ML20040A189
| ML20040A189 | |
| Person / Time | |
|---|---|
| Site: | 05000561 |
| Issue date: | 02/28/1977 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201200495 | |
| Download: ML20040A189 (13) | |
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- 8 UNITED STATES l
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NUCLEAR REGULATORY COMMisslON I ' (L- } l_
WASHINGTON, D. C. 20555 FEB 2 5 "U Docket No. STN 50-561 MEMORANDUM FOR:
D. B. Vassallo, Assistant Director for Light Water t
Reactorr,, DPM i
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FROM:
R. L. Tedesco, Assistant Director for Plant Systems, DSS l
SUBJECT:
SAFETY EVALUATION REPORT FOR BABC0CK & WILC0X STANDARD SAFETY ANALYSIS REPORT (B-SAR-205)
Plant Name: Babcock & Wilcox Standard Safety Analysis Report (B-SAR-205)
Docket Number:
STN 50-561 Licensing Stage:
PDA Milestone Number:
24.02 1
Responsible Branch & Project Manager:
LWR-1, T. Cox Requested Completion Date: March 8,1977 Review Status: Complete The Babcock & Wilcox Standard Safety Analysis Report (B-SAR-205) has been evaluated by Plant Systems. Our review is based on the SSAR including that additional information presented by the applicant in Amendments 1 through 10 to the SSAR.
This SER review is limited to compliance by the applicant with the Codes and Standards Rule, Section 50.55a of 10 CFR Part 50, Seismic Classification and System Quality Group Classifications of fluid containing components which are part of the reactor coolant pressure boundary and other fluid system important to safety which are within the scope of the B-SAR-205 l
review. A final SER evaluation of the material within the scope of our review is enclosed.
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'$h. {, n-ka Robert L. Tedesco, Assistant Director for Plant Systems Division of Systems Safety
Enclosure:
As Stated
Contact:
R. Kirkwood Ext. 27763 C201200495 810403 PDR FOIA MADDEN 80-515 PDR
D. B. Vassallo 2
cc:
S. llanauer R. !!cineman R. Boyd l
W. Mcdonald i
J. Stolz T. Cox V. Benaroya D. Fischer P. Matthews A. Ungaro R. Kirkwood D. Bunch K. Murphy l
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Auxiliary Systems Branch Safety Evaluation Report Babcock & Wilcox Standard Safety AnalysisReport(B-SAR-205)
Docket No STN 50-561 3.2 Classification of Structures, Components and Systems 3.2.1 Seismic Classification _
l Criterion 2 of the General Design Criteria requires that nuclear power plant structures, systems, and components important to safety be designed to withstand the effects of. earthquakes without loss of capability to perform their safety function. These plant features are those necessary to assure (1) the integrity of the reactor coolant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe thutdown condition, or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the guideline exposures of 10 CFR Part 100.
We have reviewed the seismic classification of those Babcock & Wilcox fluid systems and components important to safety that are within the scope of B-SAR-205 and will be designed to withstand without loss of function, the effects of safe shutdown earthquake. These seismic Category I fluid systems and components are:
(1)ReactorCoolantSystem, (2) High-Pressure Injection System, (3) Low-Pressure Injection System, (4) Core Flood System (5) Makeup and Purification System, (6) Decay Heat Removal System, (7) Portions of the Chemical Addition and Boron Recovery System, and (8) Portions of the Fuel Hardling System. Excluded from this review are structures and balance of plant fluid systems that interface with B-SAR-205 fluid systems. The quality group and seismic
classification of such structures, systems and components will be reviewed at the construction permit stage on a case by case basis.
B-SAR-205 fluid systems and components important to safety that will be designed to withstand the effects of a safe shutdown earthquake and remain functional have been identified in an acceptable manner and classified as seismic Category 1 items, in conformance with Regulatory Guide 1.29, " Seismic Design Classification" in Table 3.2-4 of B-SAR-205 except for the following items:
In amendment 8, the applicant revised the seismic and quality group classifications of that portion of the Chemical Addition and Boron Recovery System which ytovides concentrated boric acid to the reactor coolant system, from Seismic Category 1 and Quality Group C to non-seismic Category 1 and Quality Group D.
Since the Chemical Addition and Boron Recovery System is designed to provide the concen-trated boric acid solution which is required to achieve cold shutdown of the reactor under normal operating conditions, we find this down-grading in seismic and quality group classifications to be unacceptable.
Therefore, in order to conform with the guidance provided in Regulatory i
Guides 1.26 and 1.29, we will require that portion of the Chemical Addition and Boron Recovery System which includes the concentrated boric acid storage tanks, boric acid pumps, boric acid filters, interconnecting piping and valves, and the boric acid supply lines to the Makeup and Purification System be classified Quality
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Group C and designed to seismic Category 1 requirements.
j In table 3.2-4, the applicant has classified the deborating demineral -
izers and the reactor coolant degasifier package of the Chemical Addition and Boron Recovery System as non-seismic Category 1 and Quality Group D based on radiological accident analyses which are site-dependent. We find this to be an acceptable method for classi-t fying these components for the B-SAR-205 application. However, based on the radiation source term specified as an interface requirement in t
l B-SAR-205, these components will be reviewed on a case by case basis, e
in a utility applicant's preliminary safety analysis report in order l
to confinn that the non-seismic Category 1 and Quality Group D classifications of these components are consistent with the actual site meteorology.
I l
In table 5.5-3, and on Piping and Instrumentation Diagram 5.5-8, i
l the applicant has classified in an acceptable manner those components that are an integral part of the reactor coolant pump and motor, such as, the oil lubrication system and cooling water lines to the interface t
l point with the balance of plant component cooling water system as i
seismic Category 1 and Quality Groups A or C in conformance with Regulatory Guides 1.26 and 1.29.
In order to assure an adequate supply of cooling water to the reactor coolant pump seals and bearings during normal operation, i
.g.
1 anticipated transients, and accident's we will require an applicant which references B-SAR-205, to apply to the portion of the component cooling water system that interfaces with the reactor coolant pumps and motors the following criteria:
(1) A single active failure in the component cooling water system shall not result in fuel damage or result! in a breach of the reactor coolant pressure boundary when a loss of cooling occurs to one or more reactor coolant pumps.
j (2) A pipe crack or other unanticipated occurance shall not result in either a breach of the reactor coolant pressure boundary or result in excessive fuel damage when a loss of cooling occurs to two or more reactor coolant pumps.
In order to meet these criteria we will require a utility applicant which references B-SAR-205 to provide cooling to the reactor coolant pump seals, and the motor and pump bearings, so that the system meets the following conditions:
(1)
The portion of the component cooling water system which supplies j
cooling water to the reactor coolant pumps and motors may be designed to non-seismic Category 1 requirements and classified Quality Group D if it is demonstrated that the reactor coolant pumps will operate without component cooling water for at least 30 minutes without loss of function or the need for operator protective action.
In addition, safety grade instrumentation d
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cooling water to the reactor coolant pumps and motors, and to I
notify the operator in the control room. The entire instrumen-tation system, including audible and visual alarms, should meet the requirements of IEEE Std 279-1971.
f (2)
If it cannot be demonstrated that the reactor coolant pumps and motors will operate at least 30 minutes without loss of function or operator protective action, then the design of the Component Cooling Water System must meet the following requirements.
1.
Safety grade instrumentation consistent with the criteria for the reactor protection systen shall be provided to initiate automatic protection of the plant. For this case, the com-j ponent cooling water supply to the seals and pump and motor l
bearings may be designed to non-seismic Category 1 requirements l
and Quality Group D; or 2.
the component cooling water supply to the pumps and motors shall be capable of withstanding a single active failure or a i
moderate energy line crack as defined in our Branch Technical Position APCSB 3-1 and be designed to seismic Category 1 Quality Group C and ASME Section III, Class 3 requirements.
Should an analysis be provided to demonstrate that loss of component cooling water to-the reactor coolant pumps and motor assembly is
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J acceptable, we will require certain modifications to the plant tech-y nical specifications and a reactor coolant pump test conducted under operating conditions and with component cooling water terminated for a specified period of time to verify the analysis.
Except for those seismic Category 1 B-SAR-205 fluid systems and components identified above all other B-SAR-205 fluid systems and components that may be required for operation of the facility will be J
designed to other than seismic Category 1 requirements.
Included in this classification are those portions of Category 1 systems which will not be required to perform a safety function.
The basis for acceptance in our review has been in conformance of the applicant's designs, design criteria and design bases for fluid systems, and components important to safety with the Commission's regulations as set forth in General Design Criterion 2, Regulatory Guide 1.29,
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staff technical positions and industry codes and standards.
I Except for the items identified above, we conclude that B-SAR-205 fluid systems and components important to safety will be designed to withstand the effects of a safe shutdown earthquake and remain functional and have been properly classified as seismic Category 1 items in conformance with the Coninission's regulations, the applicable regulatory codes and standards are acceptable. Design of these guide, and industra items in accordance with seismic Category 1 requirements provides reasonable assurance that in the event of a safe shutdown earthquake,
the plant will perform in a manner providing adequate safeguards to the health and safety of the public.
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3.2.2 System Quality Group Classification Criterion 1 of the General Design Criteria requires that nuclear power plant systems and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety function to be performed.
l We have reviewed Babcock & Wilcox's classificatioit system for pressure-retaining components such as pressure vessels, heat exchanger, storage tanks, pumps, piping, and valves in fluid systems important to safety and the assignment by the applicant of quality classes to those portions of systems required to perform safety functions.
Babcock & Wilcox has applied a classification system. Quality Classes 1, 2, 3 and 4 which corresponds to the Commission's Quality Groups A, B, C and D in Regulatory Guide 1.26, " Quality Group Classifications and Standards," to those fluids containing components which are part of the reactor coolant pressure boundary and other fluid systems important to safety where reliance is placed on these systems (1) to prevent or mitigate the consequences of accidents and malfunctions originating within the reactor coolant pressure boundary, (2) to permit shutdown of the reactor and maintain it in the safe shutdown condition, and (3) to contain radioactive material. The major components of these i
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B-SAR-205 fluid systems except for those items identified in Section
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3.2.1 have been classified in an acceptable manner in conformance with Regulatory Guide 1.26 in Tables 3.2-1, 3.2-3 and 3.2-4 and on system piping and instrumentation diagrams in B-SAR-205. Fluid systems pressure-retaining components important to safety that are classified 4
Quality Group A, B or C will be constructed to the ASME Boiler and l
Pressure Vessel Code as follows:
NRC B&W Component Code Quality Group Quality ClassSection III, Division 1, 1974 Edition A
1 Class 1 l
B 2
Class 2 I
C 3
Class 3 l
Components that are classified Quality Group D (Quality Class 4) will be constructed to the following codes as appropriate: ASME Boiler and Pressure Vessel Code,Section VIII, Divisions 1 ANSI B31.1-1973, 4
API-620, API-650, AWWA D100 or ANSI B96.1. All components classified by the applicant as Quality Class 1, 2 or 3 are also designed to seismic Category 1 requirements. Those components classified as Quality Class 4 i
are not designed to remain functional during or after the SSE (non-seismic Category 1).
I Quality Group A components will comply with Section 50.55a of 10 CFR Part 50. Quality Groups B and C components will comply with subsection NA-1140 of the ASME code.
The basis for acceptance in our review has been conformance of the applicant's designs, design criteria, and design bases for pressure-E
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retaining components such as pressure vessels, heat exchangers, storage tanks, pumps, piping and valves in fluid systems important to safety with the Commission's regulations as set forth in General Design Criterion 1, the requirements of the Codes specified in Section 50.55a of 10 CFR 50, Regulatory Guide 1.26 staff technical positions and industry codes and standards.
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Except for the items identified in Section 3.2.1, we conclude that B-SAR-205 fluid systems pressure-retaining components important to safety will be designed, fabricated, erected and tested to quality standards in conformance with the Commission's regulations, the appli-cable regulatory guide, and industry codes and standards, are accep-table. Conformance with these requirements provide rea.* mable assurance that the plant will perform in a manner providing adequata safeguards to the health and safety of the public.
5.2.1.3 Compliance with 10 CFR Part 50, Section 50.55a_
4 We have reviewed the information provided in B-SAR-205 and conclude that pressure-retaining components of the reactor coolant pressure boundary as defined by the rules of 10 CFR Part 50, Section 50.55a, have been properly identified in Table 5.2-1 and classified as ASME Section III, Code Class 1 components. Babcock & Wilcox states that reactor coolant pressure boundary components will be constructed in accordance with the requirements of the applicable codes and addenda as specified by the rules of 10 CFR Part 50, Section 50.55a.
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t conformance with these requirements, the code edition and t'he applicable addenda identified in B-SAR-205 as minimum requirements for each ASME Section III, Code Class 1 component will be upgraded as required based on a utility applicant's PSAR docket date or the actual purchase order date for a specific component and specified in a utility applicant's PSAR.
We et.lude that construction of the components of the reactor coolant pressure boundary in conformance with the ASME code and the Commission's
_ regulations provides adequate assurance that component quality will be commensurate with the importance of the safety function of the reactor coolant pressure boundary and is acceptable.
5.2.1.4 Applicable Code Cases Babcock & Wilcox has made a commitment that no ASME code cases considered unacceptable to the Commission will be applied in the construction of pressure-retaining ASME Section III, Class 1, components within the l
reactor coolant pressure boundary (Quality Group Classification A).
j The applicant has also indicated his intent to comply with Regulatory l
Guides 1. 4 " Code Case Acceptability ASME Section III Design and I
Fabrication", and 1.85, " Code Case Acceptability ASME Section III Materials".
In the event the use of new ASME Council approved code cases are planned, authorization will be requested of the Canmission prior to their application in the construction of Section III, Class 1 components.
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1 We conclude that compliance with the Commissions regulations in the ut.e of approved code cases will result in a component quality level commensurate with the importance of the safety function of the reactor coolant pressure boundary and is acceptable.
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FEB 15 E73 i
Mr. James H. Tay1or, Panager fluclear Power Generation Division Pabcock and Wilcox Comp 1ny P. O. Box 1260 Lynchburg, Virginia 24505 P
Dear Mr. Tay1 or:
In Volume 3 of NUREG-0460, the t uclear Regulatory Commission's (NRC's) staff report on Anti.cipated Transients Without Scram ( AWS), it was recommended that prior to the Commission's consideration of a proposed AWS regulation, certain generic safety analyses should be perfomed.
These analyses are to confim that the proposed modi fications for various classes of Light Mater Reactor (LWR) designs accomplish the degree of ATWS prevention and mitigation ^ described by the staff in Volume 3 of NUREG-0460.
The Regulatory Requirenents Review Committee has concurred with the generic analysis approach and the Director of the Office of Nuclear Reactor Regulation tus authorized the staff N
to proceed. If the generic analysis approach is successful, the rule to be proposed for. Commission action will not treat AWS as a design basis accident and will not require a new sa fety analysis of AWS on each licensing case. There might be specific exceptions in the future where an analysis for a particular design would be desirable or necessary because the present generic analyses do not envelop that specific design or some future, unanticipated mode of nomal operation.
Generic questions and guidelines are provided in Enclosure 1 for two kinds of plant modifications recommended in Volume 3 of NUREG;0460.
,e These are the Alternative 3 modifications for plants recei~ving a Con-struction Pemit prior to January 1,1978, and the Alternative 4 ' mod-ifications for plants receiving a Construction Pemit after January 1, 1978. The plants listed in Enclosure 2 which began operation prior to Dresden 2 will be treated according to Alternative 2 of Volume 3 a nd will be exami ned on a ca se-by-ca se ba si s a fter the ATWS rul e i s pronul ated in its final, effective form.
9 We require that by April 15,'1979, the four LWR vendors provide respon-ses to the questions in Enclosure 1 applicable to their designs.
Responses to some of the questions can be delayed until' June 1,1979.-
These are noted by an asterisk or footnote in the enclosure.
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Mr. James H. Tayl or FEB 15 073
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for this generic analysis approach to be successful, it is imperative tha t: a) the responses be complete; b) the ce'sponses cover all LWR designs for eacf1 vendor, except the plants in Enclosure 2; c) consider-ation be given in the selection of analysis parameters to envelope the nominal conditions for these designs and their anticipated modes of operation as specified in Enclosure 1 so as to minimize the need for A1WS reanalysis in the future; and d) applicants and licensees provide the n.ecessary support to the four LWR vendors to complete these generic analyses in the required time frame.
The time available to complete the generic analyses is short.
Therefore, it is important that the questions be fully understood and that the answers be as complete as possible so that our~ review does not bog down with an iteration of questions and answers.
To this end we have scheduled a meetin'g in' Bethesda, ftryland, Room P-118, for all day
, Parch 1,1979, to explain and discuss the questions with representatives fran the four LWR vendors.
It may be necessary to further subdivide the question list at that time to assure timely submission of the generic analyses necessary for the staff to complete its drafting of J
I the proposed ATMS rule in Pay. The meeting will be open to interested I
manbers of the public.
Representatives of interested and potentially affected utilities are also invited to attend by copy of this letter.
Si ncerely, I
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Roger J. Pa ttson, Di rec' or Division of Systems Qfety Office of fiuclear Reactor Regulation j
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Enclosures.
1.
Generic Questions 2.
List of Plant for Alternative 2
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April 5. 1979 s,
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MEMORANDUM FOR:
B. H. Grier, Director, Region I J. P. O'Reilly, Director, Region II J. G. Keppler,. Director, Region III K. V. Seyfrit, Director, Region IV
.R. H. Engelken, Director, Region V FROM:
Norman C. Moseley, Director, Division of Reactor Operations Inspection, OIE
SUBJECT:
IE BULLETIN-79-05A, NUCLEAR INCIDENT AT THREE MILE ISLAND The subject IE Bulletin should'be dispatched for action by April 5, 1979, to all B&W power reactor facilities with an operating license.
The' Bulletin must be faxed to B&W facilities by April 5.
Subject bulletin and enclosures should also be dispatched for informa-tion to all other power reactor facilities with an operating license and to all power reactor facilities with a construction permit.
The text of the Bulletin, Enclosures thereto and draft letters to the licensee are enclosed for this purpose.
The continuing Preliminary
. Notifications of the incident should continue to be forwarded as they are received in accordance with the transmittal memorandum for IE Bulletin 79-05.
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. Moseley, Director
/",,Jorma Divisi n of Reactor Operations Inspection Office of Inspection and Enforcement
Enclosures:
1.
Draft Transmittal Letter to all B&W Operating Licensees 2.
Draft Transmittal Letter to all other Operating Licensees and Construction Permit Holders.
3.
IE Bulletin No.79-05A (w/en. closures - 2)
CONTACT:
D. C. Kirkpatrick, IE I
49-28180 C
n 49 R / W., s a
I (Draft letter to B&W power reactor facilities with an operating license.)
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IE Bulletin No.79-05A Addressee:
Enclosed is IE Bulletin No.79-05A, which requires action by you with regard to your power reactor facility (ies) with an operating license.
Should you have any questions regarding this Bulletin or the actions required by you, please contact this office.
Sincerely, Signature (Regional Director)
Enclosure:
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with Enclosures l
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(Draft letter to all power reactor facilities with an operating license or a construction permit.)
IE Bulletin No.79-05A Addressee:
The enclosed Bulletin 79-05A is forwarded to you for information.
No written response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely.
Signature (Regional Director)
Enclosure:
IE Bulletin No.79-05A with Enclosures
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UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION NiD ENFORCEMENT WASHINGTON, DC 20555 APRIL 5, 1979 IE Bulletin 79-05A NUCLEAR INCIDENT AT THREE MILE ISLAND - SUPPLEMENT Description of Circumstances:
Preliminary information received by the NRC since issuance of IE Sulletin 79-05 on April 1,1979 has identified six potential human, design and mechanical failures which resulted in the core damage and radiation releases at the Three Mile Island Unit 2 nuclear plant. The information and actions in this supplement clarify and extend the original Bulletin and transmit a preliminary chronology of the TMI accident through the first 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> (Enclosure 1).
1.
At che time of the initiating event, loss of feedwater, both of the auxiliary feedwater trains were valved out of service.
2.
The pressurizer electromatic relief valve, which opened during the initial pressure surge, failed to close when the pressure decreased below the actuation level.
3.
Following rapid depressurization of the pressurizer, the pressurizer level indication may have lead to erroneous inferences of high level in the reactor coolant system.
The pressurizer level indication apparently led the operators to prematurely te'rminate high pressure injection flow, even though substantial voids existed in the reactor coolant system.
4.
Because the containment does not isolate on high pressure injection (HPI) initiation, the highly radioactive water from the relief valve discharge was pumped out of the containment by the automatic initiation of a transfer pump. This water entered the radioactive waste treatment system in the auxiliary building where some of it overflowed to the floor.
Outgassing from this water and discharge through the auxiliary building ventilation system and filters was the principal source of the offsite release of radioactive noble gases.
5.
Subsequently, the high pressure injection system was intermittently operated attempting to control primary coolant inventory losses throUgh the electromatic relief valve, apparently based on pressurizer level indication. Due to the presence of steam and/or noncondensible voids elsewhere in the reactor coolant system, this led to a further reduction in primary coolant inventory.
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4 (This item clarifies and expands upon item 4. of IE Bulletin 79-05.)
Review the actions directed by the operating procedures and training instructions to ensure that:
Operators do not override automatic actions of engineered a.
sa fety. features.
pg Operating procedures currently [, or are revised tofspecifd ~
b.
that if the high pressure injection (HPI) system has been automatically actuated because of low pressure condition, it must remain in operation until either:
(1)
Both low pressure injection (LPI) pumps are in operation and flowing at a rate in excess of 1000 gpm each and the
-g situation has been stable for 20 minutes, or f
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(2)
The HPI system has been in operation for 20 minutes, g,W '
and all hot and cold leg temperatures are at least 4f k 50 degrees below the saturation temperature for the
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F existing RCS pressure.
If 50 degree subcooling cannot be maintained after HPI cutoff, the HPI shall be reactivated.
h c.
Operating procedures currently or are rev that in the event of HPI initiation, with reactor coolant pumps (RCP) operating, at least one RCP per loop shall remain operating.
d.
Operators are provided additional information and instructions to not rely upon pressurizer level indication alone, but to also examine pressurizer pressure and other plant parameter indications in evaluating plant conditions, e.g., water inventory in the reactor primary system.
5.
(This item revises item 5. of IE Bulletin 79-05.)
Verify that emergency feedwater valves are in the open position in accordance with item 8 below.
Also, review all safety-related valve positions and positioning requirements to assure that valves are positioned (open or closed) in a manner to ensure the proper-operation of engineered safety features.
Also review related procedures, such as those for maintenance and testing, to ensure that such valves are returned to their correct positions following' necessary manipulations,
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- 10. Review and modify as necessary your maintenance and test procedures to ensure that they require:
Verification, by inspection, of the operability of redundant a.
safety-related systems prior to the removal of any safety-related system from service, b.
Verification of the operability of all safety-related systems when they are returned to service following maintenance or testing.
A means of notifying involved reactor operating personnel c.
whenever a safety-related system is removed from and returned to service.
11.
All operating and maintenance personnel should be made aware of the extreme seriousness and consequences of the simultaneous blocking of both auxiliary feedwater trains at the Three Mile Island Unit 2 plant and other actions taken during the early phases of the accident.
12.
Review your prompt reporting procedures for NRC notification to assure very early notification of serious events.
For Babcock and Wilcox pressurized water reactor facilities with an operating license, respond to Items 1, 2, 3, 4.a and 5 by April 11, 1979.
Since these items are substantially the same as those specified in IE Bulletin 79-05, the required date for respense has not been changed.
Respond to Items 4.b through 4.d. and 6 through 12 by April 16, 1979.
Reports should be submitted to the Director of the appropriate NRC Regional Office and a ccpy should be forwarded to the NRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, DC 20555.
For all other reactors with an operating license or construction permit, this Bulletin is for information purposes and no written response is required.
Approved by GAO, B 180225 (R0072); clearance expires 7-31-80.
Approval was given under a blanket clearance specifically for identified generic problems.
Enclosures:
1.
Preliminary Chronology of TMI-2 3/38/79 Accident Until Core Cooling Restored.
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2.
t.ist of IE Bulletins issued in last 12 months.
IE 80lletin 79-05A April 5, 1979 PRELIMINARY CHRONOLOGY OF TMI-2 3/28/79 ACCIDENT UNTIL CORE COOLING RESTORED TIME (Approximate)
EVENT about 4 AM Loss of Condensate Pump (t a 0)
Loss of Feedwater Turbine Trip t = 3-6 sec.
Electromatic relief valve opens (' 255 psi) 2 to relieve pressure in RCS t = 9-12 sec.
Reactor trip on high RCS pressure (2355 psi)
= 12-15 sec.
RCS pressure decays to 2205 psi (relief valve should have closed)
- = 15 sec.
RCS hot leg tamperature peaks at 611 degrees F, 2147 psi (450 psi over saturation) t = 30 sec.
All three auxiliary feedwater pumps running at pressure (Pumps 2A and 2B started at turbine trip).
No f?)w was injected since discharge valves were closed.
t = 1 min.
Pressurizer level indication begins to rise rapidly t = 1 min.
Steam Generators A and B secondary level very low - drying out over next couple of minutes.
t = 2 min.
ECCS initiation (HPI) at 1600 psi t = 4 - 11 min.
Pressurizer level off scale - high - one HPI pump manually tripped at about 4 min.
30 sec.
Second pump tripped at about 10 min. 30 sec.
l t = 6 min.
RCS flashes as pressure bottoms out at 1350 psig (Hot leg temperature of l
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584 degrees F)
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t = 7 min., 30 sec.
Reactor building sump pump came on.
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'. TIME EVENT t = 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Operator opens electromatic relief valve to depressurize RCS to attempt initiation of RHR at 400 psi t = 8 - 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> RCS pressure decreases to about 500 psi Core Flood Tanks partially discharge t = 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 28 psig containment pressure spike, containment sprays initiated and stopped after 500 gal. of NaOH injected (about 2 minutes of operatien) t = 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Electromatic relief valve closed to repressurize RCS, collapse voids, and start RC pump t = 13.5 - 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> RCS pressure increased frcm 650 psi. to 2300 psi t = 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> RC pump in loop A started, hot leg temperature decreases to 560 degrees F, and cold leg temperature increases to 400 degrees F.
indicating flow through steam generator Thereafter S/G "A" steaming to condensor Cordensor vacuum re-established RCS cooled to about 280 degrees F.,
1000 psi Now (4/4)
High radiation in contair.:nent All core thermocouples~ less than 460 degrees F.
Using pressurizer vent valve with small makeup flow Slow cooldown RB pressure negative s
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UNITED STATES NUCLEAR REGULATORY COMMISSION f
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ItAY 171979 Mr. James H. Taylor Manager, Licensing Babcock & Wilcox Company l
P. O. Box 1260 Lynchburg, Virginia 24505
Dear Mr. Taylor:
SUBJECT:
ACRS RECOMMENDATIONS RELATING TO THI-2 ACCIDENT As you know, the Advisory Comittee on Reactor Safeguards (ACRS) has written three separate letters to the Comission relating to the THI-2 accident. These letters were issued on April 7, April 18, and April 20, 1979. Each of these letters contains specific recomendations which must be addressed.
The ACRS has discussed these recommendations with the staff and industry at recent ACRS subcomittee and full comittee meetings. We anticipate that this dialogue will continue in the forth-(
coming ACRS subcomittee meeting on TMI-2 (May 31-June 1)..
For this reason, we are requesting that all light water reactor vendors provide the staff with a concise discussion and position on each of the ACRS recommeno tions relating to TMI-2.
For your convenience, we have provided in the enclosure a list which contains each ACRS recomendation of which we are aware. We request that you provide your response to each of these ACRS recomendations to be received by the staff no later than Me 2 3, 197 9.
If you require any clarification of the matters discussed herein, please contact C. J. Heltemes, Jr. Pr. Heltemes' phone number is (301) 492-7745.
Sincerely, b
r h
ss, l'eputy Di r/
D. F.
ector Division of Project Management Office of Nuclear Reactor Regulation
Enclosure:
As stated op) yc
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o ENCLOSURE ACRS RECOMMENDATIONS _
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Letter, M. Carbon to Chairman Hendrie, dated April 7,1979 A.
Recomendation 1 - Perfom 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:
r (a) unambiguous RV level indication
[
(b) remotely controlled vent for RCS high points Recomendation 3 - Item 4b of Bulletin 79-05A considered unduly pre-scriptive in view of uncertainties in predicting course of anomalous small break transients / accidents.
Letter, R. Fraley to Comissioners, dated April id,1979 l
B.
Recommendation 1 - Natural Circulation-related Items
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Detailed analyses of natural circulation mode, P
a.
supported as required by experiment, by licensees and NSSS vendors.
Develop procedures for initiating natural circulation.
b.
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Provide operator means for assurance that natural 3
circulation has been established, e.g., by install-c.
ation of instructions to indicate flow at low I
velocities.
Expeditiously survey operating PWR's to determine d.
whether suitable arrangements of PZR heaters and
.I reliable on-site power distribution can be provided to assure this vital aspect of natural circulation capability.
[
Operator should be adequately informed concerning RCS F
e.
conditions which affect natural circulation capa-bility, e.g.,
(1) indication that RCS is approaching saturation L
/
condition by suitable display to operator of T &T and PIR pressure in conjunction with g
sfeamkables (2) use of flow exit temperature indicator by fuel assembly themocouples, where available.
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Enclosure - page 2 ACRS Recomendations
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Recomendation 2 - Thermocouples used to measure fuel assembly exit temperatures to determine core performance should be used, where currently available, to guide operator
-f concerning core status (full range capability).
Recomendation 3 - Operating reactors should be given priority regarding definition and implementation of. instrumentation to t
diagnose and follow the course of a serious accident, I
including i
I (a) improved sampling procedures under accident
[
conditions (b) improved techniques to provide guidance to offsite authorities.
b Recomendation 4 - Reiterates previous recomendations that high priority ~
{
be given to "research to improve reactor safety" 6
(a) research on behavior of LWR's during anomalous transients (b) NRC to develop capability to simulate wide range l_
of postulated transients and accident conditions.
Recomendation 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 Recomendation 1 - Initiate immediately a survey of operating procedures for achieving natural circulation, including:
(a) event involving loss of offsite power f
(b) consideration of role of PZR heaters.
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