injection into the core and the absence of any resulting power excursion.
If the operator cannot determine the reactor is shutdown, then level control actions are transferred to the level / power However, for this event the outcome would be nearly control contingency.
identical as the same systems specified in the containment flooding contingency would be utilized in the level / power control contingency in an effort to restore reactor water level.
[U RG197.rgm 3/14/88
For this event the operator does not need to use the MS to assess reactor
~
power to determine if recirculation pumps should be tripped '(they already are), boron should be injected (it would quickly be diluted in the sup-i pression pool), or water. level should be lowered (it already is low)._
Therefore, a MS indication failure does not significantly effect plant response or plant safety for this event.
V Small Break LOCA with Failure of High Pressure Make-up in 2)
Event:
Conjunction with Loss of Offsite Power at Time of Scram
-I
==
Description:==
The small break causes a containment pressurization above the scram setpoint. The loss of offsite power is assumed to 'cause a loss of feedwater and MSIV closure.
RPV water level decreases: due to decay -
heat bolloff and steaming through the break and the SRVs. With failure of-the high pressure systems, the RPV is depressurized by the automatic depressurization system (ADS) and low pressure systems restore RPV' water level.
Operator Actions:
The operator enters the EPGs for RPV control.and con-
]
L tainment control on the high drywell pressure scram signal. ~ The EPG specified actions for RPV control are: confirm scram and. isolation, attempt to restore high pressure systems, and manually open-- SRVs to ~
teminate SRV cycling (or confirs low-low set SRV operation).- When the operator detemines that high pressure systems cannot be restored and low pressure systems are available, the operator will' open SRVs to depressur-ize the RPV and restore RPV water level.- The' operator completes EPG specified actions related to power. control' as soon as it is determined _
j that the reactor is shut 'down or RPIS indicates that ' control rods are sufficiently inserted. _ The APRMs will have tripped - downscale, but the operator cannot use HMS to confirm reactor shutdown until the SRMs or IRMs can be driven into the core.
)
This event has an automatic ' scram signal when Environmental Impact:
drywell pressure reaches the high drywell pressure scram setpoint.
Typical drywell temperatures and the time after break occurrence are presented in Table 4-2 for a spectrum of break sizes for different type
, RGl97.rge 3/14/83-1 j
f]
Table 4-2.
Typical Time and Drywell Temperature When
(- V-Drywell Pressure Reaches the Scram Setpoint s
Drywell Tannerature (O )
F 4
2 Timeisec)
Break Size fft )
Mark I 4
170 0.1 58 141 0.01 775 146 0.002 i
Mark 11 3
167.
C.1 60 140 0.01 970 136 0.001 h
Mark III 5
168 O.1 65 148 0.01 144 144 0.005
..i Assumptions:
1.
Drywell coolers are operating 2.
Maximum technical specification allowable drywell-to-wetwell. bypass leakage area Initial drywell temperature is 135'F 3.
. Scram is assumed to occur when the drywell has pressurfred 2 psig above 4.
^b normal operating drywell pressure J RGl97.rge 3/14/88 y
-,.w w
r,,-
n
-~+
containments.
These are representative results that do not supersede
[
(
plant specific evaluations. The environment prior to scram is mild and a 25 failure would not be expected prior to scram.
The environment would be expected to gradually degrade following scram as the break continues to discharge energy to the drywell.
The extent of MS equipment surviva-bility depends upon the capability of the installed components.
Impact of 2 5 Failure:
If RPIS can confirm reactor shutdown, the operator enters the scram procedure and there is no impact from the MS failure.
This determination is made early in the event, requires RPIS operability for very short durations, and does not need to be repeated later in the If the RPIS fails and the operator cannot use MS to confirm event.
reactor shutdown, then the operator would be led to the EPG instructions to enter the Level / Power control contingency.
Other ATWS mitigating actions (trip recirculation pumps, initiate ARI, etc.) would have already occurred or have no effect on event outcome.
However, if the small break causes the suppression pool to heat up sufficiently, the operator may have to lower water level and inject boron.
The action level which requires a RPV water level reduction in the Level / Power Control Contingency is power above approximately 3 percent power (or cannot be determined), along with high suppression pool tempera-These are indica-ture, and either a SRV open or high drywell pressure.
tions that power is high, there has previously been a significant heat input to the containment, and the heat input is continuing.
With enough heat input to the containment to heat up the suppression pool, the pre-sence of the break would make it difficult for the operator to use other plant data such as steam flow and SRV position to determine that power was below approximately 3 percent power and avoid the water level reduction.
If the operator could not determine that reactor was below approximately If the 3%, then the operator would be required to reduce water level.
operator were to reduce water level, the lower water level would be maintained until the requisite amount of boron had been injected or until it could be determined that sufficient control rods had been inserted.
This water level reduction does not jeopardize adequate core cooling.
N Thus, initial failure of both the RPIS and the MS could result in un-RGI97.rgm 3/I4/88
necessary water level reduction and boron injection, but would not threat-en plant safety.
3)
Event: Control Rod Drop Accident
==
Description:==
The most limiting response to this event is when the reactor is at low power.
During the normal process of withdrawing control rods a high worth control rod sticks in the fully inserted position and becomes decoupled from its drive mechanism. After the drive is withdrawn, the rod frees and drops out of the core.
The rapid rod withdrawal causes a reactor power increase.
A high power signal scrams the reactor which terminates the accident.
The plant has been designed to accommodate this event without experiencing significant fuel failures or a radioactivity release.
Operator Actions: The operator will be monitoring neutron flux with IRMs or APRMs while pulling control rods.
Following scram the operator enters the scram procedure and uses the HMS to monitor neutron flux and confirm Q
k reactor shutdown.
The event does not generate any EPG entry condition since it does not significantly effect RPV water level, RPV pressure, or drywell pressure.
Environmental Impact: The environment near NHS equipment in the reactor, drywell and reactor building would not be effected by this event.
Impact of HMS Failure:
Following scram, if the operator cannot determine reactor power because of a NHS failure, then the operator enters the EPGs.
The operator would then rely on RPIS to determine the reactor is shut down and again enter the scram procedure.
If RPIS and NMS were both to fail, then the operator would take actions to initiate ARI.
However, other indications would show that power is below approximately 3% power and the action to trip recirculation pumps would not be required.
Furthermore, with no pool heatup and all SRVs closed even "without RPIS and NHS, boron There-injection and other ATWS mitigation actions would not be required.
fore, there is no adverse consequence from a NMS failure for this acci-dent. RG197.rgm 3/14/88
4.3.3 Transients Without Scram 1)
Event: MSIV Closure with Complete Scram Failure
==
Description:==
During full power operation all MSIVs close.
MSIV closure generates a scram signal. The scram is not successful; the reactor pres-surizes until several SRVs open and discharge steam to the suppression Plants with safety valves that discharge directly to the drywell pool.
may have these valves open briefly depending upon plant capacity and specific plant incorporated automatic ATWS mitigation features to runback feedwater and trip recirculation pumps.
The scram failure with MSIVs closed will give an EPG Operator Actions:
entry condition.
The operator will place the reactor mode switch in
" SHUTDOWN".
If automatic ATWS features have not activated he will initi-ate ARI and trip recirculation pumps.
Without control rods inserted sufficiently to assure shutdown, water level control will be transferred to the level / Power control contingency.
The rapid and continued pool heatup (along with the reactor not shutdown) will quickly generate an instruction to inject liquid boron.
With reactor power well above the approximately 3% power action level, the operator will lower RPV water to The operator will also try to drive control rods reduce reactor power.
into the core, though for this event no rod insertion is assumed.
When liquid boron has been injected sufficiently to assure hot shutdown, RPV water level is restored to its normal range.
Three-dimensional sub-scale tests have shown that if boron has collected in the lower plenum, it is mixed in the RPV volume and shuts down the reactor when the water level is restored.
After liquid boron sufficient to assure cold 0
The hot and shutdown has been injected, a 100 F/hr cooldown is begun.
cold shutdown boron amounts are pre-determined based on conservative concentrations and volumes and are not based on neutron flux measuremen The Level / Power control contingency also establishes a priority on injec-Outside the shroud injection systems are used in preference tion systems.
b to inside the shroud injection systems to promote thermal mixing and avoid
' RGl97.rgm 3/14/88
(D O
a potential power excursion that could result from injecting subcooled water into a core that is not shutdown.
In addition, if emergency RPV depressurization is required when a sufficient number of control rods are not inserted, the EPG specifies actions to assure that excessive amounts of subcooled water are not inserted into the RPV.
During this event the operator would use the NHS to determine reactor power level and trends.
The indicated pcwer immediately following the scram failure would be approximately 40% to 60% of rated power.
This is well above the approximately 3 percent power used as an action level to determine if the recirculation pumps should be tripped and if RPV water level should be lowered. The NHS would show a power reduction during the water level reduction and it would provide verification that liquid boron was in fact reaching the core and shutting down the reactor.
- However, once boron injection has begun, it is not terminated until the required amount has been injected or until control rods are inserted. Neutron flux O
indication is not used to terminate boron injection.
Environmental Impact:
Though this event has a dramatic pool temperature increase, the temperaturs increase near drywell equipment (cables, connec-tors, SP.M/IRM drive motors) and near equipment in a Mark III containment (electronic equipment, cables) would experience a slowly degrading envi-ronment as heat was transferred from the suppression pool to the surround-With a peak suppression pool temperature of 180 F to 200'F 0
ing spaces.
for this event, the HMS equipment will only be exposed to a mildly degrad-ed environment.
The extent of NHS equipment survivability depends upon the capability of the installed components.
Some BWRs are designed with unpiped safety valves which discharge directly to the drywell.
This event may cause multiple safety valve discharge for those plants over a sufficient time period to, severely degrade the drywell environment in which NHS equipment is located.
O Impact of NHS Failure:
If RPIS is available, the absence of all-control-V rods-in indication will quickly alert the operator to the scram failure even if NHS indications of high power were not available.
If NHS and RPIS RG197.rgm -
3[14/88
J..
both fail, then with reactor pressure at or above normal operating pres-sure and. several SRVs dischaviing steam 'to' the suppression pool, it will be very obvious.to the operator that the reactor is not shutdown.and that SRV dis-power is well above _ the approximately 3% power ' action level.
charge line indication (acoustic monitors or. pressure sensors) will give Therefore, the operator _
positive verification that several SRVs are open.
would take the same actions to inject boron and lower RPV water level as would be taken were MS and/or RPIS indications available.
As the event progresses, the MS is an enhancement to the operator for monitoring neutron flux during boron injection and when the water level is raised as this mixes the boron in the reactor.. volume.
However, the dramatic reduction in steam discharge through the SRVs will' be adequate-verification that reactor power is being reduced.
i MS could be used as a backup to boron concentration measurements when control rods are not. inserted but after the BWR has been shut-down with liquid boron to monitor the subcritical flux as an indication of boron dilution. The quantity of boron injected includes provisions for recircu-lation piping, RWCU, shutdown cooling system volume, etc.
Since _ boron carryover with steam is negligible, boron dilution can only occur as the This dilu-
-l result of liquid leakage or vessel flooding through the SRVs.
tion could require the makeup and injection of additional boron into the J
reactor pressure vessel if the control rods cannot be inserted.
2)
Event:
Inadvertent SRV Opening with Partial Scram Failure During full power operation a SRV opens and fails.to close.
==
Description:==
When the suppression. pool h'as heated up to the pool temperature at which reactor scram -is required,: the operator manually initiates the scram.
With a partial scram failure'some of the control rods -are inserted on the' initial scram signal and/or the operator has' success with manual attempts-to drive control rods. The _ operator still follows. the actions specified ~
in the EPGs,- but the plant consequences are less extensive than for the J
previous case with no control rod insertion.
\\
! RGl97.rgm q
3/14/88-l
Operator Actions:
Suppression pool temperature above the limiting condi-d tion for operation (LCO) causes the operator to enter the containment Actions to initiate pool cooling will not be sufft-control procedure.
cient to terminate the temperature rise and the operator will quickly enter the RPV control procedure where the instruction is to initiate With a scram failure as indicated by control rod position reactor scram.
and neutron flux indication, the operator will follow EPG power control instructions to place the mode switch in SlWTDOWN initiate ARI, run back and trip recirculation pumps, inject boron with the standby liquid control Without control rod system (SLCS) and attempt to drive control rods.
insertion sufficient to assure shutdown, water level control will be transferred to the Level / Power control contingency. The operator will use the main turbine bypass valves to control RPV pressure.
With a SRV open, reactor power still above approximately 3 percent power, and elevated suppression pool temperature, the operator will lower RPV water level per the Level / Power control contingency instructions to reduce m
natural circulation and reduce generated power.
When a sufficient pre-determined amount of boron has been inserted, RPV water level will be i
Q restored to its normal range and the operator will proceed to take the This restoration occurs when a sufficient number plant to cold shutdown.
of control rods to assure shutdown are inserted or when a specific amount of boron has been pumped into the reactor.
The level restoration action is not based on neutron flux information.
The actual response would depend upon how many control rods went in and how soon in the event they were inserted.
If the initial insertion was sufficient to reduce power below approximately 3 percent power, then the recirculation pumps would be run back but not tripped, and the reactor Furthermore, water level would not be lowered to reduce reactor power.
j liquid boron injection could be delayed or even avoided if the subsequent heat addition rate to the suppression pool did not exceed the pool cooling capability.
1 If the initial control rod insertion was not sufficient to prevent boron the RPV water level reduction and additional rod insertion p
injection, I RG197.rgm 3/14/B8 i
This would allow could reduce power below approximately 3 percent power.
/
\\
for a less extensive water level reduction than for a complete scram The IMS would be used by the operator to monitor neutron flux failure.
reductions as control rods are inserted and/or as water level is lowered, and to verify that boron is reaching the core region.
It would be used to confirm that the reactor power has dropped below approximately 3 percent power and therefore, determine which less extensive actions are warranted.
Environmental Impact:
The HMS equipment will experience an environment that is degraded even less than for the MSIV closure with complete scram failure event since the steam production is reduced and most of it goes to the main condenser instead of to the containment.
The BWRs which are designed with.unpiped safety valves may have their high setpoint valves open for a short duration.
The resulting environment and extent of NMS equipment survivability would require plant specific review.
- However, part of the plant's design basis is to assure that the unpiped safety valves do not open for events with scram when relief valves function
[~
properly.
Any safety valve opening would be further evidence to the operator that a scram failure has occurred and assure that appropriate 4
ATWS mitigation actions are taken even if the degraded environment causes a NHS failure.
Impact of HMS Failure:
If RPIS is available, absence of all-control-rods-in indication will quickly alert the operator to the scram failure even if NMS indications of high power are not available.
If NHS and RPIS both fail, the positive control room indication of the open SRV in addition to steaming through the turbine bypass valves, etc. will be obvious indica-tions that the reactor is not shutdown and that power is above the ap-proximately 3% power action level. However, as control rods are inserted, the steaming rate will decrease. The turbine bypass valves will close and For these j
the RPV will begin to depressurize through the stuck open SRV.
conditions, a NHS indication failure will make it more difficult for the operator to determine if power is above or'below the approximately 3%
power action level. However, the inability to determine reactor power has been incorporated conservatively into the EPGs; if the actual event has Oe reduced power below approximately 3 percent power, but all indications are RGl97.rgm 3/14/88
inadequate (including a IMS failure), then the operator must take the These actions as though power were above approximately 3 percent power.
(j actions which trip recirculation pumps, initiate liquid boron injection, and maximize the RPV water level reduction are more extensive than would need to be taken, but do not threaten plant safety.
If liquid boron is unnecessarily injected, it would have to be cleane1 up, but the actions do not threaten adequate core cooling.
Furthermore, if the control rod insertion (as monitored by the RPIS) is sufficient to assure reactor shutdown under all conditions without liquid boron, then the more exten-sive ATWS control operator actions can still be terminated or avoided.
In summary, with a NHS failure the partial rod insertion for this event may make it more difficult for the operator to determine if power is above or below the approximately 3% power action level.
The EPGs assure that whether the operator can determine this or not, plant safety is main-tained.
4.3.4 Other Occurrences Without Scram p
Recirculation pump seal leakage, failure to scram when in-1)
Event:
itiated by the operator
==
Description:==
During normal full power operation a recirculation pump seal begins to leak excessively. The operator runs back recirculation pumps to The scram does not minimum speed and manually initiates reactor scram.
occur.
Operator Actions:
The operator enters the EPGs for RPV control when the scram does not occur.
The operator uses feedwater to control reactor water level, trips the turbine and uses turbine bypass to control reactor pressure, initiates the alternate rod insertion system (ARI), trips recirculation pumps, and if ARI has not inserted them, attempts to manual-ly drive control rods.
The containment heatvp for this event would be small because the reactor is not isolated from the main condenser and the drywell coolers prevent the leak from causing a substantial drywell Therefore, other ATVS mitigation actions such as Q
temperature increase. RG197.rgm 3/14/88
The boron injection and water level reduction would not be required.
(~(
operator would use RPIS to monitor rod position and MS to monitor power /-
neutron flux as control rods are inserted.
As power was reduced, the operator would insert IRMs and SRMs to continue monitoring neutron flux until the reactor was fully shutdown.
The EPG actions related to power control are completed as soon as it is determined that the plant is shutdown or RPIS indicates that control rods are sufficiently inserted.
Environmental Impact:
Pump seal leakage will increase temperature and humidity in the bottom of the drywell in the vicinity of the NHS The actual undervessel cabling, connectors, and SRM/IRM drive motors.
response would be less severe than the results presented for the smallest break in Table 4-2 for the small break LOCA. The extent of NHS equipment survivability depends upon the capability of the installed components.
Impact of HMS Failure:
With a MS failure the operator must use other means to monitor reactor power reductions such as turbine bypass flow as Q
discussed in Section 6.0 of this report.
Due to little containment heatup, most ATVS mitigating actions would not be required even if RPIS and HMS were both to fail.
Therefore, plant response is not adversely effected by a HMS failure for this event.
2)
Event:
Partial scram followed by scram discharge volume leakage During normal full power operation a spurious scram signal
==
Description:==
is generated.
The plant only partially scrams.
Following the partial scram a leak develops in the scram discharge volume which adds heat to the As this reactor building (primary cor.tainment for Mark III designs).
event requires multiple failures of safety related equipment it is not considered to be of significant concern; the analysis of this event was suggested by the NRC.
The operator enters the RPV control procedure when the Operator Actions:
The actions taken to control reactor pressure, scram does not occur.
water level and power are essentially the same as for the leak inside Oh containment discussed above.
Other ATWS mitigation actions such as boron RGI97.rgm 3/14/88
[
injection and RPV water level reduction would still not be required since the reactor does not isolate from the main condenser and the suppression t
pool does not heat up for this event.
The operator also enters the secondary containment control procedure on high temperature or high water level in a sump or area of the secondary containment.
If the leak propogates the high temperature or water level to more than one area of the secondary containment, emergency RPV depres-surization would be required. This is to assure that if equipment in the secondary containment begins to be effected by the leak, the RPV will be in a low energy condition with the maximum number of systems available to provide core cooling. Special level control actions would be required for a blowdown with the reactor not shutdown as specified in the level / power control contingency to assure that a cold water induced reactivity excur-sion does not occur.
Environmental Impact:
The scram discharge volume is in the vicinity of NHS electronic equipment for some plant designs.
Thus, the leak could cause NHS electronic equipment failure under these conditions.
Each plant A
would have to evaluate the location of NHS equipment relative to compon-ents that could leak water on them to determine the potential for this failure.
Impact of NHS Failure:
A NHS failure would have little impact on the operator or plant response to this event.
The operator would continue to monitor rod position with RPIS as control rods are inserted. This event has essentially no containment heatup (a small heatup for Mark III con-tainments) and most ATWS mitigating actions would not be required even if RPIS and NHS were both to fail. Therefore, plant response is not adverse-ly effected by a NHS failure for this event.
w i RGI97.rgm 3/14/88
3)
Event: Loss of drywell coolers, failure to scram Q'
==
Description:==
During normal full power operation all drywell coolers simultaneously fail.
The drywell heats up and pressurizes until it reaches the scram setpoint where a scram is initiated. The scram does not occur.
Operator Actions:
The operator enters the EPGs for primary containment control on high drywell temperature and for RPV control when the high drywell pressure scram signal occurs.
If drywell temperature approached 0
the qualification temperature for ADS solenoids (typically 340 F), drywell spt ay would be initiated and/or the RPV would be blown down.
But these temperatures are not expected for this event. The operator uses feedwater to control reactor water level, turbine bypass to control reactor pres-sure, initiates the ARI system, trips recirculation pumps and if ARI has not inserted them, attempts to manually drive control rods. The operator would use RPIS to monitor rod position and HMS to monitor power / neutron flux as control rods are inserted.
As power was reduced, the operator would insert IRMs and SRMs to continue monitoring neutron flux until the reactor was fully shutdown.
The EPG actions related to power control are completed as soon as it is determined that the plant is shutdown or RPIS indicates that control rods are sufficiently inserted.
High drywell pressure is one of the conjunctive criteria for lowering RPV water level, but there is no suppression pool heatup for this event so neither water level reduction nor boron injection would be required.
Environmental Impact:
The drywell would heatup with a relatively low humidity for this event.
This heatup could cause a slow degradation of NHS equipment in the drywell, but is not expected to cause a rapid NHS failure.
The extent of NHS equipment survivability depends upon the capability of installed components and actual drywell temperature response to this event.
Impact of NHS Failure:
With a NHS failure the operator must use other means to monitor reactor power reductions such as turbine bypass flow as discussed in Section 6.0 of this report.
If the operator could not use RGl97.rgm 3/14/88
.. ~ ~
1945 to confire reactor shutdown, then the operator must continue current-actions until RPIS indicates that control rods sufficient for shutdown are inserted.
If RPIS also fails, the operator would have to wait to cool' down, etc. untti some means can be employed (see Section 6.0) to determine
~
reactor shutdown. However..with all steam going to the main condenser, no other ATWS mitigating actions would be required.
Therefore, plant re--
sponse is not adversely effected by a MS failure for this event.
4.4 Conclusions The events analysis considered the operator's use of the MS for tran-sients with scram, accidents with scram transients without scram, and; other occurrences without scram.
The analysis details how the operator uses the MS if available, and the impact on event outcome if the MS were to fail.
The events selected provide a spectrum of impacts, but they bound the MS importance for all events that are within-the scope of the l
1 Reg. Guide 1.97 criteria.
For Transients With Scram the long tens post-accident function _ for neutron
('
flux monitoring is not necessary after reactor shutdown is confirmed.
These events have very little environmental.' impact on letS equipment and j
operator actions are not significantly affected by the loss of neutron j
monitoring capabilities.
For the bounding transient with scram events the l
~
l operator nonna11y uses the MS to confirm low power, but upon lets failure there are other clear indications which will show that power is low..
1 Boron injection or other abnormal operator actions are not expected to be required as a result of the lets failure.
Therefore, these events do'not-4 set design requirements for the letS.
For Accidents With Scram the long tern post-accident function for neutron j
flux monitoring is not necessary after reactor shutdown is confirmed.
These events impose severe environmental conditions for large pipe breaks, i
but the automatic plant response makes l#tS indication of low importance to the operator.
Note that under these conditions the plant automatically i
scrams, the water level drops to the top of jet pump elevation -(approxi-O mately 2/3 of core height),
and low pressure injection systems 36 -
RG197.rge 3/14/88
.=ir wi-g b
74---*w-
--wy
-.--.w,p.yP=.%q 5-y-e v
y--
yryr-
-l For non jet pumps automatically provide for required. core cooling.
plants, the core is completely uncreared for large recirculation line Note also that fort this breaks and cooling is provided by core spray.
j event boron injection would be of little value since the boron would very rapidly be diluted in the suppression pool.
For smaller breaks the MS can be used along with the RPIS to verify the plant has been shutdown.
Analysis of these ' events have shown that the
-l operator actions are not affected by the loss of MS as -long as the RPIS I
Furthermore, the initial environment is not harsh and remains operable.
under these conditions neither MS or RPIS equipment would be expected to I
Accident with scram events, fail prior to verification of plant shutdown.
therefore, do.not establish design requirements for the MS.
For Other Occurrences Without Scram, the MS would be used to monitor These events may reactor power while control rods are being inserted.
local environmental conditions that could potentially fail 'or
')
cause degrade MS equipment, but the bulk suppression pool temperature is not
]
.( O significantly impacted'since these events do not isolate from the main
~i Therefore, most ATWS mitigation actions would not be required j
condenser.
for these events even if the MS were to fail and these events do no l
design requirements for the M S.
the MS provides the primary means of For Transients Without Scram, neutron flux monitoring and power ' level indication as-ATWS mitigation.
actions required by the EPGs are taken.
Other indications are available to verify MS indications or to be the primary source of reactivity information if the MS fails. The importance of the NMS to the operator-is dependent upon the severity of the ATWS, :0nce control rods are suf-If the plant is ficiently inserted, this monitoring is not required.
l required to remain shutdown on liquid boron.over long periods, boron' sampling and laboratory analysis becomes the. primary means for reactiv This boron measurement is a monitoring, with the MS serving as backup.
more reliable reactivity variable since the MS would not detect dilution l
when the boron concentration is well below the concentration
(
recriticality would occur.
, RGl97.rge
-l 3/14/88
_ _ =., _. _
Transients Without Scram do not impose a harsh environment except for plants with unpiped safety valves during high power ATWS events which isolate from the main condenser. However, for large ATWS events, the lack of all-rods-in indication and the containment response will assure that the operator takes appropriate ATVS mitigation actions even if the NMS fails.
For lesser ATWS events (when partial control rod insertion occurs or the plant is not isolated from the main condenser) the high setpoint safety valves would open for only a very short duration if at all and the result-There is, of course, no impact on ing environmental impact is not harsh.
the environment for the majority of BWRs which have only piped safe-ty/ relief valves.
For these lesser ATWS events the HMS enhances the operator actions, since successful verification that power is below approximately 3% power can avoid various non-routine actions.
The lesser ATWS events, therefore, establish design requirements for the NHS.
Note, however, that even if the operator takes the most extensive ATWS mitigating actions for these 7
less severe ATWS events, plant safety would be maintained.
Whb V RG197.rgm 3/14/88
5.0 FUNCTIONAL DESIGN CRITERIA FOR POST-ACCIDENT NEUTRON MONITORING 5.1 Scope The purpose of this section is to define and justify alternate post-accident requirements for the MS and to compare these requirements to the Category I requirements in R.G. 1.97.
These criteria are developed as a result of the post-accident operational uses of HMS instrumentation dis-cussed in the previous section.
A general evaluation of existing NMS instrumentation to meet this criteria is also included. Note that this is not a complete MS design criteria specification since it does not address criteria for startup, normal operation, automatic trips, or shutdown. The scope of the criteria is limited to post-accident conditions.
5.2 Reouirements. Bases and Existino Ctoabilities 5.2.1 Rance O
Alternate Requirement:
1 to 100%
RG 1.97 Requirement:
10-6% to 100%
Basis:
If successful scram occurs, post-accident neutron monitoring is not meaningful and reactivity control is assured by the control rod latching design.
The alternate requirement covers the possible ATWS conditions from immediately after the scram failure until power has been reduced to below the APRM downscale trip of approximately 3% power.
This will allow the monitoring of reactor pown as AT' S mitigating actions are
/
taken as instructed by the EPGs.
An indication range below 1% is not justified since neutron flux informa-tion would only confirm that the plant is shutdown currently, it would not ensure reactivity control as reactor temperature decreases.
Monitoring neutron flux under such a partial shutdown situation is of relatively low significance to plant safety due to the inherent safety of the BWR design which establishes negative reactivity feedback for control of the fission
(
reaction.
Furthermore, if the plant is shutdown with liquid boron, it is RGl97.rgm 3/14/88
O more important to measure boron concentration directly by sampling reactor i
water than to measure it indirectly with neutron flux indication.
4 All BWRs_ meet the alternate. criteria with existing Existing Capability:
APRM equipment.
1 5.2.2 Accuraev Alternate Requirement: 12% of rated. power I
RG 1.97 Requirement: None Specified It is not necessary to know the post-accident power with a high Basis:
degree of accuracy until power has been reduced to around 15 or less.
EPGs specify a specific power level (approximately 3%) as an action level in several places and also base boron injection requirements on suppres.
ston pool temperature as a function of power in the 2 to 10% range.
an exact value would help the During events without a complete scram, However,.the operator in assessing the status of reactivity control.-
O reactivity effects of changing RPV pressure, core voids, core injection flow, etc. complicate the operators' ability - to accurately determine Partial rod insertion events -
t reactor power by neutron flux measurements.
To support these events, an.
place the greatest demand on IMS accuracy.
instrument accuracy of 12% of rated power is judged to be sufficient to allow the operator to make appropriate action decisions.
s Although no accuracy requirement is specified by RG 'l.97 directly, refer-ence is made to ANS Standard 4.5 which establishes performance require-ments Instrument loop accuracies are! highly plant speci-Existing capability:-
By proper and frequent calibration of the LPRMs, the power range fic.
accuracy level can be met.
Calculations at one BWR/6 indicated that the APRM loop accuracy is about 2% of scale based on a 1% power supply accura-
[A plant specific evaluation would have to be conducted.]
cy.
O
. RGl97.rge 3/14/88-a
5.2.3 Response Characteristic Alternate Requirement: 5 sec/10% change RG 1.97 Requirement: None specified f
The power range monitors should respond within a few seconds of Basis:
i the actual change in fission rate. The alternate requirement is judged to infonnation to verify the provide operators with sufficiently current accomplishment of reactivity control.
Although RG 1.97 does not directly
[
specify response characteristics, this requiremer.t has been added to be consistent with the ANS Standard 4.5 performance requirements.
Existing Capabilities:
Power range monitors are designed with a response time of I second for a 100% change in flux. This is more than adequate to meet the response characteristic requirement.
5.2.4 Ecuioment Oualification
(
Alternate Requirement: Operate in ATWS Environment RG 1.97 Requirement:
RG 1.89 and RG 1.100 The event analysis in Section 4 of this report identifies the Basis:
limiting events for 25 operation and describes the equipment environment ATWS events for those events where the 25 is important to the operator.
are determined to result in the most limiting environmental conditiens during which the MS operation is needed.
Qualification to design basis environmental standards required by RG 1.89 is not necessary since the MS does not need to function to mitigate Because of its importance to operator accident events.
design basis actions the lesser ATWS events therefore define an appropriate level of system performance when needed.
Qualification qualification to assure standards for the 25 are consequently established on the qualification This rule specifies standards established by the ATWS Rule (10CFR50.62).
ATWS environmental conditions and does not require seismic qualification.
RG 1.100 compliance is therefore not justified for the 25. RGl97.rgm 3/14/88
Existing Capability:
MS equipment is typically designed for abnormal
[]
environments shown in Table 5-1.
These environmental conditions are not expected to be exceeded in the vicinity of MS equipment for the majority of ATWS events.
[A plant specific evaluation of ATWS environments in comparison with design specifications is needed to assure system perfor-mance.]
Table 5-1.
Typical Design Conditions (Abnormal Operation)
Irma Press Humidity Duration 0
Drywell undervessel 135-185 F 0-2 psig 90%
2 hrs Reactor Bldg.
104'F 0.25" rg 90%
100 days Control Room 75 F 0-1" wg 60%
Unlimited a
5.2.5 Function Time Alternate Requirement:
I hour RG 1.97 Requirement: None specified Basis: The function time is tied to the event in which the equipment must survive.
Since the lesser ATWS :vants set the environmental requirements for the WS, those events also set function time requirements.
The key operator actions for those events which relate to power level monitoring are water level reduction and boron injection.
These actions are no longer required when the cold shutdown boron weight has been injected to the RPV or control rods sufficient for shutdown arE inserted. One hour is judged to be sufficient time for the operator to have successfully com-pleted these actions.
Existing Capability:
25 equipment has generally been designed for function times in abnormal operating envircnments which exe.eed this
(,)
specification (see Tablo 5-1).
[A plant specific evaluation of design l
v RG197.rgm 3/14/88
(3
(,) _
specifications and ATWS environments is needed to ccnfirm that this specification is met.)
5.2.6 Seismic Qualification Alternate Requirements: Seismic qualification not required RG 1.97 Requirement:
Seismically qualify Cat.1 equipment as important to safety per RG 1.100 and IEEE-344 The events analysis in Section 4 of this report identifies the Basis:
limiting events for post-accident neutron monitoring.
ATWS events are determined to-set this requirement.
The MS qualification standards are consequently established to be consistent with the ATWS Rule (10CFR50.62).
This rule specifies ATWS environmental conditions and does not require seismic qualification. RG 1.100 compliance is therefore not justified for the NHS.
A)
Existing Capability:
The NHS equipment which provides automatic trip functions have been seismically qualifted to assure that the seismic event does not prevent the automatic trip function.
The remainder of the NHS equipment has generally not been seismically qualified.
Therefore, existing HMS equipment meets or exceeds the alternate criteria.
5.2.7 Eqdundancy and Seoaration I
Alternate Requirement: Redundancy to As u - Reliability RG 1.97 Requirement: Redundant in Diviu n M :,ing RG 1.75 Redundant indication of the power range monitors should be provid-Basis:
ed to assure the operator that the scram function or alternate shutdown measures have been achieved. This criteria is to provide a grecter moni-toring reliability to the control room operat'or in the event one :hannel Due to the capacity to achieve reactivity control withoat the is lost.
NMS, and the brief function time when indication is required, sepa' ation (A) of these signals is considered desirable, but not essential.
, RG197.rgm 3/14/88
Existing Capability: The existing MS meets the alternate criteria.
A i
i L /5.2.8 Power Sources Alternate Requirement: Uninterruptable and Re!iable Power Sources RG 1.97 Requirement: Standby Power Source (RG 1.32)
Basis: Power supp1tes should be reliable and available during most events in order to avoid unnecessary actions in some events such as are described in Section 4.0.
They should be from uninterruptable sources in order to monitor neutron flux continuously during any automatic load shed events, but because of the many alternate methods to establish reactor power (see Section 6), it is not necessary that Class IE power be provided.
Existing Capability: The power supplies for MS equipment may vary among plants.
Most utilities power the sensors and displays from the RPS instrument bus.
[A plant specific evaluation is required to review the power distribution to the MS including the recorders to verify that the
/^}
instrument power is not lost during events by load shedding logics or similar schemes.]
I 5.2.9 Channel Availability i
Requirement: Available Prior to Accident RG 1.97 Requirement: Available Prior to Accident Basis:
The WS should be fully available during power operation, w inform the operator of a high flux level when the scram did not occur. No deviation from the RG 1.97 requirement is intended.
I Existing Capability: Since power range instrumentation is available while the plant is at power, existing MS designs meet this criteria. Addition-ally Section 2.6 indicates that the WS availability is extremely high.
(D c)
RG197.rgm 3/14/88
5.2.10 Quality Assurance Alternate Requirement: Limited QA Requirements Based on Generic Letter l
85-06 RG 1.97 Requirement: Application at Specified Reg. Guides Basis:
The NHS should have QA requirements applied consistent with the importance of this instrumentation to verify a safety function.
NRC generic letter 85-06, " Quality Assurance Guidance for ATWS equipment that is not Safety Related", should be applied to ES monitoring equipment since it is consistent with the use of the NMS to support ATWS events.
Existing capability:
Much of the NHS equipnent is Class IE since it provides trip functions to the reactor protection system.
The remainder of the equipment is designed, procured, and installed as non-safety related.
[Lompliance with Generic Letter 85-06 should be verified on a plant specific basis.]
Disolav and Recordino O'5.2.11 Requirement:
Continuous Recording RG 1.97 Requirement:
Continuous Recording Basis:
Recording of the NMS signals should be provided for post accident diagnostic review. No deviation from the RG 1.97 requirement is intended.
1 Existing Capability:
Every NHS channel is recorded in existing desig.s.
Therefore, this requirement is satisfied.
5.2.12 Eautoment Identification Requirement:
Identify in Accordance with CRDR RG 1.97 Reautrement:
Identify as Post-Accident Monitors 1
I Basis:
NMS recorders should be clearly marked to be consistent with 1
(\\
results of the detailed control room design review (CRDR).
This does not tQ RG197.rgm 3/14/88
deviate from the RG 1.97 intent except to add that integration with the k)N
/
CRDR be accomplished to be consistent with NUREG 0737, Supplement I requirements.
Existing Capability:
Recorders are normally clearly marked.
[This item should be verified on a plant specific basis.]
5.2.13 Interfaces Alternate Requirement: No Interference with RPS trip functions RG 1.97 Requirement:
Isolators to be used for alternate functions Basis:
Non-lE portions of the IMS should be separated from the Class IE portions of the NMS in accordance with plant licensing requirements so that they do not interfere with reactor prote* tion system (RPS) functions.
This alternate requirement is intended to be consistent with the ATWS Rule (10CFR50.62).
Existing Capability:
Existing designs fulfill she alternate requirement.
5.2.14 igrvice. Test and Calibration Requirement: Establish in Plant Procedures RG 1.97 Requirement:
Establish in Plant Procedures Basis:
The NMS should be included in normal maintenance programs estab-lished by the plant staff. The capability to demonstrate recorder opera-bility should be provided in addition to out of service alarms if channels fail.
The power range accuracy is dependent on calibration of the LPRM signals and heat balances to provide an accurate measurement of average core wide power. The calibration schedule should be such that the overall loop accuracy requirements are met.
Existing Capability:
[This item is to be verified on a plant specific p
basis.)
f RGl97.rgm 3/14/88
p.2.15 lig an Factors 5
l 1
G Requirement:
Incorporate HFE Principles RG 1.97 Requirement:
Incorporate HFE Principles Basis:
The NMS should be consistent with good human factors engineering (HFE) practices as established by the plant's control room design review.
No deviation from the RG 1.97 requirement is intended.
I Existing Capability:
[This item is to be verified on a plant specific basis.]
5.2.16 Direct Measurement Requirement: Direct measurement of neutron flux RG 1.97 Requirement: Direct measurement of neutron flux o
Basis: To accurately monitor power trends the HMS should directly measure neutron flux. No deviation from the RG 1.97 requirement is intended.
Existing Capability:
Fission type detectors meet the requirement that detectors should directly monitor the neutron flux.
5.3 Conclusion In general, BWR NHSs meet appropriate post-accident design requirements defined by the alternate requirements.
Some plant-unique assessment will be required to confirm compliance with specific alternate requirements.
(v RGl97.rgm 3/14/88
6.0 ALTERNATE OR SUPPORTING INSTRUMENTATION
/
\\
fU The most direct method of determining reactor power is through the use of the NMS.
IMS can also indicate if the reactor is currently shutdown though it doesn't guarantee that the reactor will stay shutdown as condi-tions change.
If the rod position infonsation system (RPIS) indicates "all-rods-in' (or some other positions with less than all-rods-in) then i
the plant design shutdown margin requirement assures that the reactor is shutdown for all conditions.
If RPIS is not available or does not indi-cate sufficient control rod insertion to assure shutdown, and should direct indication from the NHS become unavailable, alternate indications are employed to ascertain reactor power levels.
Inferences can be drawn with respect to reactor power by monitoring other indications including the reactor coolant boron concentrations, flux levels from the traversing in-core probe system (TIP), or the status of plant parameters or compon-ents which are in someway linked to reactor power.
A summary of each of these alternate or supporting methods follows.
6.1 Rod Position Infonnation System (RPIS)
The RPIS is a highly reliable monitoring system which provides individual rod position information in addition to " full in" and " full out" indicat-ing lights.
RPIS provides immediate indication of successful core reac-tivity control. When all control rods can be determined to be inserted to the " maximum subtritical banked withdrawal position" (MSBWP as defined by the EPG's), the reactor will remain shutdown under all conditions and all coolant temperatures without liquid boron injection.
If control rod position indication is available, but all rods are not inserted to the MSBWP, then other criteria may be used to determine core reactivity such as the existence of the core design basis shutdown margin with the single strongest control rod full-out and all other control rods full-in, or compliance with the Technical Specification requirements 90verning control rod position and the allowable number of inoperable control rods.
LJ RG197.rgm 3/14/68
If direct control rod position indication is not available, signals from
,q RPIS may nevertheless be providing information to the control rod with-e drawal/ insert circuitry, to the plant process computer, to various annunciators or status indicating lights, and other logic systems.
These signals can be queried to determine if the reactivity function has been achieved.
6.2 Traversina In-Core Probe (TIP)
Although time consuming, neutron flux could be determined with the tra-versing in-core probe (TIP) system.
The TIP system is normally used to calibrate the LPRMs at power and, when inserted into the reactor, is capable of sensing flux in the immediate vicinity of the permanently installed LPRM fission chambers. From a lack of positive reading a shut-down condition could be inferred.
6.3 Other Plant Parameter Indications of Reactor Power A{)
There are many other plant parameters which are linked to reactor power.
Observing their values and trends will give valuable indication of reactor power to the operator.
SRMs and IRHs which are withdrawn will provide ex-core monitor informa-tion. They will not be calibrated to provide an accurate measurement for those conditions but will indicate reactivity trends of increasing, stable, or decreasing neutron flux.
The main steam safety / relief valve positions can be used to determine the approximate power level.
Each valve passes a known steam flow as a frac-tion of rated steam flow.
SRVs typically discharge steam at a rate of 6 to 7% of rated steam flow.
Thus if three SRVs are open and reactor pressure is stable then the reactor is approximately 20% of rated power.
Turbine bypass valve flow and other steam dr' inn equipment such as HPCI and RCIC would give the operator similar information.
SRY position is redundantly and diversely sensed including open/close indicators which n
(
have been installed in the tailpipes of these valves.
(R.G. 1.97 requires RG197.rga 3/14/88 j
these position monitors to be Category 2.) Observing RPV water level and p
pressure (both of which are monitored by R.G. 1.97 Category 1 instruments) values and trends as well as the effect of mitigating actions upon their control will also indicate power level.
For instance, if there is no indication of a 'oreak, RPV pressure is stable, HPCI is operating properly, and water level is still decreasing, then it is quite obvious that power is well above 3%.
There are various other indicators that are useful for determining whether a reactivity control action as been successful.
SLCS status indications including boron tank level will indicate that boron injection actions are being accomplished.
Sampling RPV water will confirm that boron has in fact reached the vessel. Suppression pool temperature (Category 1) trends and the effectiveness af RHR operation will indicate the rate at which energy is being discharged to the containment.
Similarly, containment pressure (Category 1), and containment temperature, including trends or oscillations of these parameters are indirect, but potentially useful indications for determining whether a reactivity control action has been Oi successful.
These indications of plant parameters provide useful infonnation by them-selves or they may be used in conjunction with related plant parameters, such as in the performance of a heat balance ar><und the RPV or the primary containment.
6.4 Sumary In sumary, failure of the most direct indication of reactor power does not preclude the ability of the reactor operator to determine reactor power levels.
Many alternate indications derived Som both component status and parameter status are available from which reactor power may be inferred.
Some alternate indications may require more than one input to determine reactor power. However, based on the multiple inputs available to the operator, sufficient information should be available upon which to base operational decisions and to conclude that reactivity control has p
(
been accomplished. RG197.rga 3/14/88 I
L O7.0 WROG CONCLUSIgi$
v The WROG recognizes the need to identify post-accident monitoring re-quirements for WR reactivity control instrumentation.
It was determined that post-accident neutron monitoring while useful to' the operator is not essential for any event to assure post-accident plant safety is main-
'tained.
It was also concluded that for WRs the Rod Position Indication System (RPIS) provides the primary verification for determining. plant shutdown. However, based on the intent of R.G. 1.97 to provide effective control room monitoring of post-accident plant conditions, specific design criteria for post-accident neutron monitoring capability have been estab-itshed.
The proposed criteria has been compared to the RG 1.97 require-ments and deviations are justified.
After evaluating the existing 25 equipment against the proposed criteria, it was concluded (subject to certain plant unique confirmations) that the existing neutron monitoring system design is generally adequate for every postulated event.
Some plant-specific evaluations may be required to k
confirm adherence with certain requirements.
The BWROG WS Committee believes that the proposed functional criteria represent an acceptable
- alternate to the Category I requirement specified in RG 1.97. While the NHS would be useful to the operator under certain scenarios,.a fully qualified IE NHS for post accident monitoring is not appropriate or justified.
' RGl97.rgm 3/14/88
8.0 REFERENCES
(1) APED 5706, "In-core Neutron Monitoring System for General Electric Boiling Water Reactors", APED-5706, April 1969, General Electric Company.
(2) " Technical Specification Improvement Analyses for BWR Reactor Protec-tion System", NEDC-30815P, May 1985 General Electric Company.
- 31331, (3)
" Emergency Procedure Guidelines, Revision 4",
NEDO March 1987, General Electric Company (4) Letter, C. T. Young to J. S. Post, "Drywell Pressure and Temperature Response to a LOCA", CTY8727, December 8, 1987.
(5)
" Lessons from the Chernobyl Accident for the BWR", GE, September 25, 1986.
~
O RG197.rge l.
3/14/88
O 9.0 LIST OF ACRONYMS AND ABBREVIATIONS v
ADS Automatic Depressurization System ARI Alternate Rod Insertion APRM Average Power Range Monitor ATWS Anticipated Transient Without Scram BWR Boiling Water Reactor BWROG Boiling Water Reactor Owners' Group CRD Control Rod Drive CRDR Control Room Design Review ECCS Emergency Core Cooling System E0P Emergency Operating Procedure EPG Emergency Procedure-Guidelines GE General Electric HFE Human Factors Engineering INPO Institute of Nuclear' Power Operation IRM Intemediate Range Monitor LCO Limiting Condition for Operation LER Licensing Event Report LTR Licensing Topical Report LOCA Loss of Coolant Accident LPRM Local Power Range Monitor A
MSBWP Maximum Suberitical Banked Withdrawal Position f.
/
MSIV Main Steam Isolation Valve
\\U NHS Neutron Monitoring System NRC Nuclear Regulatory Commission PWR Pressurized Water Reactor QA Quality Assurance RWCU Reactor Water Cleanup RG Regulatory Guide RBM Rod Block Monitor RPV Reactor Pressure Vessel RPIS Rod Position Indication System RPS Reactor Protection System SRM Source Range Monitor SRV Safety Relief Valve SLCS Standby Liquid Control System TIP Traversing In-Core Probe e
(
RGI97.rgm -
3/I4/88
()Nb OUJNERS' GROUP 1;;;R:M;"
c/o CPU NUCLEAR e 1 UPPER POND ROAD, BUILDING E e PAAS BVR00-8833 June 13, 1988 U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 Attention: Thomas T. Martin, Associate Director for Technical Review ar.d Inspecticn Subj ect:
BVR OWNERS' CROUP LICENSING TOPICAL REPORT " POSITION ON NRC REGULATORY CUIDE 1.97, REVISION 3 REQUIP.EMENTS FOR POST-ACCIDENT NEUTRON MONITORING SYSTEM" (CENERAL ELECTRIC REPORT NED0 31558)
Centlemen:
,/-('\\)
The BVR Owners' Croup (BUROG) submitted the subject Licensing Topical Report (LTR) to the NRC on April 1,
1988.
This document proposes functional criteria for post-accident neutron monitoring tha; represent an acceptable, alternate to the Category 1
requirements specified in Regulatory Guide 1.97.
The report concludes that a _ fully qualified 1E-post accident Neutron Monitoring System is not appropriate or justif;able
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from a cost-benefit basis.
The NRC was formally notified of the BVROG effort on August 25,1987 (BVROG-8745) and shortly thereafter assigned an NRC contact.
The BVROG committee had a program briefing meeting with the NRC in January Ja88 and co=pleted the document in March 1998.
Recently System Energy Resources, Inc. has been informed that the NRC review of this LTR will not be complete until December 1988, and that the NRC is unwilling to give additional extensions at this time unless effected utilities provide commitment to specific modifications.
We believe that because of plant specific licensing conditions, it is extremely important that NEDO 31558 be expeditiously reviewed by the NRC.
In addition, we' believe the existence of this report gives cause to extend license conditions for the upgrading of neutron monitoring equipment until the NRC has completed its I
review.
If selected BVRs are required to proceed with this' plant modification and the NRC staff later concurs with the NEDO 31558 I
conclusion, very expensive instrumentation will have been purchased without cost benefit justification.
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US Nuclear Regulatory Commission k
BWROC 8833 June 13,1988 Page 2 The BUROC strongly encourages the expeditious technical review of this documsnt to meet the needs of the affected BVRs and recommends additional extensions until the NRC has completed its review.
In order to satisfy 1
the current license conditions, several affected BWRs will have to initiate procurement of qualified instruments in the next month or two.
This letter has been endorsed by a substantial number of the members of the BVR Dwners' Group; however, it uhould not he interpreted a a commit-ment of any individual member to a specific coprge7.of action. Each member must formally endorse the BWROG position in order for that position to become the member's position.
If you desire to discuss this request in more detail, please contact me at your convenience.
Very truly yours, 6
D. N. Grace, Chairman BVR Deners' Group DNG/ dis Attachment ec:
BVR Owners' Croup Primary Representatives BVR Owners' Croup Executive Overview Committee SD Floyd, BVROG Vice Chairman RF Janecek, CECO CC Lainas. NRC l
SA Varga, NRC l
AC Thadani, NRC l
5 Newberry, NRC MV Hodges.
NRC i
JP Joyce, NRC R Evans, NUMARC VS Green, INPO H Vyckoff, EPRI LS Cifford, CE/Bethesda t-
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