NRC Staff Review of Responses to I&E Bulletin 79-08 (Generic Letter 79-23)
GL 79-23
UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D.C. 20555
June 20, 1979
TO ALL BOILING WATER REACTOR LICENSEES
SUBJECT: NRC STAFF REVIEW OF RESPONSES TO I&E BULLETIN 79-08
The purpose of this letter is to advise you of the status of our review of
licensee responses to I&E Bulletin 79-08 (IEB 79-08), and to identify related
concerns which we believe require your further consideration.
The priority established for the review of I&E Bulletins related to the Three
Mile Island, Unit 2 accident (TMI-2) is (1) B&W plants (2) C-E and
Westinghouse plants, and (3) boiling water reactor plants. For this reason,
as well as the limited resources within the NRC staff to apply to these
reviews, we have now targeted for completion in August 1979. In the interim,
we have scheduled a joint meeting at our Bethesda, Maryland office with all
boiling water reactor licensees. This meeting is scheduled for 9:00 am on
June 28, 1979, in Rooms P110/114 of our Phillips Building office in Bethesda,
Maryland. You will be expected to attend the meeting and be prepared to
discuss, among other things, those matters listed below along with a schedule
and procedure for providing the information needed by the staff to complete
the review of these issues.
(1) When our preliminary review of the responses to IEB 79-08 is complete,
we will advise you of any items not satisfactorily resolved. In certain
instances, this may include cases where licensee responses differ, without
apparent justification, from General Electric Company (GE) recommendations.
A copy of the GE recommendations is provided as Enclosure 1.
(2) We expect to prepare a generic report on (TMI-2) matters related to
boiling water reactor operating plants. We expect that this report will
recommend, among other things, further analyses of transients and small
reactor coolant system breaks (including stuck-open relief valves), the
development of appropriate written procedural guidance to operators as
indicated by these analyses, and further training of operators in the use of
these new procedures.
7908130339
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To All Boiling Water Reactor Licensees - 2 -
(3) In certain instances, licensees are using fuel and relying on safety
analyses which are not provided by the nuclear steam supply system (NSSS)
designer. As a result, it is not clear to us what the respective roles of the
licensees, NSSS designers, fuel suppliers, and/or other parties should be in
implementing any requirements that may evolve from item (2) above.
(4) The Advisory Committee on Reactor Safeguards (ACRS) has issued five
letters to the Commission as a result of their examination of the TMI-2
accident. We believe that it would be mutually beneficial for each utility to
provide specific comments on those ACRS recommendations having potential
impact on plant design and/or operation. A summary of the ACRS
recommendations is provided as Enclosure 2.
(5) Individual pressurized water reactor licensees have indicated an
interest in meeting directly with the staff regarding the bulletin items for
their facilities. Experience to date has demonstrated that the staff does not
have time to meet individually with each licensee to resolve those items that
are truly generic to all licensees.
It is clear that there are a significant number of technical issues yet to be
resolved for a large number of operating plants. There are limited resources
available within the staff to perform the necessary work. This situation is
exacerbated by the need to conduct similar and concurrent activities with
owners of pressurized water and boiling water operating plants. At the same
time, there is a need to resolve these issues promptly.
In this regard, we believe there is a compelling need to establish an owner's
group for boiling water reactor operating plants. We expect that such a group
would be needed for the remainder of calendar year 1979. Owner's groups have
worked effectively in the past in minimizing staff and industry resource
requirements to resolve other generic problems. We strongly urge your to meet
with other owners of boiling water reactor operating plants to consider the
formation of such a group prior to the forthcoming generic meeting with the
staff discussed above. This will be one of the principal agenda items at that
meeting.
Please note that our investigation of a number of areas related to the TMI-2
accident, including the ACRS recommendations and the action items from
NUREG-0560, will be included specifically as part of the staff's "Lessons
Learned" activity. You can expect additional correspondence in the future on
those items.
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To All Boiling Water Reactor Licensees - 3 -
If you require any clarification of the matters discussed in this letter,
please contact William F. Kane, the staff's assigned project manager for these
activities on boiling water reactor operating plants. Mr. Kane may be reached
at 301-492-7745.
Sincerely,
Thomas A. Ippolito, Chief
Operating Reactors Branch No 3
Division of Operating Reactors
Enclosures:
As stated
cc w/enclosures:
Service List
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Enclosure 1
(GE Recommendations)
April 20, 1979
Subject: IE BULLETIN 79-08, EVENTS RELEVANT TO BOILING WATER REACTORS
IDENTIFIED DURING THREE MILE ISLAND INCIDENT DATE APRIL 14, 1979
NRC Bulletin 79-08 issued to all BWR reactor facilities with operating
licenses and requested responses to eleven questions. While most of the
questions pertain to operator performance and plant procedures, several are
related to the BWR to a loss of feedwater flow and should be helpful in
responding to question 1 of the bulletin. The attached writeup is provided
for your use in preparing a total response to questions 2,3,4,5b and 10 and
is intended to assist BWR owners in preparing a consistent response to the
bulletin. We suggest that BWR owners communicate with each other to assure
consistency of their response.
The attachment provides the basis for our conclusion that the design of the
BWR has already considered the concerns referenced in the NRC bulletin, such
as:
(1) response to transient caused by equipment failures,
(2) methods of auxiliary heat removal,
(3) redundancy of reactor water level indication,
(4) containment isolation, and
(5) handling of hydrogen
Since the writeup was developed for a typical BWR-2, BWR-3, and BWR-4, some of
the information may need to be modified to be directly applicable to your
specific plant.
Should you have any additional questions, please contact this office.
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NRC QUESTION 2
Review the containment isolation initiation design and procedures, and prepare
and implement all changes necessary to initiate containment isolation, whether
manual or automatic, of all lines whose isolation does not degrade needed
safety features or cooling capability, upon automatic initiation of safety
injection.
RESPONSE SCOPE
This proposed response covers the review of the containment isolation design.
The discussion covers initiation of the manual and automatic features of all
lines whose isolation does not degrade needed safety features. BWR owners
should include the NRC required evaluation of procedures or specific Balance
of Plant systems.
RESPONSE APPROACH
A review was made of drawings and specifications for one Boiling Water Reactor
(BWR) of each type (ie, BWR/4, BWR/3, and BWR/2). Items reviewed were General
Electric Process and Instrument Diagrams (P&ID's), elementary diagrams and
specifications for Nuclear Steam Supply System safety systems, Reactor
Protection System (RSP), the Nuclear Boiler System, and other systems.
CONCLUSIONS
The BWR design provides containment and reactor coolant pressure boundary
(RCPB) isolation (excluding emergency core cooling and make-up systems). The
isolation occurs upon reactor vessel low water level or high drywell pressure
prior to or simultaneous with initiation of emergency core cooling and safety
injection systems. The isolation valves will remain closed until operator
action is taken even if the initiating signal clears. The operator cannot
shutoff Emergency Core Cooling Systems (ECCS) if the low water level condition
exists; therefore, the containment and RCPB isolation is adequate and does not
degrade required safety functions or core cooling capability and no changes
are necessary. During initial plant startup containment isolation was tested
prior to operation and is periodically retested in accordance with the plant
Technical Specifications to assure continued operability.
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ADDITIONAL INFORMATION FOR NRC QUESTION 2
DISCUSSION
For convenience in providing the discussion of the isolation system, systems
are separated and Group B systems represent Balance of Plant or Architect
Engineer (AE) supplied systems. The isolation of each typical group is
described separately.
SYSTEMS GROUPS DEFINITION
Category A - The Nuclear Steam Supply System (NSSS) includes the following
four general groups of systems:
Group A1 - Non-safety systems which connect to the Reactor Coolant Pressure
Boundary (RCPB), penetrate the containment and are not part of the Emergency
Core Cooling Systems (ECCS) or other safety-related coolant injection systems.
Examples of these non-safety systems are:
Reactor Water Cleanup Systems (RWCU)
Main Steam Lines
Feedwater Lines
Group A2 - Systems which connect to the RCPB and are part of the ECCS or
safety related coolant injection system such as:
Reactor Core Isolation Coolant System (RCIC)
High Pressure Coolant Injection System (HPCI)
Low Pressure Core Spray System (LPCS)
Residual Heat Removal/Low Pressure Injection (RHR/LPCI)
Isolation Condensers
Automatic Depressurization System (ADS)
Group A3 - Systems which are shutdown cooling systems that are connected to
RCPB and/or ECCS (eg. RHR shutdown cooling) but which are required to isolate
so as to preserve the ECCS function.
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ADDITIONAL INFORMATION FOR NRC QUESTION 2
Group A4 - Systems which connect directly to containment, such as:
Drywell Floor Drain Pumping Systems
Equipment Drain Sump Pumping Systems
Containment Spray System
Category B - Systems which are outside General Electric scope of supply and
connect directly to the containment consist of two general groups follows:
Group B1 - Containment ventilation systems or other Balance of Plant (BOP)
Systems which connect to the containment atmosphere, including AE supplied
Radwaste Systems.
Group B2 - Closed systems which penetrate the containment (e.g., closed
cooling water, plant air, etc.) These systems do not connect to the RCPB or
containment atmosphere.
ISOLATION FUNCTION
Piping penetrating the containment which do not degrade the core cooling
capability are isolated prior to or simultaneous with the ECCS initiation in
the following ways:
The Group A1 systems (except feedwater lines) automatically isolate at low
reactor water level (level 2). This trip set point is the same one that
automatically initiates the HPCI and RCIC systems. High drywell pressure also
isolates this group of systems, except for the main steam lines.
Feedwater lines are automatically isolated with checkvalves. Feedwater can be
manually isolated from the control room by remote manual switches. The
reactor water cleanup (RWCU) system also isolates upon high temperature in the
RWCU equipment spaces or on high differential flow in the RWCU lines. These
isolation signals prevent loss of coolant. BWR/2, 3 and some BWR/4's isolate
the RWCu at the higher level 3 - BWR owner needs to verify.
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ADDITIONAL INFORMATION FOR NRC QUESTION 2
Automatic isolation of the main steam lines is also initiated upon either 1)
high temperatures in the steam tunnel or steam line spaces in the turbine
building, 2) excess flow in any main steam line, 3) high radiation in the
steam lines, 4) low steam line pressure, or 5) excess pressure in the main
condenser.
THE GROUP A2 systems do not automatically isolate except for RCIC and HPCI
which isolate upon excess steam flow or high temperature in their steam line
or their equipment spaces.
THE GROUP A3 systems are automatically isolated by the low level set point
(level 3) which is also the level at which the reactor scrams. Non-ECCS
portions of the RHR system are also isolated by a high drywell pressure signal
(2PSIG). This assures no loss of the LPCI function.
THE GROUP A4 systems isolate at either reactor water level 2 or upon high
drywell pressure (2 PSIG).
THE GROUP B1 systems are automatically isolated at either low reactor water
(level 2) or upon high drywell pressure. BWR owners should check
implementation. High radiation instruments in the containment exhaust ducts
are provided to isolate the containment ventilation system. BWR owners should
check implementation.
THE GROUP B2 systems isolation is determined by the AE. Reactor low water
level and high containment pressure signals are provided to the AE for this
function. BWR owners should check for implementation.
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NRC QUESTION 3
Describe the actions, both automatic and manual, necessary for proper
functioning of the auxiliary heat removal systems (e.g., RCIC) that are used
when the main feedwater system is not operable. For any manual action
necessary, describe , in summary form, the procedure by which this action is
taken in a timely sense.
RESPONSE SCOPE
This proposal response describes both automatic and manual actions necessary
for proper functioning of the auxiliary heat removal systems. These systems
are used when the main feedwater system is not operable. The procedures are
described in summary form assuming the reactor is scrammed and isolated from
the main condenser.
RESPONSE APPROACH
GE has determined the necessary actions and procedures from design and
operating procedures documents. BWR plant owners should check plant operating
procedures for consistency.
CONCLUSION
Automatic action provides abundant makeup water to the core for initial
cooling. Long term core and containment cooling can be provided with few
manual actions. Information is available to the operator in the control room
to assist him in taking the required manual actions. Information in the
control room permits the operator to verify that the objective of these
actions is being achieved.
5
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ADDITIONAL INFORMATION FOR NRC QUESTION 3
DISCUSSION
The auxiliary heat removal system provided to remove decay heat from the
reactor core and containment following loss of the feedwater systems are:
High Pressure Coolant Injection System (HPCI)
Reactor Core Isolation Coolant System (RCIC)
Low Pressure Core Spray System (LPCS)
Residual Heat Removal (RHR)
The description that follows details the operation of the systems needed to
achieve initial core cooling followed by containment cooling and then followed
by extended core cooling for long term plant shut down.
INITIAL CORE COOLING
Following a loss of feedwater and reactor scram, a low reactor water level
signal (level 2) will automatically initiate main steam line isolation valve
closure. At the same time this signal will put the HPCI and RCIC Systems into
the reactor coolant makeup injection mode. These systems will continue to
inject water into the vessel until high water level signal (level 8)
automatically trips the systems.
Following a high reactor water level 8 trip, the HPCI System will
automatically re-initiate when reactor water level decreases to low water
level 2. The RCIC System must be manually reset by the operator in the
control room before it will automatically re-initiate after a high water level
8 trip.
The HPCI and RCIC Systems have redundant supplies of water. Normally they
take suction from the condensate storage tank (CST). The HPCI System suction
will automatically transfer from the CST to the suppression pool if the CST
water is depleted or the suppression pool water level increases to a high
level.
The RCIC system suction must be manually transferred from the CST to the
suppression pool using controls located in the main control room. This action
would be taken when control room alarms indicate low CST or suppression pool
high water level.
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ADDITIONAL INFORMATION FOR NRC QUESTION 3
The operator can manually initiate the HPCI and RCIC Systems from the control
room before the level 2 automatic initiation level is reached. The operator
has the option of manual control after automatic initiation and can maintain
reactor water level by throttling system flow rates. The operator can verify
that these systems are delivering water to the reactor vessel by:
a) Verifying reactor water level increases when systems initiate (see
water level discussion in response to Question 4).
b) Verify systems flow using flow indicators in the control room.
c) Verify system flow is to the reactor by checking control room
position indication of motor operator valves. This assures no
diversion of system flow to the reactor.
Therefore, the HPCI and RCIC can maintain reactor water level at full reactor
pressure and until pressure decreases to where low pressure systems such as
the Core Spray (CS) or Low Pressure Coolant Injection (LPCI) can maintain
water level.
CONTAINMENT COOLING
After reactor scram and isolation and establishment of satisfactory core
cooling, the operator would start containment cooling. This mode of operation
removes heat resulting from safety relief valve (SRV) discharge to the
suppression pool. This would be accomplished by placing the Residual Heat
Removal (RHR) System in the containment (suppression pool) cooling mode, ie,
RHR suction from and discharge to the suppression pool.
The operator could verify proper operation of the RHR system containment
cooling function from the control room by:
a) Verifying RHR and Service Water (SW) system flow using system
control room flow indicators.
b) Verify correct RHR and SW system flow paths using control room
position indication of motor-operated valves.
c) On branch lines that could divert flow from the required flow paths,
close the motor-operated valves and not the effect on RHR and SW
flow rate.
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ADDITIONAL INFORMATION FOR NRC QUESTION 3
Even though the RHR is in the containment cooling mode, core cooling is its
primary function. Thus, if a high drywell pressure signal is received at any
time during the period when the RHR is in the containment cooling mode, the
RHR system will automatically revert to the LPCI injection mode. The Core
Spray (CS) system would automatically initiate and both the LPCI and CS
systems would inject water into the reactor vessel if reactor pressure is
below system discharge pressure.
EXTENDED CORE COOLING
When the reactor has been depressurized, the RHR system can be placed in the
long term shutdown cooling mode. The operator manually terminates the
containment cooling mode of one of the RHR containment cooling loops and
places the loop in the shutdown cooling mode as follows:
1. trip the RHR pumps,
2. close motor-operated valves in the suppression pool suction and
discharge lines,
3. open suction valves from and discharge valves to the reactor vessel,
and
4. restart the PHR pumps.
In this operating mode, the RHR system can cool the reactor to cold shutdown.
Proper operation flow paths in this mode can be verified by methods similar to
those described for the containment cooling mode.
8
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NRC QUESTION 4
Describe all uses and types of vessel level indication for both automatic and
manual initiation of safety systems. Describe other redundant instrumentation
which the operator might have to give the same information regarding plant
status. Instruct operators to utilize other available information to initiate
safety systems.
RESPONSE SCOPE
The proposed response to this question covers the uses and types of reactor
vessel water level indication (or recording) for both automatic and manual
initiation of Nuclear Steam Supply Systems (NSSS) safety sytems. Descriptions
are also given of other instrumentation available to the operator for
determining reactor water level. Further information on other available
instrumentation is in the response to question 5b.
BWR plant owners should review operators' instructions and any Balance of
Plant (BOP) instrumentation.
RESPONSE APPROACH
A review was made of drawings and specifications for one Boiling Water Reactor
of each type (ie, BWR/4, BWR/3. and BWR/2). Items reviewed were General
Electric Process and Instrument Diagrams (P&ID's), elementary diagrams and
specifications for Nuclear Steam Supply System safety systems, Reactor
Protection Systems (RPS) and the Nuclear Boiler Systems.
CONCLUSIONS
Reactor vessel water level in the BWR is continuously monitored by 7 (5 in
BWR/2) indicators or recorders for normal, transient and accident conditions.
Those monitors used to provide automatic safety equipment initiation are
arranged in a redundant array with two instruments in each of two or more
independent electronic divisions. Thus, adequate information is provided to
automatically initiate safety actions and provide the operator with assurance
of the vessel water level at all times.
These water level measurement devices have operated in BWR plant for 20 years.
Test of BWR water level instrumentation under simulated steam and water line
breaks have been conducted showing satisfactory performance.
9
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ADDITIONAL INFORMATION FOR NRC QUESTION 4
DISCUSSION
The range of reactor vessel water level from below the top of the active fuel
area up to the top of the vessel is covered by a combination of narrow and
wide-range instruments. Level is indicated and/or recorded in the control
room. Typically, there are seven in BWR/3 and 4, and five BWR/2 plants. (see
figure.)
A separate set of narrow-range level instrumentation on separate condensing
chambers (typically five per plant) provides reactor level control via the
reactor feedwater system. This set also indicates or records in the control
room (typically three level indicators and one level recorder).
The safety-related systems or functions served by safety-related reactor
water level instrumentation are:
Reactor Core Isolation Coolant System (RCIC)
High Pressure Coolant Injection System (HPCI)
Low Pressure Core Spray System (LPCS)
Residual Heat Removal/Low Pressure Injection (RHR/LPCI)
Automatic Depressurization System (ADS)
Nuclear Steam Supply Shutoff System (NS4)
All systems automatically initiate on low reactor water level. In addition,
the RCIC and HPCI systems shutdown on high reactor vessel level. In all cases,
except the RCIC, these systems automatically restart if low reactor level is
again reached. (See response to questions 3 for further discussion on this.)
In the case of RCIC, manual resetting is required if high reactor vessel water
level is reached.
10
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NRC QUESTION 5
Review the action directed by the operating procedures and training
instructions to ensure that:
b. Operators are provided additional information and instructions to
not rely upon vessel level indication alone for manual action, but
to also examine other plant parameter indications in evaluating
plant conditions.
RESPONSE SCOPE
The proposed response identifies alternate instrumentation to provide the
operator with information to take corrective actions in the event of loss of
reactor coolant or other abnormal conditions.
RESPONSE APPROACH
In determining the proposed response, a review was made of drawings and
specifications for one typical Boiling Water Reactor (BWR) of each type (ie,
BWR/4, BWR/3, and BWR/2). Items reviewed were General Electric Process and
Instrument Diagrams (P&ID's), and elementary diagrams and specifications.
CONCLUSIONS
Over a dozen other types of instrumentation in the BWR provide the operator
with indirect indication of reactor vessel coolant inventory changes and could
inform the operator of the need to take corrective actions.
12
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ADDITIONAL INFORMATION FOR NRC QUESTION 5
DISCUSSION
Some of the instrumentation which the operator can use to determine changes in
reactor coolant inventory or other abnormal conditions are:
Drywell High Pressure
Drywell High Radioactivity Levels (owner check applicability)
Suppression Pool High Temperature
Safety Relief Valve (SRV) Discharge High Temperature
High Feedwater Flow Rates
High Main Steam Flow
High Containment and Equipment Area Temperatures
High Differential Flow-Reactor Water Cleanup System
Abnormal Reactor Pressure
High Suppression Pool Water Level
High Drywell and Containment Sump Fill And Pumpout Rate
Valve Stem Leakoff High Temperature
An example of the use of this additional information by the operator is a
follows: Drywell high pressure is an indirect indication of coolant loss.
Coincident high suppression pool temperature further verifies a loss of
reactor coolant. High SRV discharge temperature would pinpoint loss of
coolant via an open valve.
Other instrumentation that can signal abnormal plant status but not
necessarily indicative of loss of coolant are:
High Neutron Flux
High Process Monitor Radiation Levels
Main Turbine Status Instrumentation
Abnormal Reactor Recirculation Flow
High Electrical Current (Amperes) to Pump Motors
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NRC QUESTION 10
Review operating modes and procedures to deal with significant amounts of
hydrogen gas that may be generated during a transient or other accident that
would either remain inside the primary system or be released to the
containment.
RESPONSE COVERAGE
The proposed response to this question discusses the system that are available
for removing hydrogen from the primary system as well as treatment and control
of hydrogen in the containment.
Procedures for these operations are unique to each plant and should be
available addressed by each plant owner.
CONCLUSIONS
The BWR has design features to safely dispose of any significant amounts of
hydrogen either in the primary system or in the primary containment.
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ADDITIONAL INFORMATION FOR NRC QUESTION 10
DISCUSSION
During normal operation, the Reactor Pressure Vessel dome is filled with
steam, which flows to the turbine. During reactor isolation, the dome is
automatically vented through the SRVs to the suppression pool. In addition,
the Reactor Pressure Vessel head has a vent line with a valve remotely
operated from the Control Room.
In the event of significant hydrogen release to the primary containment, the
containment atmosphere dilution system maintains hydrogen below flammability.
In addition, there are usually several other systems such as the containment
atmospheric monitoring system, hydrogen recombiners, and containment purge via
standby gas treatment to provide long term hydrogen control. For those plants
with inerting, the possibility of hydrogen flammability is precluded.
15
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ENCLOSURE 2
ACRS RECOMMENDATIONS
A. Letter, M. Carbon to Chairman Hendrie, dated April 7, 1979
Recommendation 1* - Perform further analyses of small break transients
and accidents.
Recommendation 2 - Provide operator additional information and means to
follow the course of an accident; as a minimum,
consider expeditiously:
(a)** unambiguous RV level indication
(b) remotely controlled vent for RCS high points
Recommendation 3 - Item 4b of Bulletin 79-05A considered unduly
prescriptive in view of uncertainties in predicting
course of anomalous small break transients/accidents.
B. Letter, R. Farley to Commissioners, dated April 18, 1979
Recommendation 1 - Natural Circulation-related Items
a.* Detailed analyses of natural circulation mode,
supported as required by experiment, by
licensees and NSSS vendors.
b.* Develop procedures for initiating natural
circulation.
c.* Provide operator means for assurance that
natural circulation has been established, e.g.,
by installation of instructions to indicate
flow at low velocities.
d.* Expeditiously survey operating PWR's to
determine whether suitable arrangements of PZR
heaters and reliable on-site power distribution
can be provided to assure this vital aspect of
natural circulation capability.
e.* Operator should be adequately informed
concerning RCS conditions which affect natural
circulation capability, e.g.,
(1) indication that RCS is approaching
saturation condition by suitable display
to operator of Tc & Th and PZR pressure in
conjunction with steam tables
(2) use of flow exit temperature indicator by
fuel assembly thermocouples, where
available.
* Amplified in Interim Report #2
** Amplified in Interim Report #3
.
Enclosure - page 2
ACRS Recommendations
Recommendation 2*- Thermocouples used to measure fuel assembly exit
temperatures to determine core performance should be
used where currently available, to guide operator
concerning core status (full range capability).
Recommendation 3 - Operating reactors should be given priority regarding
definition and implementation of instrumentation to
diagnose and follow the course of a serious accident,
including
(a)* improved sampling procedures under accident
conditons
(b)** improved techniques to provide guidance to
offsite authorities.
Recommendations 4 - Reiterates previous recommendations that high
priority be given to "research to improve reactor
safety"
(a)* research on behavior of LWRs during anomalous
transients
(b)* NRC to develop capability to simulate wide
range of postulated transients and accident
conditions.
Recommendation 5 - Consideration should be given to additional monitoring
of ESF equipment status, and to supporting services,
to help assure availability at all times.
C. Letter, M. Carbon to Acting Chairman Gilinsky dated April 20, 1979
Recommendation 1 - Initiate immediately a survey of operating procedures
for achieving natural circulation, including:
(a)* event involving loss of offsite power
(b)* consideration of role of PZR heaters.
* Amplified in Interim Report No 2 dated 5/15/79
** Amplified in Interim Report No 3 dated 5/15/79
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Enclosure - page 3
D. Interim Report No. 3 dated May 16, 1979
Recommendation 1 - Examine operator qualifications, training and
licensing, and requalificatin training and testing.
Recommendation 2 - Establish formal procedures for the use of LER
information:
(a) in training supervisory and maintenance personnel
(b) in licensing and requalification of plant
operating personnel
(c) in anticipating safety problems
Recommendation 3 - Consider formal review of operating procedures for
severe transients by inter-disciplinary team, and
develop more standardized formats for such procedures.
Recommendation 4 - Re-examine comprehensively the adequacy of design,
testing and maintenance of offsite and onsite AC and
DC power supplies with emphasis on:
(a) failure mode & effects analyses
(b) more systematic testing of power system
reliability
(c) improved quality assurance and status monitoring
of power supply systems
Recommendation 5 - Make a detailed evaluation of current capability to
withstand station blackout, including:
(a) examination of natural circulation capability
under such circumstances
(b) continuing availability of components needed for
long-term cooling under such circumstances
(c) potential for improvement in capability to
survive extended blackout
Recommendation 6 - Examine a wide range of anomalous transients and
degraded accidents which might lead to water hammer,
with emphasis on:
(a) controlling or preventing such conditions
(b) research to provide a better basis for control or
prevention of such conditions
.
Enclosure - page 4
Recommendation 7 - Plan and define NRC role in emergencies, including
consideration of:
(a) assurance that formal emergency plans, procedures
and organizations are in place
(b) designation of emergency technical advisory teams
(names and alternates)
(c) compilation of an inventory of equipment and
materials needed in unusual conditions or
situations
Recommendation 8 - Review and revise within three months:
(a) licensees' bases for obtaining offsite advise and
assistance in emergencies from within and outside
company
(b) licensees' current bases for notifying and
providing information to offsite authorities in
emergencies
Recommendation 9 - Examine the lessons learned at TMI-2, including
consideration of the following:
(a) behavior, failure modes, survivability and other
aspects of TMI-2 components and systems as part
of the long-term recovery process
(b) determine if design changes are necessary to
facilitate decontamination and recovery of major
nuclear power plant systems
Recommendation 10- Expedite resolution of unresolved safety issues by the
following means:
(a) suitable studies on a timely basis by licensees
to augment NRC staff efforts
(b) use of consultant and contractor support by NRC
staff
Recommendation 11- Augment expeditiously the NRC staff capability to deal
with problems in reactor and fuel cycle chemistry in
the following areas:
(a) behavior of PWR & BWR coolants and other
materials under radiation conditions
(b) generation, handling & disposal of radiolytic (or
other) H2 at nuclear facilities
(c) performance of chemical additives in containment
sprays
(d) processing and disposal techniques for high and
low level radioactive wastes
.
Enclosure - page 5
(e) chemical operations in other parts of nuclear
fuel cycle
(f) chemical treatment operations involved in
recovery, decontamination or decommissioning of
nuclear facilities
Recommendation 12- Reconsider whether or not use of the Single Failure
Criterion establishes an appropriate level of
reliability for reactor safety systems
Recommendation 13- With respect to safety research:
(a) consideration should be given to augmentation of
the FY80 NRC safety research budget
(b) consider orienting a larger part of the safety
research budget toward exploratory (as opposed to
confirmatory) research
Recommendation 14- Perform design studies of a filtered venting or
purging option for containments for possible use in
the event of a serious accident
E. Interim Report No 2, dated May 16, 1979
Amplified many of the recommendations included in earlier ACRS letters
dated April 7, April 18, and April 20, 1979, including ACRS views on
relative priorities to be assigned a number of those earlier
recommendations. (Address amplifications and suggested priority
assignments as appropriate.)
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