ML19343D239
| ML19343D239 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/10/1981 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1833, NUDOCS 8105040027 | |
| Download: ML19343D239 (37) | |
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/$35 f f) j,,2 i s; ISSUE DATE: 4/10/81
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10 k I MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON d
i REACTOR OPERATIONS WASHINGTON, DC MARCH 9 & 10, 1981 Siarch 9 &
The ACRS Subcommittee on Reactor Operations held a meeting of t 10, 1981 in Room 1046, 1717 H St., NW, Washington, DC. The purpose of the meeting on March 9, 1981 was to hear a discussion of the NRC Staff's plans to fomulate a Human Engineering Guide to Control Room Evaluation and on March to begin the Subcommittee's review of Rep. Udall's inq11 ries on ATWS 10, 1981 which were prompted by the June 28, 1980 Browns Ferry 3 parti.al failure to scram.
Noticeofthismeetingwaspubl[ishedintheFederalRegisteronFriday, A list February 20, 1981. A copy of this notice is included as Attachment A.
of attendees for this meeting is included as Attachment B, and the schedule for the meeting is included as Attachment C.
A complete set of meeting Attachment D is a list of the handouts has been included in the ACRS Files.
There were no writ..s handouts and documents associated with the meeting.
statements or requests for time to make oral statements received from members The meeting was entirely open to the public. The Designated of the public.
Federal Employee for the meeting was Richard Major.
DISCUSSION WITH NRC STAFF _
Mr. Kramer of the Division of Human Factors Safety (DHFS) gave an He noted that there are overview of the responsibilities of the DHFS.
control room o
basically four elements to the program which are:
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l MARCH 9 & 10, 1981 REACTOR OPERATIONS MARCH 9, 1981 plant procedures, operator training, and management structure and the When asked by the Subcom-interaction between these four components.
mittee if the optimum mix of these four components has been addressed.
nr. Kramer noted it had not but that the need to address this is recognized and will be addressed by Research.
Mr. Kramer noted the need for the PHFS was perceived after the accident Areas The Division has been in existence for eccut one y' ear.
at TMI-2.
in the Action' Plan that gave rise to the DHFS were cited and included:
I(A), (B), (C), (D), (E), (F), and (G).
Mr. Kramer explained that NUREG/CR-1580 is basically hardware oriented.
As this document evolves into NUREG-0700, more consideration will be given to an analysis of what information operators need to know.
Mr. Kramer noted, in response to questions, there is a fairly high degree of unanimity inhouse ove.r the Staff position taken in the Supplement to NUREG/CR-1580 which is NUREG-0659 (Draft Report)
Supplement to the Draft Report on Human Engineering Guide to Contr The Supplement contains a response to comments, san pie Room Evaluation.
check lists, draft systems review guidelines, and evaluation procecures.
Over the past several months, more specialists in the area of human l
factors engineering have been added to the Staff.
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MARCH 9 & 10, 1981 REACTOR OPERATIONS MARCH 9, 1981 Mr. Voss Moore, Chief cf the Human Factors Engineering Branch, explained that the purpose of the day's session is to obtain general comments from the Subcommittee and its consultants. He noted a letter from the He explained that comments would be welcomed on the is not necessary.
overall program and specifically on the Supplement to the Draft Guidelines Comments would be required in 6-8 weeks to be factored of NUREG/CR-1580.
into the effective guidelines.
The guidelines ' presented during the, meeting were for use by the utilities At a later time, Mr. Moore noted in reviewing their own control rooms.
the Staff would explain the evaluation criteria to be used by the Staff in analyzing the utilities' control room reviews.
As a result of the various recommendations and findings resulting from the investigations of the Three Mile Island accident, the Task Action Plan (and specifically Plan I.D.1) was instituted to implement the various The Staff contracted with Essex Corporation control room recommendations.
Essex was hired to to develop the guidelines for a control room review.
f provide human factors' expertise, and idditional manpower the Staff was Essex had experience in the defense and aerospace fields and lacking.
had worked on the Rogovin Report.
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MARCH 9 & 10, 1981 4-REACTOR OPERATIONS MARCH 9, 1981 4
The draft guidelines issued in the summer of 1980 were the sole product The intent of the Staff in of Essex Corporation with no Staff input.
issuing the guidelines so quickly was to get public comments rapidly.
At this point, the NUREG/CR-1580 was issued for comment in July 1980.
Essex contract was terminated. The Staff has contracted with Lawrence Livermore, who is using Bio-Technology as its human factors subcontractor to assist the Staff in resolving public comments, and develop.ng the i
effective guidelines.
The Staff Supplement to the Dra.ft Report on the Human Engineering Guide to Control Room Evaluation not only contains a response to public comments, but identifies additional items not in the draft quidelines, such as, identifying information an operator may need to know for some process The Supplement will be out for public that is not in the control room.
The effective guidelines will be issued comment until the end of April.
l by the end 'of May provided public comments are not too extensive.
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Mr. Froelich described the control room guidelines' development.
i began with informal guidelines and checklists in NUREG/CR-1270 develo Draft guidelines in NUREG/CR-1580 were sent out l:
by Essex Corporation.
Concurrently, the Staff was increasing the degree
' i for public review.
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Public was issued as a result of public comments on NUREG/CR-1580.
meetings to discuss NUREG-0659 will be held on April 22nd and 24th.
Additional public comments and Staff comments will be factored into NUREG-NUREG-0700 which will contain control room evaluation criteria.
0700 will be ready for publication by the end of tiay.
Mr. Froelich explained that the purpose of NUREG/CR-1580 was to' produce a set of instru'ctions for a utility to identify operator / control room interfaces, compare them with a set of standar( criteria or guidelines applicable to control rooms, and on the basis of the comparison, uncover any human engineering deficiencies.
The guidelines in NUREG/CR-1580 were broken into a number of sections which include control room environment, visual displays, controls, workspace, auditory displays, control / display integration, performance aids, and communication.
It did NUREG/CR-1580 was characterized as basically hardware oriented.
not identify the operators' role in the control room, or identify " missing" l
There was no guidance for integration l
(other) or unneeded components.
with other Task Action Plans.
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FOIA EXEMPTION (b)5 MARCH 9 & 10, 1981 REACTOR OPERATIONS MARCH 9, 1981 As a result of the public comment period on NUREG/CR-1580, the Staff summarized, for the Committee, some of the comments they received as follows:
1.
the guide was too detailed 2.
data requirements are excessive many guidelines were irrelevant guidelines were not specific enough for nuclear power plants 3.
more guidance was needed to help utilities prioritize the dis-4.
5.
crepancies found between guidelines and the control room Mr. Beltracchi discussed systems reviews, which are a part of the control room design r'eview. These will serve as the bcsis of the functional review of operation and will help to integrate several items in the Task Action Plan such as procedures upgrading, the use of a safety parameter display system, and the display of post-accident monitoring equipment (Reg. Guide 1.97).
In systems review, the scope will encompass a review of the control room l
operators' tasks for events like anticipated operational occurrence and postulated energency conditions.
Mr. Beltracchi discussed the phases of the control room design review.
The process begins with identifying operating events with the emphasis The abnormal operation -
on " abnormal and/or emergency operations."
l Licensees-studies should be coupled with postulated multiple failures.
will be asked to define and document all systems and subsystems to facil-For itate the definition of functions associated with operating events.
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FOIA EXEMPTION (b)5 MARCH 9 & 10, 1981 REACTOR OPERATIONS MARCH 9, 1981 each operator function, the guidelines will suggest analyzing the function This type of study by breaking the function down into operator tasks.
Reviewers could then is aimed at revealing the man-machine interface.
proceed with verifying tasks and validating procedures.
When NUREG-0700 is released, the review it describes will have an assessm design, and implementation section, which will describe how a particular proposed improvement is correlated with operator training requirements, and assurances are reached that a particular improvement does not itself create another human factor's deficiency.
Mr. Moore indicated that the job of planning the tasks to be performed in This takes into improving control rooms could be completed in one year.
He said much of the effort account the current work load on utilities.
can be done on a generic (non-site), vendor-related basis.
Mr. Moore noted that he would be pleased to accept comments on the draft NUREG-0659 for the next 6-8 weeks from the Committee or its c Mr. Mathis requested those present on the Subcommittee to send any com-ments to Richard Major, who will compile and forward them to Voss Moore.
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MARCH 9 & 10, 1981 8-REACTOR OPERATIONS MARCH 10, 1981 (Congressman Udall's ATWS Concerns Prompted by the Browns Ferr to Scram, June 28,1980)
In his introductory statement, Mr. Mathis, Subcommittee Chairman, went over He noted the Subcommittee was meeting with the the purpose of the meeting.
NRC Staff to begin its review of Rep. Udall's inquiries on ATWS, which were Mr.
prompted by the June 28, 1980 Browns Ferry 3 partial f4flure to scram.
Mathis traced the chronology of the correspondence between Rep. Udall and l
He summarized Mr. Udall's concerns as follows:
the Commission.
An indication of the level of confidence placed on the Staff's ability 1.
to calculate the consequences of an ATWS.
The level of confidence in and the adequacy of actions taken subsequent 2.
As a second part to this to the Browns Ferry 3 control rod failure.
i concern, what additional ATWS-related concerns does the Commission and the ACRS deem appropriate to consider?
The extent to which emergency procedures at operating plants contain l
3.
l nstructions for the operators given an ATWS.
i An assessment of the causes of the Browns Ferry 3 partial failure 4.
to scram.
ACRS review of previous Commission responses to Mr. Udall's inquiries.
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MARCH 10, 1981 Mr. Mathis concluded his opening remarks by noting he hoped to be in a position where the Subcommittee could at least begin formulating a response to Congressman Udall on summary items 2, 3, and 4 mentioned above.
Mr. Check, Office of Nuclear Reactor Regulation, introduced the Staff's presentation for the day. He explained the purpose of the day's presenta-tion was to describe how the NRC dealt with the partial failure to scram event that occurred at Browns Ferry 3.
No tingle element of tre Staff Several offices have been had an exclusive responsibility for the issue.
He noted that, initially, following an event at an I
heavily involved.
operating reactor, the Office of Inspection and Enforcement has the lead In this case, I&E had ad hoc assistance provided by NRR and agency role.
Mr. Check noted that the Staff will be trying to convey a sense of AE00.
the process by which an operating event is handled.
Mr. Panciera presented the overall chronology of Staff actions taken He gave subsequent to the Browns Ferry.3 partial failure to scram.
the Subcommittee a perspective on how the Staff responded to the event I
Mr. Panciera and the time sequence in which the response occurred.
noted that, before the Browns Ferry 3 event I&E, as a result of severa,1 e
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. MARCH 9 & 10, 1981 REACTOR OPERATIONS MARCH 10, 1981 LERs, had been locking into the problem of level indication in the scram discharge volume (SDV). Both Brunswick and Hatch plants had damaged A bulletin was issued in floats on their SDV level indicating instruments.
response to these events. Mr. Panciera described the near term re-sponse to the accident and noted the early involvement of both NRR and AEOD in the analysis of the incident.
In addition to the near term response, which was the issuing of I&E bulletins, Mr. Panciera described the establishment of a multi-disciplined team to work on the long range The involvement of both Staff and the BWR owners solutions to the event.
The evolution was described during the evolut' ion of the long term fixes.
of the reports by AE00 and the evaluations contained in the generic SER were explained, as well as, the issuance of additional supplements to the
^* riose coupling of effort between the various Staff I&E bulldx offices and BWR owners groups was stressed.
I Mr. Mills, Office of Inspection and Enforcement, discussed the scram l
system design for a boiling water reactor, the sequence of events for j
the Browns Ferry 3 partial failure to scram, and identification of the Mr. Mills explained that when a scram signal is received, causes.
Water the scram outlet valves open just before the scram inlet valves.
a' present in the area above the control rod drive piston is vented and
MARCH 9 & 10, 1981 REACTOR OPERATIONS MARCH 10, 1981 higher water pressure is applied to the bottom side of the piston, As the control rod is driven driving the control rod into the core.
into the core, water is displaced from the area above the drive piston About 3/4 of a gallon of and flows into the scram discharge volume.
water is displaced above each control rod drive, the scram discharge Once a scram volume is sized to receive about 31/3 gallons per drive.
is initiated, seal leakage around each control rod drive mechanism Reactor water, from the vessel, through the seals, flows into occurs.
If the scram signal is not reset within the the scram discharge volume.
a few minutes, the scram discharge volume will fill with water and pres-When an operator takes manual surize to the reactor vessel pressure.
action to reset a scram, this closes the scram inlet and outlet valves The system will drain and be and opens the vent and drain valves.
capable of receiving water from the next scram.
Browns Ferry 3 was in the process of shutting down for On June 28, 1980, The power level of Browns routine maintenance on the feedwater system.
On the first manual scram, control rods on Ferry 3 was approximately 35%.
the west side of the reactor core fully inserted, however, 75 control The reactor power leve'l was rods on the east side did not fully insert.
The operator reset the scram signal and tried a.
reduced to about 2%.
second manual scram.
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. MARCH 9 & 10, 1981 REACTOR OPERATIONS MARCH 10, 1981 The rods on the east side moved an average of 12 inches, 33 rods were The operator again reset the scram signal, and a third fully inserted.
manual scram was executed, rods on the east side moved an average of 7 1
inches, 47 rods fully inserted. Once again the operator reset the scram The reactor automatically scrammed on a high scram discharge signal.
It took 14 minutes, from volume level signal, all rods fully inserted.
the start of the sequence (the first scram attempt), until the control rods were fuily inserted.
Following this incident, support people were inmediately called to the site to begin an investigation to determine the The testing and evaluation performed on the cause of the partial scram.
day of the event included a check of the valve alignment on the control Correct alignment was verified.
The east bank vent rod drive system.
valve operability was also verified.
A survey of the drain lines that connect the east scram discharge volume to the instrument volume was conducted, no indication of blockage was found. A survey of the drain Again, the purpose of the survey was to inspect for sumps was conducted.
foreign objects or debris that.might have been an indication of blockage.
No indication of blockage was found. A calibration check was performed Two on the level switches on the scram discharge instrument volume.
Two level switches were out of calibration during problems were found.
However, during the event itself, these switches did actuate.
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MARCH 10, 1981' An evaluation of the electric'al systems, which initiates a scram, was Based on a number of tests, it was clear that an electrical conducted.
malfunction could not have created the west only scram.
I&E concluded it was the retention of water in the east scram discharge The cause of the water retention volume that caused the partial scram.
The Staff concluded that generic action was required was not determined.
This action consisted of verifying at immediately following the event.
BWRs that the scram discharte volume was f:lly operable and that the scram discharge volume is periodically checked to make sure it is empty.
This was the basic philosophy behind the bulletins issued shortly after the Browns Ferry 3 event.
Mr. Rubin of AE00 discussed that office's investigations and activities Within a few days since the Browns Ferry 3 partial failure to scram.
after the Browns Ferry event, AE00 technical representatives went to the site as part of an NRC team to begin to gather information about the event, the scram system design, and results of systems tests and With this initial contact, AE00 began its inspections performe, by TVA.
t own individual investigation of the B.rowns Ferry event; its causes,and lessons learned.
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MARCH 9 & 10, 1981 REACTOR OPERATIONS MARCH 10, 1981 AE00 concluded as did the other offices, that the reason for the Browns Ferry 3 partial failure to scram was due to the scram system hydraulics.
The observed rod me, tion was best explained on the basis of at least a Tests performed by GE at partially filled scram discharge volume.
San Jose showed that the drainage rate of the east scram discharge volume was consistent w th the average rod motion that occurred during the scram i
9 attempts. The rod motion was consistent with expectations, given the amount of free volume made available by drainage.
Mr. Rubin explained how it was possible for water to accumulate in the in the scram discharge volume header and not scram the reactor as a result of high water level in the scran instrument volume (SIV).
The vent line between the scram discharge header and the instrumented volume Tests were done to drops only 1 ft. 7 in. over a total length of 150 ft.
determine the draining characteristics of both the east and west scram Both headers were filled with room temperature discharge volume systems.
At time zero, the instrumented volume water with the vent valves open.
The was opened and both headers were allowed to drain simultaneously.
The east header took 25 min. before it f
west header emptied in 9 1/2 min.
The slower drainage' characteristic of the east he'ader finally emptied.
allowed the instrumented volume to drain, clearing high water indicatica l
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. MARCH 9 & 10, 1981 REACTOR OPERATIONS MARCH 10, 1981' instruments before the east bank had drained. AE00 concluded that the drainage rate of the west header is about 35 gpm, and the east header drains at about 116 gpm, while the average drain rate of the instrumented volume, based on the rate of clearing the instrument switches is about 24.5 gpm.
Mr. Rubin summarized the findings of the study by AE00. The findings are as follows:
Water in the east scram discharge volume header is the likely cause 1.
of the scram system failure.
The scram instrument volume high level scram function did not pro-2.
vide protection against an accumulation of water in the east scram discharge volume header, for normal venting and draining of the he ade,r.
A single blockage could disable, both east and west discharge 3.
volume header protection, if the faster draining line is plugged, its lack of contribution would prevent the instrumented volume from filling and giving a scram signal.
With the current scram discharge volume design, a blockage in 4.
the vent or drain path can cause water to accumulate and at the same time disable the protection function.
The current scram discharge instrument volume results in the 5.
automatic high level scram safety function being dependent on the nonsafety related reactor building clean radwaste drain Venting of the instrumented volume is controlled by system.
the downstream clean radwaste drain system characteristics.
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REACTOR OPERATIONS MARCH 10, 1981 There are numerous sources of water which can fill the scram Possible sources 6.
discharge system if drainage does not occur.
of water are from previous scrams, multiple scran outlet valve leakage, and injection from the scram discharge volume flush Mechanisms which can trap water in the scram discharge volume include blockage of vent piping; a plugced line leading lines.
from the scram discharge volume into the scram instrumentedThe sp volume; or a closed vent valve.
caused the Browns Ferry 3 problem is not known, however, enough mechanisms can be postulated to cause a concern.
Float type level switches have been unreliable.
7.
Re-scram attempts are not always possible.
8.
A blowdown can occur outside of primary containment, if a vent or 9.
drain valve remains open during a scram.
There was no emergency procedures for a scram system failure 10.
at Browns Ferry 3.
There were five recommendations resulting from the AEGO initial invest These recom-gation into the Browns Ferry 3 partial failure to scram.
mendations were:
SDV system protective function should not depend on vent orThis 1.
drain arrangements.new design requirements which will combine the scram di volume and scram instrumen*.cd volume.
Provide diverse water levei monitoring instruments.
8 2.
This Provide redundant SDV system vent and drain valves.
recommendation is to provide extra protection to prevent 3.
an unisolatable blowdown outside primary containment.
Provide emergency operating procedures for scram system 4.
failures.
Consider improving SDV system drain reliability.
5.
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REACTOR OPERATIONS MARCH 9 & 10, 1981 MARCH 10, 1981 These Mr. Rubin presented the four conclusions to the AE00 investigation.
conclusions were:
Water in the east SDV caused the Browns Ferry 3 scram failure.
1.
The current scram capability protection system is unacceptable.
2.
Given a single valve failure, the potential exists for an uniso-3.
latable blowdown outside containment.
The SDV system will require modification to reduce ATWS risk.
4.
Mr. Rubin discussed an AE00 study on the potential for control air and scram system interaction. The concern arists as a result of the effects of degraded air on the scram outlet valves.
If the control air pressure were to drop slightly below 45 psi, the scram outlet valves could par-tially open.
Flow passing the control rod drive seals, at the rate of 1-2 gpm, could occur without significant rod motion for a partially opened scram outlet valve. The accumulated flow rate of 93 scram outlet valves could be in excess of the drainage rate of the scram discharge volume header. Water could accumulate in the header as the result of this degraded air situation. As the SDV headers fill with water, and automatic scram has not occurred due to poor hydraulic cou-pling between the SDV and SIV, a concern is raised for approaching a' situation in which the reactor cannot scram. At the same time, the degraded control air supply would be adversely affecting other regulating valves in the plant, e.g., the feedwater system. Thus, a plant transient 1
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MARCH 9 & 10,1981 REACTOR OPERATIONS MARCH 10, 1981 such as a water level drop in the reactor, could also be initiated even-This scenario considered in its entirety tually leading to need to scram.
shows the plant within a few minutes evolving towards and ATdS, because of degraded air conditions.
Mr. Rubin explained tnat in the long term, the combination of the SDV and In the SIV will alleviate concerns over loss of control air situations.
the near term, additional surveillance and level indication is being re-Bulletins have required operators to quired for th'e SDV and SIV systems.
An manually scram the reactor on indications of loss of control air.
automatic scram on degraded control air was required by the Staff.
Mr. Graves discussed a BWR plant transient analysis program performed at Mr. Graves noted that in recent years Brookhaven National Laboratory.
the NRC in conjunction with technical assistance from BNL, has developed a reasonable capability of analyzing the consequences of a full ATWS.
This capability has been used in calculations for selected ATWS events in The calculations had been performed to improve the BWR-4 type plants.
Staff's understanding of the consequences of ATWS events and to formulate the Staff position with respect to the AT45 rule now under consideration.
Following the Browns Ferry 3 partial failure to scram, the Staff asked ~
Brookhaven to conduct some calculations of the consequences of a similar This program will be conducted during FY 81 and FY 82.
incident.
MARCH 9 & 10, 1981 REACTOR OPERATIONS MARCH 10, 1981 A second part of this program is for BNL to prepare input tapes that model typical BWR 3s, 5s and 6s for ATWS and other transient conse-quences.
The objectives of the BNL program are to develop the capability to audit vendor and licensee analyses, develop capabilities to perform confirmatory analyses of all BWRs to determine the safety impact of operating transients and to provide a better basis for decisions involving operating reactors.
Further objectives are to develop a better understanding of the transient /
accident behavior response of BWRs for developing emergency guidelines and plant operating procedures, and independent audit assessment of the For the case of the main steam isolation adequacy of safety features.
valve closure plus a Browns Ferry 3 partial scram at full power, the Mr.
critical parameter became the heat load to the suppression pool.
Graves noted that a full ATWS would be a much more severe event than a He noted that, in the past, a series of partial failure to scram.
calculations had been performed, which would show the consequences of a full ATWS and the results of fixes intended to correct the situation.
The calculations performed for a partial failure to scram as occurred at Browns Ferry 3 indicated that consequences were mild enough so that operator action could be taken in order to avoid serious consequences.
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MARCH 9 & 10, 1981 MARCH 10, 1981 Mr. Mills discussed the short term actions necessary to justify continued operation of BWRs and long term actions needed to provide the SDV design with improved reliability. He also discussed BWR ATWS-related procedures and modifications. He noted that I&E Bulletin 80-17 was sent from IaE The main thrust Headquarters within 5 days after the Browns Ferry event.
of this bulletin was to keep the scram discharge volume empty and operable.
Mr. Mills described a number of deficiencies that were discovered in scram systems' as a result of a response to Bulletin 80-17.
Some of the deficiencies discovered included crushed' floats in the high level alarm and rod block instruments and scram discharge volume high level scram instruments.
(These events were discovered before the Browns Ferry 3 Additional deficiencies included scram discharge volumes incident.)
that did not drain properly at several plants. Mr. Bender asked for additional information to clarify how these deficiencies might affect a Mr. Mills discussed some of the requirements of Browns Ferry type event.
Bulletin 80-17, Supplement 1, and noted that, among other items, this Supplement required that a continuous monitoring system be installed in the scram discharge volume by September 1st of 1980. The Supplement also required a design review of the vent system by September 1st, and proce-dural controls for the availability and use of the standby liquid contro'1 systems.
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REACTOR OPERATIONS MARCH 10, 1981 Bulletin 80-17, Supplement 2 required, "each BWR with a scram discharge volume vent system, that depends on any component other than the vent valve alone for prope venting, must provide an alternative vent path continu-ously open to the building atmosphere within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of notification to It must be positive in its functions at all times."
continue operation.
This requirement was to eliminate the potential that a vent problem would As a result result in retaining water in the scram discharge volume.
of this requi.rement, about 15 plants modified their vents.
Supplement 3 to Bulletin 80-17 $as issued after the concern raised by AE00 concerning a loss of control air effect, that could result in the loss of scram capability. Supplement 3 requires the operators to manually scram I
in the event of loss of control air.
The responses to the bulletin sprolement requiring a schedule for installa-Confirmatory tion of a continuous monitorsng system were not definitive.
orders were issued which required all plants to begin installing by December 1st a continuous monitoring system to detect the presence of water in the scram discharge volume.
I&E Bulletin 80-17, Supplement 4 required in-place operability testing of I :
!I The Bulletin the continuous monitoring system with water in the the 50V.
also required periodic surveillance testing of the continuous monitoring This Bulletin corrected some installation problems that had been system.
i discovered in the continuous monitoring systems.
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MARCH 10, 1981 One of the items in Bulletin 80-17 was a requirement for plants to have I&E conducted a survey of all operating ATWS operating procedures.
reactors and procedures were inspected for their acceptability with Mr. Mills reported that all BWRs regards to responding to ATWS events.
Mr. Mills concluded that as were found to have acceptable procedures.
result of the Bulletin, corrective actions had been taken to ensure that Corrective actions taken the SDV is maintained during power operation.
are necessary,and sufficient to justify continued operation. Long term action to improve the scram system'is nece{sary and will be performed.
Mr. Schwenk of the I&E Staff discussed the results of the survey to deter-mine the adequacy of licensee emergency operating procedures to respond to an ATWS event at PWRs. Guidance was supplied from I&E headquarters to the resident inspectors in the form of guidelines to compare ATWS pro-The results of the survey indicated that 20 plants had cedures against.
^
Five plants procedures with no exceptions to the inspection requirements.
meet the inspection requirements, but did not have them labeled as specific Twenty plants had some minor exceptions to inspection ATWS procedures.
requirements.
Mr. Rubin described AE00's assessment of the interim protective measures These are the at Browns Ferry 3 required by the first I&E bulletin.
measures which were put in place a few weeks after the event and were I
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MARCH 9 & 10, 1981 -
REACTOR OPERATIONS MARCH 10, 1981 intended to provide a basis for assuring continued safe operation of the hardware modifications to the SDV system.
plant, pending long term AE0D's findings were that the then existing interim system, which in-volved a newly-installed ultrasonic water detection equipment and spe-cial procedures did not restwe the level of scram capability protection However, these temporary thought to be assured in the original design.
arrangements were felt to be adequate for sources for water which would For fast-fill involve slow wa,ter accumulation in the SDV headers.
scenarios of the SDV, and in particular for degraded control air situa-tions, the interim measures in place at the time AEOD did its review were considered to be less than adequate. Accordingly, AE00 made a recom-mendation for an imrnediate, manual scram required solely on the basis of AE00 also an indication of low control air pressure in the control room.
felt it was necessary to move the UT monitoring indication into the AE00 felt that con-control room to improve operator response time.
l sideration should be given to providing an automatic scram signal based on degrading control air pressure.
A number of Mr. Panciera discussed the development of the generic SER.
The objective of regional meetings were held in July and August of 1980.
~
these regional meetings was to obtain an in-depth understanding of the as-built conditions of scram discharge volume, instrumented volume, interconnecting piping, and vent and drain systems.
--_r
MARCH 9 & 10, 1981 24 -
REACTOR OPERATIONS MARCH 10, 1981 Mr. Panciera noted that a BWR Owners Group was formed around the 20th of August 1980 to address the problem with the hydraulics of the scram By the 19th of September, the Owners Group had de-discharge volume.
veloped criteria that they felt addressed all the problems that were uncovered by the Browns Ferry event and the problems uncovered at Hatch and Brunswick concerning crushed scram level indicating instrument floats.
The Owners The Staff reviewed the early criteria and made comments.
As the result Group submitt,ed a revised set of criteria on October 15th.
of two rounds of review, the Staff's SER, in effect, endorses the criteria developed by the BWR Owners.
Mr. Panciera discussed the sections of the SER which addressed justi-The justi-fication for continued operation and the long tcrm program.
fication for continued operation was based on an evaluation of the licensee is compliance with Bulletin requirements, recommended short term The Staff addressed the modifications, and plant-specfic modifications.
fast-fill of the SDV scenario that could result from a degraded control supply air by a requirement to automatically scram the reactor when air pressure reaches 10 psi above the scram discharge volume actuation A
setpoint or if other indications of degraded air become apparent.
specific requirement to install an air dump valve which will automaticaTTy scram the reactor when pressure reaches a point of 10 psi above the scram O
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MARCH 9 & 10, 1981 REACTOR OPERATIONS MARCH 10, 1981 outlet valve setpoint pressure was an interim measure contained in the The basis for continued operation was satisfaction of all the SER.
Bulletin requirements as well as implementing the short term modification of an air dump valve. Once an improved hydraulic coupling in the scram discharge volume and scram instrumented volume is attained there will be no need for such a scram. Mr. Panciera discussed long term requirements.
~
The functional criteria for the scram discharge volume was to have sufficient ca,pacity to, receive and contain water exhausted by a full reactor scram without adversely affecting control rod drive scram per-Mr. Panciera also discussed safety criteria for the long term formance.
fixes. There are 5 such criteria, including:
No single active failure of a component or service function 1.
shall prevent a reactor scram under the most degraded conditions that are operationally acceptable.
No single active failure shall permit uncontroled loss of coolant.
2.
The scram discharge system. instrumentation shall be designed 3.
to provide redundancy to operate reliably under all condition, and shall not be adversely affected by hydrodynamic forces or flow characteristics.
Syst:m operating conditions, which are required for a scram, 4.
shall be continuously monitored.
Repair, replacement, adjustment, or surveillance of any system -
5.
component shall not require the scram function to be bypassed.
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five criteria were:
Level instrumentation shall ta desigred to be maintained, 1.
tested, or calibrated during plant operation without causing a scram.
The system shall include sufficient supervisory instrumentation 2.
and alarms to permit surveillance of system operation.
The system shall be designed to minimize the exposure of. operating 3.
personnel to radiation.
Vent path's shall be provided to assure adequate drainage in 4.
preparation for scram reset.
- Vent and drain functions shall not be adversely affected by 5.
The objective of this requiremer.t other system interfaces.
is to preclude water backup in the scram instrument volume which could cause spurious scrams.
These Mr. Panciera explained the design criteria specified in the SER.
criteria established the need for good hydraulic coupling between the Mr. Panciera also scram discharge volume and the instrumented volume.
explained the surveillance criteria for the scram discharge system.
Mr. Pittman of the Division of Systems and Reliability Research Staff dis-Both cussed the review of the two basic BWR designs performed by that group.
the single instrument volume and the dual instrument volume (coupled SDV and SIV system) were reviewed with an eye toward additional improvements which v,,e
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m
. MARCH 9 & 10, 1981 REACTOR OPERATIONS MARCH 13,19[1 These Also included in the SER were a number.of operational criteria.
five criteria were:
Level instrumentation shall be designed to be maintained, 1.
tested, or calibrated during plant operation without causing a scram.
The system shall include sufficient supervisory instrumentation 2.
and alarms to permit surveillance of system operation.
The system shall be designed to minimize the exposure of operating 3.
personnel to radiation.
Vent paths shall be provided to assure adequate,drair. age in 4.
preparation for scram reset.
Vent and drain functions shall not be adversely affected by 5.
The objective of this requirement other system interfaces.
is to preclude water backup in the scram instrument volume which could cause spurious scrams.
These Mr. Panciera explained the design criteria specified in the SER.
criteria established the need for good hydraulic coupling between the Mr. Panciera also scram discharge volume and the instrumented volume.
explained the surveillance criteria for the scram discharge system.
Mr. Pittman of the Division of Systems and Reliability Research Staff dis-cussed the review of the two basic BWR designs performed by that group.
Both the single instrument volume and 'the dual instrument volume (coupled SDV and SIV system) were reviewed with an eye toward additional improve-ments which could be made in future plants. Basically, the findings from DSSR supported those earlier findings of AE00 and NRR, highlighted in te Mr. Pittman reiterated that the new design provides an instrumented SER.
l I
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MARCH 9 & 10,1981 REACTOR OPERATIONS l
MARCH 10, 1981 volume attached directly to the headers of the scram discharge volume.
This design eliminates the hydraulic coupling between the header and the instrumented volume.
Mr. Rubin described additional AE00 BWR scram system investigations.
Since the Browns Ferry 3 case study, AEOD has extended their initial review to include a more thorough study of the safety concerns associated with single passive failures, i.e., pipe breaks in the scram. discharge The two points under review during this extension of the volume system.
study is the ability of the recto'r coolant boundary to isolate and the primary containment isolation function.
~
The Subcommittee held a brief executive session following the prepared During the executive session the Subcommittee decided to presentations.
explore the calculational capabilities of the Staff and their contractor, Brookhaven National Laboratory, during the next Subcommittee meeting.
The ability of the Staff to predict the consequences of an ATWS during both a full ATWS and a partial failure to scram will be the topics of the The meeting will focus primarily on ATWS scenarios associated meeting.
The Subcommittee agreed to meet on the afternoon of April 8, with BWRs.
1981; the meeting is to take place in Washington, DC.
The meeting was adjourned at 5:45 p.m.
For additional details, a complete transcript of the meeting is availabl in the NRC Public Document Room,7th St., SW,, Was,hington, DC 1717 H St.
NW Washington DC 20 NOTE:
554-2345.
or from Alderson Rep,orters, 300 I
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- 20. 1961 / Notices 13132
~
Feder:I Register / Vcl. 46. No. 34 / Friday. Februtry cf the meeting & a transcnpt to being' /
dvisory Committee Assetor ASafeguards; Sut> committee on Virgil C.
F. 'NRC DegradedCore kept. and questions may be asked only Summer Nuclear Station; Change of Ru/emaAmp--ACRS position / action re by members of the Subcommittee.Its Location consultants. and Staff. Persona desiring proposed rule.
G. *t.' orth Anna Nuclear Power to make oral statements should notify The ACRS Subcommittee on Virgt! C.
Station Umt 2-decay heat removal the Designated Federaftmployee as far Summer Nuclear Station wi!! hold a s3 stems.
fp ae to n$a e'a'tra 4em 9
ts can made e
ocat on
'""o-
- naaa'r "== d=rias *
- cavuounn =ns2-s1==
r andIAe ED m ssio NR AsumMy Strut.Columba. SC dM"n'to2;e!','e'nd""3MS d
8t8temen*
ng Thursday. February 26.1981 at 100 p.m.
l recommendations on the safety research he entire meeting will be open to and Friday. February 27.1961 at 8.30 public attendance except for those a.m. until the conclusion of business program. proposed revision of the fission product source term used in the sessions during which the Subcommittee each day instead of the Holiday Inn-siting and design of nuclear plants. and Tmds it necessary to discuss proprietary No-thwest. US-1 & I-26. Columbia. SC.
j ACRS comments on the proposed NRC and Industrial Security information. One All other items regarding this meeting Long Range Safety Research Program.
or mon closed sessions may be remain the same as announced in the
- 1.
- Reports of ACRS Subcomimitees necessary to discuss such infonnation.
Federal Register published Wednesday, ond Alembers-etatus of ACRS review (Sunshine Act Exemption 4.)To the February 11.1981.,
ci matters such as the proposed astent practicable, these closed sessions Further infmmation may be obtained prelicensing review of a passite w211be held so as to minimize by a prepaid telephone call to the containment system. status of generic inconvenience to members of the public cognizant Designated Faderal Employee for s meeung. Mr. PauMoehnert items applicable to LWRS. control in attendance.
system failures that could cause or De agenda for subject meeting shall (te!ephone 202/ca-3267) between 8:15 exacerbate nuclear power plant be as foHows:
eccidents. and the ATWS resolution IDM8#
- Da February 7 en.
8'C lohn C Hode.
Y., ',;
,_,_,,,y gy, Adr:sory ommstree Management O&er en er Ian Bro n
- l.
- Reports and briefi:'gs by y
representottves of the Department of
,,chs M & s n on es-rut suw us a Energy (DOEJ ond NRC- -application of Durms the mitialportion of the meettr4
,,g,,,,,,,3,,,,,
Three Male Island Unit 2 lessons learned
&e Subcommutu. along wnh any of na to DOE fac21ities and an independent consuhants do may be pnwat.wul IOocketNo.50 2551 design review program implemented by exchange prelimmary views ressed:ng be conssdered dunns the balance gg ma Amendment to Provisional Operating K. *Fissjon hoduct Source Term-.-
The Subcommirtw win then hur Ucense ACRS review of the fission product the siting and design presentations by and hold discussions with The U.S. Nuclear Regulatory
((]u representatives of the NRC Staff. their Commission (the Commission) has Aptd s-f J. Issf Agenda to be c nauhants and otherinterested persons issued Amendment No. 64 to Provisional announced.
MS*
- 8th28 M N *-
Operating License No. DPR-20. Issued to Afoy 7-9.1981: Agenda to be Funhet information regarding topics Consumers Power Company (the announced.
to be discussed.whether the meeting licensee). which amended the license for D.ted. February 13.1961 bas been cancelled or rescheduled. the f,
y')
- * *
- fp e
7,3 r onn ujsts o Van Buren County Michigan.ne John C % ).
a s
.Adusory Commarter Manopement Q!ficer and the time allotted therefor can be amendment is effective sa of its date of
. obtained by a prepaid telephone can to I"fhe amendment revises Paragraph 3 E the cognizant Designated Federal of the license to incorporate Supplement Employee. Mr. Richard K. Major yAdvisory Committee Reactor (te:ephone 202/634-1414) between 8.15 Nos.1 and 2 to the Fire Protection Safegaards; Subcommittee on Reactor e.m. and 5.00 p.m EST.
Safey Evaluation.
The filings comply with the standards h
Operations; Meeting I have determined. in accordance witand requirements of the Atomic Energs The ACRS Subcommittee on Reactor Subsection 10(d)of the Federal Act of 1954, as amended (the Act), and Operations will hold a meeting on Advisory Committee Act, that it may b e
the Commission's rules and regulations.
Monday and Tuesday. March 9-10.1981 necessary to close some portions of thi The Commission has made appropnate s
at 1717 H Street. N.W., Washington. DC meeting to protect proprietary and findings as required by the Act and the in Room IM6.The Subcommittee will Industrial Security information. The Commission's rules and regulations in to e
beer a briefing on the NRC StafTs authority for such closure is Exemption CFR Chapter L which are set forth in the s
Ituman Engineering Guide to Control (4; to the Sunshine Act. 5 U.S.C.
license amendment. Prior pnblic notice s
Room Evaluation and wiU begin work 552b(c)(4).
of this action was not nquired since the on a response to recent inquiries by Congressman Udall on ATWS.
Dared. Febraary 13.1981.
amendment does not involve a significant hazards consideration.
in accordance with the procedures I*h" C D The Commission has determined that outlined in the Federal Register on Ad"8 cry Committee Menegement O@cer the issuance of this amendment will not October 7.1980 (45 FR 66535). ora; or p on s:-ru m um os ami result in any significant environmental nritten statemer.ts may be presented by siamo coot rsawus impact and that pursuant to 10 CFR members of the public, recordings will 8
be permitted only during those portions P90R 031GINAl.
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TENTATIVE SCHEDULE FOR THE MARCH 9 & 10, 1981 ACRS SUBCOMMITTEE MEETING ON REACTOR OPERATIONS 1717 H ST., W WASHINGTON, DC MONDAY, MARCH 9,1981 APPROXIMATE TIME _
I.
OPENING REMARKS 1:00 p.m.
Discussion of Schedule a.
b.
Meeting Goals BRIEFING BY THE DIVISION OF HlMAN FACTORS HUMAN ENGINEERING GUIDE TO CONTROL ROO II.
1:15 p.m.
TOPIC:
EVALUATION (NUREG/CR-1580)
Detailed agenda to be provided by Division of ihman Factors Safety RECESS 5:00 p.m.
9
,I. -
TENTATIVE SCHEDULE 4
ACRS SUBCOMMITTEE MEETING ON REACTOR OPERATIONS TUESDAY, MARCH 10, 1981 APPROXIMATE TIME 8:30 a.m.
1.
OPENING REMARKF 8:45 a.m.
2.
OPENING STAFF COMMENT (P. Check) 8:50 a.m.
3.
CHRONOLOGY OF STAFF ACTIONS (V. Panciera) 9:10 a.m.
4.
PRINCIPLES OF OPERATION AND SCRAM SYSTEM DESIGN (W. Mills) 9:25 a.m.
5.
BF-3 EVENT-SEQUENCE OF ACTIONS AND PRELIMINARY IDENTIFICATION OF CAUSE (W. Mills) 9:40 a.m.
6.
AE00 INVESTIGATION OF BF-3 EVENT (S. Rubin) 10:40 a.m.
7.
POTENTIAL FOR UNACCEPTABLE CONTROL AIR-SCRAM SYSTEM INTERACTION (S. Rubin) 10:55 a.m.
8.
IE BULLETIN REQUIREMENTS (W. Mills) 11:25 a.m.
9.
IE SURVEY OF ATWS PROCEDURES (W. Mills, G. Schwenk) 12:00 noon LUNCH 1:00 p.m.
- 10. ATWS CALCULATIONS (T. Spies) 1:15 p.m..
11.
INTERIM MEASURES AT BROWNS FERRY TO PREVENT WATER ACCUMULATION IN SDV (S. Rubin) 1:30 p.m.
- 12. DEVELOPMENT OF SER (V. Panciera) 1:45 p.m.
- 13. GENERIC SER - JUSTIFICATION (V. Panciera, FOR CONTINUED OPERATION M. Goodman) 1.
Evaluation of Bulletin Responses 2.
Short Term Modifications 3.
Human Factors Consideration 2:15 p.m.
- 14. GENERIC SER - LONG TERM (V. Panciera)
MODIFICATIONS 1.
Criteria 2.
Acceptable Means of Compliance l
3.
Implementation l
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TENTATIVE SCHEDULE REACTOR OPERATIONS TUESDAY, MARCH 10, 1981 APPROXIMATE TIME, 3:00 p.m.
15.
SCRAM RELIABILITY EVALUATION (J. Pittman) 3:50 p.m.
- 16. ONGOING AE00 SCRAM DISCHARGE VOLUME SYSTEM.,EVIEW ACTIVITIES (S. Rubin) 4:00 p.m.
- 17. OPEN EXECUTIVE SESSION ADJ0URNMENT l
b y_-c
.-.m_,
6 LIST OF DOCUMENTS PROVIDED AT MEETING 1.
Tentative Meeting Schedule Slides used by V. Moore on Development Guidelines and Criteria 2.
for Control Room Design Review (7 slides)
Slides used by R. Froelich, Guidelines Development - Control 3.
Room Design Review (22 slides)
Slides used by L. Beltracchi, Systems Review (5 slides) 4.
Advance Draft Copy of NUREG-0659. " Staff Supplement to the Draft 5.
Report on Human Engineering Guide to Control Room Design" Slides used by V. Panciera, Chronology of Staff Actions (2 slides) 6.
Slides used by W. Mills, Browns Ferry 3 (Partial Scram - June 28,1980) 7.
(11 slides)
Slides used by C. Graves, BWR Plant Transient Analysis Program 8.
at Brookhaven National Lab. (7 slides)
Slides used by W. Mills, I&E Bulletins (13 slides) 9.
Slides used by G. Schwenk, Survey by Residen Inspectors to Determine 10.
the Adequacy of Licensees Emergency Operating Procedures to Respond to ATWS Events (4 slides)
Slides used by G. Schwenk, Emergency Instruction, I-4.3, Reactor Trip 11.
(4 slides)
Slides used by V. Panciera, SER Development (18 slides) 12.
Slides used by J. Pittman, Scram Reliability Evaluation (10 slides) 13.
Slides used by S. Rubin, AEOD Investigations and Activities Since 14.
the Browns Ferry 3 Scram System Failure (32 slides)
Attachment D
<v
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