ML19341A526
| ML19341A526 | |
| Person / Time | |
|---|---|
| Issue date: | 03/25/1980 |
| From: | Stello V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | |
| Shared Package | |
| ML19341A503 | List:
|
| References | |
| FOIA-80-516, REF-QA-99900400 NUDOCS 8101260227 | |
| Download: ML19341A526 (13) | |
Text
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e li,AR 2 51980 For:
The Commissioners From:
Victor Stello, Jr., Director Office of Inspection and Enforcemen
,,3.s ).D'a:ks Thru:
Executive Director for Operations
Subject:
BABCOCK & WILC0X HANDLING OF INFORMATION RELATIVE TO THE RESPONSE OF B&W PWRs TO SMALL LOCAs
Purpose:
To inform the Commission of the results of IE investigations of, this subject and the proposed enforcement action as a result of these efforts.
Discussion:
The matter of how B&W dealt with information related to the response of a B&W-designed PWR to a small LOCA has been examined by I&E during the course of one inspection and three investiga-tions, as well as by the Kemeny Commission and the NRC Special TMI Inquiry Group.
We have previously forwarded the pertinent IE reports to the Commission in a memorandum dated March 14, 1980.
It should be noted that a parallel investigation was conducted by OIA of the NRC staff's handling of the Michelson Report.
That investigation disclosed that the NRC staff received a copy from Mr. Michelson, but had not been informed of its existence or implications by B&W.
Subsequently, the Commission requested a briefing in this matter.
This paper has been prepared in advance of that briefing.
We have reviewed the results of the above inspection, investigations, and special inquiries and have reached the following conclusions:
1.
Based on the information at hand we did not find objective '
evidence that a B&W " director" or responsible officer as defined in 10 CFR 21.3(f) was in noncompliance with 10 CFR 21.61.
Contact:
W. J. Ward, IE 49-27246 i
8101260 h
2-2.
There were at least four matters related to the response of B&W-designed reactor systems to small LOCAs (see Enclosure I, Appendix A) that should have been identified and evaluated as required by 10 CFR Part 21.
The, failure ~
to properly identify'and evaluate these matters resulted in at least two cases not being reported to the appropriate NRC licensee and to the NRC.
3.
Had the generic aspects of those four matters been properly identified, evaluated, and reported, the TMI accident might not have occurred.
4.
Procedural violations relative to the handling of 10 CFR 21 matters were also identified.
In summary, it is our view that matters were identified that have a high degree of severity in terms of insights they would have provided in coping with a small LOCA at a B&W-designed ~
f aci.li ty.
We are convinced that the intent of 10 CFR Part 21 was to ensure that information of this nature would be properly identified, evaluated and reported.
In this case we have information that was identified, but was neither properly evaluated nor reported to the NRC even though some of it was subsequently characterized by a B&W supervisory engineer as being of such importance that had it been known by the proper persins, the TMI accident would not have occurred.
It should be noted that the engineer's recommendations were in part e
implemented after TMI.
This information, in concert with other items identified in the IE reports, cast serious doubt upon the efficacy of B&W's implementation of Part 21.
The above conclusions indicate that a higher threshhold enforcement action should be imposed.
Thus we believe that a civil penalty is indicated on the basis that the safety informa-tion improperly handled by B&W 1ed to an event that could have been prevented, and that some of the,information has subsequently led to the implementation or development of corrective actions. -
We do not believe that other sanctions would be appropriate at this time.
Based on the above, we are proposing a civil penalty in the amount of $100,000.
The draft civil penalty package and Notice of Noncompliance are enclosed as Enclosure 1.
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l Coordination:
The Office of Nuclear Rbactor Regulation concurs in the proposed l
issuance of a civil penalty.
Alternatives-to-the proposed-action l
w4ll-be-discussed at-theiom:nissit>n-beief4cgr 1
Victor Stello, Jr.
Director Office of Inspection and Enforcement
Enclosures:
1.
Draft Letter to B&W with Appendix A - Notice of Noncompliance and Appendix B, Notice of Proposed Imposition of Civil Penalties Distribution:
V. Stello, IE R. C. DeYoung, IE NRR ELD T. W. 8rockett, IE W. J. Ward, IE D. Thompson, IE H. D. Thornburg, IE L. N. Underwood, IE (H11-2170-H11)
G. Ertter (E00-8233)
IE Files IE Reading E00 Reading
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DD:IE D:IE EDO RCDeYoung VStello 3/ /80 3/ /80 3/ /80 I
WPU:SM:DM Office X005:IE X005:IE X005:IE RCI:IE NRR ELD 3/20/80 Surname TWBrockett WJWard DThompson HDThornburg JOB A(2) Date 3/ /80 3/ /80 3/ /80 3/ /80 3/ /80 3/ /SE
The Babcock & Wilcox Company Nuclear Power Generation Station AlTH:
Mr. J. H. MacM'llan Vice President Post Office Box 1260 Lynchburg, Virginia 24505 Gentlemen:
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We have completed our investigations into the matters related to the review, evaluation, and reporting of knowledge and information related to factors that may have contributed to the TMI-2 accident and available to B&W before and after the accident.
We have also reviewed the relevant information generated by the President's Commission on Three Mile Island and NRC Special Inquiry Group.
On the basis of our review we have concluded that B&W did
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not and does not yet have an effective system for collection, review and
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evaluation, and reporting of crucial safety information.
Further, we believe that the B&W procedures actually did impede this process.
I It distresses us to find that senior B&W technical personnel stated that some of these matters should not be reported to NRC licensees or to the NRC becaus'e of fear of overreaction on the part of the NRC.
These matters certainly were important to safe,ty based on our present insights to the TMI-2 accident and should have been reported to NRC licensees and to the NRC in accordance with 10 CFR part 21 and other of the Commission's reporting requirements.
.a s As indicated above, your organization failed to collect this vital safety information, evaluate and report on it in a responsible manner so that action could have been taken to avoid or reduce the effects of the THI-2 accident.
The
. CERTIFIED MAIL RETURN RECEIPT REQUESTED n
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,The Babcock and Wilccx Company provisions of 10 CFR 21 require evaluation and reporting of significant safety' information.
In our view your staff's application of your existing procedures and the attitudes of your senior technical personnel did not contribute to _
accomplishing that end.
Accordingli, we believe that you must provide assurance that you have developed a system for the timely and effective collection, evaluation, and dissemination of significant safety information.
This system must ensure that proper under-standing and handling of all criteria in NRC regulations associated with the,
reporting of matters affecting safe operation of nuclear power plants to the-NRC.
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Based on the results of the NRC investigations conducted on May 29, and November 6-8, 1979, an inspection conducted on September 25-28, 1979, and review of the official transcripts of sworn testimony before the President's Commission on the Accident at Three Mile Island, it appears tnat certain of your activities were not conducted in full compliance with NRC requirements, as set forth in t'he Notice of Violation, enclosed herewith as Appendix A to this letter.
You are required to respond to the Notice of Violation and in preparing your response you should~ follow the instructions in Appendic,es A and B.
AsindicatedinAppendixB,weintendtoimposecivilpena5tiesin the cumulative amount of One Hundred Thousand Dollars for the items of noncompliance.
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The Babccck and Wi.1ccx Company,
In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, code of Federal Regulations, a copy of this letter and the enclosures will be placed in the NRC's Public Document Room.
Sincerely Victor Stello, Jr.
Director Office of Inspection and Enforcement
Enclosures:
1.
Appendix A, Notice of Noncompliance 2.
Appendix B, Notice of Propos&d Imposition of Civil Penalties a
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APPENDIX A NOTICE OF NONCOMPLIANCE The Babcock & Wilcox Company Vendor No. 99900400
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Based on the results of NRC investigations conducted on May 29 and November 6-8, 1979, an inspection conducted at the Babcock & Wilcox Company on September 25 through September 28, 1979, and review of the official transcripts of sworn testim 6ny before the President's Commission on the Accident at Three Mile Island, it appears that certain of your activities were not conducted in full' compliance with NRC requirements as indicated below:
10 CFR 21.21(a) requires that "each individual, corporation, partnership or other entity subject to the' regulations in this part shall adopt appropriate procedures to (1) provide for (i) evaluating deviations or 4
(ii) informing the licensee or purchaser of the deviation in order that" the licensee or purchaser may cause the deviation to be evaluated unless the deviation has been corrected; and (2) assure that a director or responsible officer is informed if the construction or operation of a facility, or activity, or a basic component supplied for such facility or activity:
(i) Fails to comply with the Atomic Energy Act of 1954, as amended, or any applicable rule, regulation, order or license of the Commission relating to a substantial safety hazard, or (ii) Contains a defect..."
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Appendix A 2-10 CFR 21.21(b) requires that each " director" or responsible officer subject to the regulations of this part or a designated person shall notify the Commission when he obtains information reasonably ind'icating a failure to comply or a defect affecting (i) the construction or operation of a facility or an activity within the United States that is subject to the licensing requirements under Parts 30, 40, 50, 70 or 71 and that is within his organization's responsibility or (ii) a basic component that is within his organization's responsibility and is supplied for a facility or an activity within the United States that is subject to the licensing requirements under Parts 30, 40, 50, 70 or 71. "
g.
B&W Administrative Procedures NPG 1707-01 (Processing of Safety Concerns) and No. 1716-Al (Reporting of Defects and Noncompliance Concerning Safety) provide the Company's process for identifying, evaluating and initiating resolution of safety concerns relating to or affecting B&W supplied components, systems, or, services, and to assure compliance with 10 CFR Part 21.
1.
Contrary to the above, in February,1978, two Babcock & Wilcox senior engineer's brought to the attention of Babcock & Wilcox management certain concerns about reactor operation procepures for certain small
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break LOCA conditions on Babcock & Wilcox reactors.
These concerns were about the sufficiency of guidance for operator actions to preclude interruption of high pressure safety injection, which could lead to possible uncovering of the core and fuel damage.
Sufficient instructions are necessary to insure proper operator actions in response to a LOCA O
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Appendix A 3-when the reactor is at full power to assure safe operations.
This absence of sufficient instructions constitutes a defect as defined in 10 CFR 21.3(d).
This insufficiency in operator instruction.was _
documented in internal Babcock & Wilcox memoranda, but was not identified as a matter which should have been evaluated and considered in accordance with procedures implementing the requirements of Part 21 for handling such defects.
This failure to consider the sufficiency of guidance for operator actions in accordance with Part 21 procedures resulted in the failure. to report'_-
such condition or circumstance to the Nuclear Regulatory Commission at the time Babcock & Wilcox became aware of the matter.
5 2.
Contrary to the above, senior engineers at Babcock & Wilcox were aware as early as November 1978, that analyses had not been performed for certain small break LOCAs using an auxiliary feedwater (AFW) level 1
control of 10 feet for the Davis-Besse Nuclear Plant as required under 10 CFR 50.46.
This lack of analysis could have created a condition or circumstance involving a basic component that could have contributed to the exceeding of a safety limit, as defined i,n the technical specifica-tions of a license for operation issued pursuant to Part 50 of this chapter.
This lack of analysis which represented a potential condition or circumstance was documented in internal Babcock & Wilcox memoranda, yet it was not identified as a technical matter which should have been i
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Appendix A -
s evaluated and considered in accordance with procedures implen.enting the requirements of Part 21 for handling su h a potential condition or circumstance.
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l 3.
Contrary to the above, senior engineers at Babcock & Wilcox were aware as early as Novamber 1978, that a small break LOCA analysis with reactor i
coolant pumps running had not been performed for Babcock & Wilcox plants.
This lack of analysis which is required by 10 CFR 50.46 is considered a defect under 21.3(d).
The lack of analysis and absence of.
conclusions needed from such an analysis to assure that current Babcock,--.
& Wilcox plants will be operated in a safe manner under normal and accident conditions by itself, represented a condition or circumstance and a defect as defined in 10 CFR 21.3(d).
This lack of analysis was documented in internal Babcock & Wilcox memoranda, yet it was not identifed as a technical matter which should have been evaluated and considered in accordance with procedures implementing the requirements of Part 21 for handling such a condition er circumstance.
U This failure to consider a matter which represent.ed a defect as defined in 10 CFR 21.3(d) in accordance with Part 21 procedures resulted in the. :.;-
.a failure to report such defect to the Nuclear Regulatory Commission at the time Babcock & Wilcox became aware of the matter.
4.
TVA transmitted a letter dated April 27, 1978, to Babcock & Wilcox (received May 3,1978) identifying a matter which represented a potential i
deviation or condition or circumstance concerning small break LOCAs at TVA's
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s, Appendix A.
Bellefonte Nuclear Plant.
Th s letter included a report by C. Michelso'n entitled, " Decay Heat Removal During a Very small Break LOCA for a B&W 205-Fuel Assembly PWR." (Michelson, Report).
This potential deviation or_
condition or circumstance was documented in internal Babcock & Wilcox memoranda, yet it was not considered in accordance with procedures implementing the requirements of Part 21 for evaluating potential devia-tions or conditions or circumstances.
Each day of noncompliance with the regulations in 10 CFR 21.21 constitutes a separateviolationforwhichacivilpenaltyisimposed(December 1,1978-"I y-c.
March 28, 1979, a total of'102 day 1).
Cumulative Civil Penalty - $510,000.
The Atcaic Energy Act of 1954, as amended, limits the total civil penalty to
$25,000 within any 30-day period, thus limiting penalties for those items cited above to 525,000 for each 30-d,ay period resulting in subtraction of $410,000.,.
Therefore a total penalty of $100,000 is proposed.
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This Notice of Noncompliance is sent to you pursuant to the provisions to Section 2.201 of the NRC's " Rules of Practice," Part 2. Title 10, Code of Federal Regulations.
You are hereby required to submit to this office, within,. -
.. s twenty (20) days of your receipt of this notice, a written statement or explana-tion in reply, including for each item of noncompliance: (1) admission or denial of the alleged item of noncompliance; (2) the reasons for the item of noncompliance, if admitted; (3) the corrective steps which have been taken and the results achiefed; (4) the corrective steps which will be taken to avoid further noncompliance; and (5) the date when full compliance will be achieved.
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Appendix 8 2-i or explanation in reply pursuant to 1 CFR 2.201, but may incorporate by specific reference (e.g., giving page and paragraph numbers) to avoid
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repetition.
l The Babcock & Wilcox Company's attention is directed to the other provisions of 10 CFR 2.205 regarding, in particular, failure to answer and ensuing orders; answer, consideration by this office, and ensuing orders; requests for hearings; hearings and ensuing orders; compromise; and collect' ion.
Upon failure to pay any penalty due, which has been subsequently determined in. --
accordance with the spplicable profisions of 10 CFR 2.205, the matter may be '
referred to the Attorney General, and the penalty, unless compromised, remitted, or mitigated, may be collected by civil action.
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APPENDIX B NOTICE OF' PROPOSED IMPOSITION OF CIVIL PENALTIES The Babcock & Wilcox Company Vendor No. 99900400 This Office has considered the enforcement options available to the NRC including administrative actions in the form of written notices of violation, civil monetary penalties, and orders pertaining to the modification, suspen-sion, or revocation of a license.
Based on these consideration's we propose to impose civil penalties pursuant to Section 206 of the Energy Reorganization Act of 1974, as amended, (42 USC 5846), and to 10 CFR 2.205 in the cumulative f;
-,.n amount of One' Hundred Thousand Dollars for the specific items of. noncompliance set forth in Appendix A to the cover letter.
The Babcock & Wilcox Company may, within twenty (20) days of receipt of this notice, pay the civil penalties in the amount,of One Hundred Thousand Dollars.
or may protest the imposition of the civil penalties in'whole, or in part, by a written answer.
Should The Babcock & Wilcox Company fail to answer within the time specifie#d, this office will issue an order imposing the civil penalties in the amount proposed above.
Should The Babcock & Wilcox Company elect to file an answer protesting the civil penalties, such answer may (a) de'ny the items of noncompliance listed in the Notice of Violation in'whole or in part, (b) demonstrate extenuating circumstances, (c) show error in the Notice of Violation, or (d) show other reasons why the penalties shoul[i not be imposed.
In addition to protesting the civil penalties in whole or in part, such answer may request remission or mitigation of the penalties.
Any written answer in Accordance with 10 CFR 2.205 should be set forth separately from the statement e
n.
x NRR response to Item 3 F01A-80-516 Davis-Besse - 1 Documentation Relating to Event of September 24, 1977 September 25, 1977 Preliminary notification of Loss of Reactor Coolant Pressure due to failure of pressurizer power operated relief valve.
October 6,1977 Memo from McDermott to Skovholt describing event of Ssptember 24 at Davis-Besse -1.
November 14, 1977 Letter from TECO enclosing LER on Event of September. 24, 1977.
November 28, 1978 Letter to Region from TECO identiffing reporting a " Reportable Occurrence" - Procedure deficiency
- controlling S.G. level in the event of a transient followed by a small break.
December 8,1978 Letter from TEC0 with LER which reported inadaquacy of station procedures to provide the complete procedures to apply when the plant would experience a. transient such as a turbine trip or loss of feedwater.
December 11, 1978 Letter from TEC0 - supporting analyses for the procedures referenced in letter dated December 8,1978.
Dec -ber 22,1978
- Additional Safety Analysis which supplements analysis in letter dated December 11, 1978.
February 26, 1979 Letter from TEC0 with Safety Analysis required by 10 CFR 50.59 which supports the modification to S.G. level setpoint instrumentation.
Memo from Vassallo to Christenberry ; Board Noti-
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March 6,1979 fication - Inspectors concerns regarding B&W plants' S
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Davis-Besse - 1 February 8, 1979 Letter enclosing start up report - Date on Turbine and Reactor Trip Test 0 40% power.
Note pressurizer level during test - Fig. 9.1.10 April 2, 1978 A turbine trip test was performed at 75% power to evablate piping modifications made on the extraction steam lines to the deaerator. The feedwater flow exceeded the feedwater demand during the runback, resulting in overfeeding the steam generators. This coupled with lifting of the pressurizer power relief valve caused a reactor trip on low RCS pressure. Reference -
Start Up Report letter dated February 8,1979.
September 2, 1977 During the initial escalation to 15% power, feedwater flow was erratic. The main feed water pump controller was placed in automatic pre-maturely.
The Steam and Feedwater Rupture Control System (SFRCS) tripped on differential pressure between steam and feedwater, leading to a reactor trip'on low RCS pressure. The excessive blowdown of the main steam safety relief valves contri-buted to the reactor trip on low pressure. All the relief valves were reset by use of a hydroset on September 16, 1977.
September 24 1977 -
With the turbine off line and the reactor at -
approximatelf 8% powe6 a " half-trip" of the SFRCS caused the 5tartispfeedwater control valves to close.
Reactor Coolant. System (RCS) pressure increased and lifted the power relief valve-on the --
pressurizer. After several cycles, this valve struck open, blowing the rupture _ disc _on the quench tank and causing a partial depressurization of the RCS. The power relief block valve was closed, and the plant was shut down for repairs.
October 23, 1977 An undetected half-tnip of the SFRCS closed the startup feedwater con
- trol valve to steam generator 1-2.
The steam generator water level decreased to 17 inches, giving a full trip of SFRCS and initiating auxiliary feedwater.
The reactor tripped on low RCS pressure as a result of the addition of 70 F auxiliary feedwater to the steam generators, and due to lif ting of the 1
pressurizer power relief valve.
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. Davis-Besse - 1 The unit was operating at 40% with the Reactor November 29, 1977 Protection System (RPS) overpower trips set at 50%. A faulty patch board was inserted into the startup test panel, producing a unit load demand signal equivalent to 50%. The plant responded to the increased demand, and the unit trippedThe on high flux when the reactor reached 50%.
automatic transfer of house loads from the aux-iliary transformer to the startup transformers was. defeated, resulting in a plant loss of AC Auxiliary feedwater initiated natural power.
circulation flow through the reactor, and the diesel generators assumed the essential loads until off. lite power was restored.
In the unit startup following the reactor trip December 16, 1977 test (TP 800.14) from 40% power, the turbine-generator was on-line and the reactor was at approximately 11% when.the startup feedwater control valves began to oscillate. These valve
-positi~on swings resulted in overfeeding of steam generator 1-1.
The reactor tripped on low RCS Additional tuning of the ICS was
.g pressure.
performed to minimize these valve oscillations during startups.
Following nine consecu.tive days of steady-state December 30., 1977 power operations at 72% power. #1 main feed pump tripped on " indicated" high exhaust casing water level. An Integrated Control System (ICS) runback was initiated, but the pressurizer power relief valve lifted resulting in a reactor trip on low RCS pressure. The response of the main feed pump speed controls was modified, using the data collected during this trip.
Two SFRCS trips occurred during startup operations.
January 6, 1978 Both were caused by feedwater flow fluctuations which caused feedwatef/ steam outlet pressure differential to exceed the limit. Following the second trip, Auxiliary Feedwater Pump 1-1 was declared inoperable because the speed control cir-cuitry malfunctioned. A circuit modification was completed and tested to correct this problem.
l Davis-Besse - 1.
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i January 21, 1978 To check out the main feedpump speed control changes made as a result of the 12/30/77 trip, a #1 feed pump trip test from 70% power was conducted.
For approximately one minute the runback went smoothly.
Then the running pump tripped on high exhaust casing level. The reactor and turbine were tripped manually, and the plant was controlled with auxiliary feedwater during the cooldown to 532 F.
January 31, 1978 An SFRCS trip at 67% power result 6d in a high pressure RPS trip of the reactor. The SFRCS trip was caused by.a spurious half-trip in'-conjunction with an intentional half-trip of the system while performing the monthly surveillance test. The monthly surveillance test has been modified to reduce the likelihood of a recurrence of this problem..
February 24, 1978 A failed RCS flow transmitter had placed RPS Channel 3 into a tripped condition. An erroneous RCS high temperature signal to Channel 2 of the RPS tripped the unit off-line.
Both problems were investigated and corrected prior to resuming power operations.
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!! arch 1,1978 The reactor wa's at 49% power. The level control valve to deaerator 1-2 failed closed. The main feed pump ran our to-feedwater which initiated an SFRCS trip on feeowater/ steam pressure differential.
___ _- The loss of feedwater and closing of the main steam isolation valves increased RCS pressure which tripped the reactor on RPS high pr6ssure.
April 2,1978 A turbine trip test was performed at 75% power to evaluate piping modifications made on~ the extraction -
steam lines to the deaerator. Ths feedwater flow exceeded the feedwater demand during the runback, resulting in overfeeding the steam generators. This coupled with lifting of;the pressurizer power relief valve caused a reactor trip on low RCS pressure.
April 5,1978 While operating at 100% FP for the first time, B&W requested an immediate reduction in power and a change to 3 RC pump operation while a complete analysis of the LBPRA problem was conducted. The unit was O
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1 Davis-Besse - 1 l
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reduced in power to 65% and RCP 1-1 was manually tripped.
Feedwater demands were not properly ratioed and the feedwater valve AP error signal
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in the ICS affected the main feedwater pump speed to such a degree that the feedwater system reached an uncontrollable oscillation, and the RPS tripped the reactor on low RCS pressure. Since that time, FCR 78-200 has been approved and implemented to de-tune the DP error signal during two MFP operations, and adjustments have been made to properly ratio feedwater after an RCP trip.
April 29,1978 While +he shutdown for the screen outage was in progress, the unit experienced a reactor trip from approximately 20% FP.
This was the first shutdown attempted with only three reactor coolant pumps in operation. As #2 steam generator approached
" low-level limit", the operator used manual control of the main feed pump to maintain 45 psid across the
' main feedwater control valves. This resulted in overfeeding the steam' generator, and although operator action was taken to stabilize the situation, a rapid cooldown took place, tripping the RPS on low RCS pressure, and initiating high pressure injection for approximately 5 minutes. The Reactor Coolant System was returned to 2155 psig/530*F and a normal controlled cooldown to Mode 5 was perfonned.
September 10, 1978 Conducted optional. turbine trip test from 75% FP per TP 800.14.
Excessive feedwater flow resulted in reactor trip on low pressure.
Septeit.oer 28, 1978 While at 90% FP, theioop 2 RCS~fl'oWtransmitter FT
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RC1A1 failed low. This low flow signal caused a trip of RPS Channel 1 and initiated an ICS runback at 20%
per minute.
The runback stopped at 700 MWe and resulted in feedwater to the steam generator ratioed as if the erroneously indicated flow condition actually existed. The o)erator took manual control of loop 2 main feedwater control valve, attempting to maintain level #2 steam generator.
This action re-sulted in feedwater flow greater than that required for the existing reactor power level, and decreased RCS pressure to below 1985 psig RPS trip setpoint.
The plant was placed in Hot Standby (Mode 3) and the RCS flow transmitter was repaired.
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Davis-Besse - 1 -
October 3,1978 While operating at 73% FP, the second EHC pump was stated to investigate the recent reduction in EHC header pressure. A hydraulic perturbation was introduced, tripping the turbine on low EHC pressure. The ICS iaitiated a reactor power runback at 20%/ minute. The increased steam generator pressure and the ICS " cross-limits" rapidly increased feedwater flow, overcooling the RCS and causing an RPS reactor trip on low RCS pressure 84 seconds after the turbine trip. The analysis of this trip resulted in a recomniended modification to the ICS cross-limits, reducing the amount of feedwater added following any turbine trip.
October 29, 1978 With reactor power at 4% of full power while lowering RCS temperature for the Natural Circulation Test, TP 800.04, the reactor operator was controlling Main Feed Pump Turbine (MFPT) 1-2 speed in manual.
The MFPT l-2 motor speed changer hung up at the high speed stop resulting in a high differential pressure across the control valve. As the operator kept trying to reduce the differential pressure, the speed changer was freed and ran MFPT l-2 back resulting in a low differential pressure and a Steam and Feedwater Rupture Control System (SFRCS) full trip at 10:43:06 hours. The resulting sequence
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of events resulted in a reactor trip from low reactor coolant pressure (1985 psig) on Channels 1 and 3 of the Reactor Protection System.(RPS).
The cause of this trip is due to both the unusual plant conditions while perfonning the Natural Circulation Test and the MFPT speed changer diffi-culties. The spped changer was repaired...It ir also believed the pressurizer electromatic relief valve or pressurizer spray valve may not have reset at the proper RCS pressure, The incident is still under investigation.
Question - Did DDR respond to or analyze in any way (before 'the TMI-2 accident) the September 24, 1977 Davis-Besse transient?
The subject Davis-Besse transient event was considered by DDR staff before the TMI-2 event to the extent of identifying potential probler. areas and considering whether the event s'hould.be reported as an Abnorral Occurrence.
The potential problem areas identifed for further study were:
- a.. The water hammer and/or excessive pipe vibration during the transient.
b.
Steam Fes'dwater Rupture Control System (SFRCS) in the secondary systems affects the control of the primary system.
The fuel and clad conditions during the event since boiling occurred in the c.
core and was considered a small LOCA.
d.
Safety features between B&W PWRs were not the same as those in the Westinghouse PWRs (i.e., Westinghouse trips reactor upon loss of mainfeedwater and turbine trip while B&Ws do not).
Because of the reviews of this event undertaken by OIE and DSS, it was agreed that completion of the review of these concerns and this event was to be completed by 3
During the period ending December 23,1978, DDR reviewed the accep,tability of a i
proposed procedural control of steam generator water level.
This review was conducted at the request of 01E.
The need for thii proceoure change may have beer.
initiated by the licensee as a result of the September 24, 1977 event, however thE cocumentation and recollection of those participating ir. :he. review does not show connection betweer. the Septenber 24, 1977 event and :he :rocedure change.
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