ML20030A898
| ML20030A898 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Peach Bottom, Nine Mile Point, Oyster Creek, Hope Creek, Pilgrim, Susquehanna, Limerick, Vermont Yankee, FitzPatrick, Shoreham |
| Issue date: | 07/24/1981 |
| From: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Counsil W, Curtis N, Daltroff S, Finfrock I, Kemper J, Lempges T, Martin T, Morisi A, Pasternak R, Pollock M, Rhode G, Rich Smith BOSTON EDISON CO., JERSEY CENTRAL POWER & LIGHT CO., LONG ISLAND LIGHTING CO., NIAGARA MOHAWK POWER CORP., NORTHEAST NUCLEAR ENERGY CO., PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC, PENNSYLVANIA POWER & LIGHT CO., POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK, Public Service Enterprise Group, VERMONT YANKEE NUCLEAR POWER CORP. |
| References | |
| NUDOCS 8107290166 | |
| Download: ML20030A898 (3) | |
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Gentlemen:
The enclosed IE Circular No. 81-11, " Inadequate Decay Heat Removal During Reactor Shutdown", is forwarded to you for appropriate action /information.
No written response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely, CC Boyce H. Grier Director
Enclosure:
IE Circular No. 81-11 with 1 attachment CONTACT:
D. L. Caphton (215-337-5287)
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LIST OF HOLDERS OF BWR OPERATING LICENSES OR CONSTRUCTION PERMITS RECEIVING IE CIRCULAR NO. 81-11 Boston Edison Company M/C Nuclear Docket No. 50-293 ATTN: Mr. A. V. Morisi Nuclear Operations Support Manager 800 Boylston Street Boston, Massachusetts 02199 Jersey Central Power and Light Conipany Docket No. 50-219 ATTN: Mr. Ivan R. Finfrock, Jr.
Vice President Oyster Creek Nuclear Generating Station P. O. Box 388 Forked River, New Jersey 08731 N.agara Mohawk Power Corporation Docket No. 50-220 ATTN: Mr. T. E. Lempges Vice President Nuclear Generation 300 Erie Boulevard West Syracuse, New York 13202 Northeast Nuclear Energy Company Docket Nos. 50-336 ATTN: Mr. W. G. Counsil 50-245 Senior Vice President - Nuclear 50-423 Engineering and Operations P. O. Box 270 Hartford, Connecticut 06101 Philadelphia Electric Company Docket Nos. 50-277 ATTN: Mr. S. L. Daltroff 50-278 Vice President l
Electric Production 2301 Market Street Philadelphia, Pennsylvania 19101 Power Authority of the State of New York Docket No. 50-333 I
James A. FitzPatrick Nuclear Pcwer Plant l
ATTN: Mr. R. J. Pasternak Resident Manager P. O. Box 41 l
Lycoming, New York 13093 l
Vermont Yankee Nuclear Power Corporation Docket No. 50-271 l
ATTN: Mr. Robert L. Smith Licensing Engineer 1671 Worcester Road Framingham, Massachusetts 01701 l
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2 Long Island Lighting Company Docket No. 50-322 ATTN:
Mr. M. S. Pollock V:ce Presid r.t - Nuclear l
175 East Old Country Road Hicksville, New York 11801 Niagara Mohawk Power Corporation Docket No. 50-410 ATTN: Mr. Gerald K. Rhode Vice Presiden'.
System Project Management c/o Miss Catherine R. Seibert 300 Erie Boulevard, West i
Syracuse, NY 13202 Pennsylvania Power & Light Company Docket Nos. 50-387 ATTN: Mr. Norman W. Curtis 50-388 Vice President Engineering and Construction - Nuclear 2 North Ninth Street Allentown, Pennsylvania 18101 Philadelphia Electric Company Docket Nos. 50-352 ATTN:
Mr. John S. Kemper 50-353 Vice President Engineering and Research 2301 Market Street Philadelphia, Pennsylvania 19101 Public Service Electric & Gas Company Docket Nos. 50-354 ATTN:
Mr. T. J. Martin 50-355 Vice President Engineering and Construction 80 Park Plaza - 17C Newark, New Jersey 07101 l
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NUCLEAR REGULATORY COMMISSION J,
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July 24, 1981 IE Circula-No. 81-11:
INADEQUATE DECAY HEAT REMOVAL DURING REACTOR SHUTDOWN
Background:
Following several losses or' decay heat removal capability at operating pressurized water reactors (PWRs), IE Bulletin 80-12 " Decay Heat Removal Systen. Operability" (issued May 1,1980) requested PWR licensees to take certain actions intended to reduce the probability of loss of decay heat removal. All operating PWRs were requested to amend the Technical Specifi-I cations for their facilities with respect to reactor decay heat removal capability by letter from D. Eisenhut, Division of Licensing, on June 11, 1980.
IE Bulletin 80-12 was issued to boiling water reactor (SWR) licensees for information with the expectation that the information would be evaluated for applicability and subsequent actic, taken as determined necessary.
- However, events involving inadequate decay heat removal at operating BWRs now indicate the need for EWR licensees to provide additional controls related to decay heat removal.
Descriotio6 of Circumstances:
1.
Brunswick - Temocracy Loss of Shutdown Cooling
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On December 8, 1980, unplanned heatup of the reactor coolant occurred at Brunswick Unit 2 when the unit was in cold shutdown (F212'F) with all rods inserted.
The heatup occurred while the service water cooling for the "A" loop of the residual heat removal (RHR) system was isolated longer than expected for repair of a service water leak.
Shutdown cooling was not lined up to loop "B" (1) because it was expected that loop "A" would be returned to service before 212 F was reach.
and (2) because of the length of i
time required to line up the "B" loop for operation.
During the repair, the recirculation pumps were off, an RHR pump was running, and the control rod drive pump was supplying water to the reactor pressure vessel (RPV) while the reactor water cleanup (CU) system was rejecting water for level I
l control.
The reactor coolant ter, rature monitored at the CU inlet (from recirculation loop) indicated F212 F during the repair.
The reactor l
2 head vents were reported to be opened during this period, with no evidence of steaming.
However, average coolant temperature at the time of completion of repair approached 212 F with an observed maximum of 217cc.
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Shutdown cooling was initiated and primary coolant temperature decreased to a normal temperature within approximately 30 minutes.
Primary containment could not be quickly established due to cables going through the personal access hatch and the torus hatch being removed.
A similar event occurred at Brunswick Unit 2 on the following day. With the primary containment and reactor head vents reported open, the conventional and nuclear service water systems were secured to repair a conventional service water pump discharge check valve.
The primary coolan l
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IEC el-11 July 24, 1981 Page 2 of 4 temperature initially was less than 120 F.
Approximately two hours after the sarvice water systems were secured, the RHR pumps in the A loop were secured to reduce coolant haat input from the pumps.
Repairs took longer than anticipated, and when the conventional and nuclear service water systems were returned to service, the primary coolant temperatwa at the vessel bottom head drain was 147'F. Approximately fifteen minutes later shutdown cooling was initiated using the B loop of the RHR.
There were indications of heatup of the coolant to approximately 212 F; however, there was no evidence of steaming through the open reactor heat vents.
Primary coolant temperature decreased to a normal temperature within approximately three hours.
2.
Dresden Unit 3 - Unolanned Repressurization On December 20, 1980, the Dresden Nuclear Power Station Unit 3 was in the cold shutdown condition.
Numerous maintenance and modification outages were in progress which resulted in the shutdown and/or isolation of all systems which communicate with the reactor vessel, and which normally provide cooling and recirculation of the primary coolant.
Subsequently, one of three loops of the shutdown cooling system (SDC) was put in service to maincain reactor water temperature at approximately 150 F.
The reactor water level was maintained at the normal operating level (instead of flooding up) to limit vessel safe end thermal stresses.
Because the design of the SDC does not allow for throttling of the cooling water flow to the SDC heat exchangers, it is standard practice to throttle SDC flow to the recirculation loop ta maintain vessal temperature when 1
in cold shutdown. As the decay heat 'oad decreased the unit operators reduced SDC flow until insufficient v isel flow existed to provide mixing of the primary coolant, and accurate t mperature measurements by the recirculation pump and SDC pump suctior temperature instruments.
Because i
the operators monitored only the recirc lation pump and SDC temperatures, a slow heatup and repressuri:ation of tre reactor vessel to 175 psig occurred over a six hour period of time.
Upon discovering the repressurization, 50': flow was increased, and a l
second SDC loop was placed in service to expedite the return to cold l
shutdown.
The indicated recirculation suction temperature rose to approximately 225 F, indicating that the entire vessel contents did not heat up to the saturation temperature at 175 psig (377 F).
During the repressurization event the containment personnel access doors l
were open, resulting in violation of the Technical Specification limiting condition for operation for primary containment integrity.
Had the l
Technical Specificatian been revised to conform to current BWR standard Technical Specifications the LC0's for the High pressure coolant injection system and isolation condenser systems would also have been exceeded.
l Post event evaluations of the circumstances leading up to the repressur-ization, and the chronology of the event itself, estsblish that the hm M%
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IEC 81-11 July 24, 1931 Page 3 of 4 licensee. did not evaluate the potential for adverse effects on plant safety resulting from procedure changes removing the vessei floodup requirement, and the effect of removing from service those systems which normally cool and recirculate the reactor coolant.
The potential for inaccurate response of normally used instrumentation was apparently not considered by the licensee, and redundant instrumentation which could have provided warning that the event was in progress was not utilized by operations personnel.
The licensees of the above facilities have committed to make administrative and procedural changes to provide personnel additional guidance when operating in the shutdown cooling mode. Additional information regarding these events and corrective actions is contained in LERs 2-80-107, 2-80-112 (Brunswick 2), and LER 80-047/01T-0 (Dresden 3).
There have been recent events at otne.r BWRs involving the loss of systems providing normal decay heat removal, and appropriate action has been taken by operating personnel to put alternate cooling in service.
These events indicate the need for timely operator response and the need to have backup systerrs available.
Recommended Action for Licensees of BWRs with an Ooerating License:
1.
Review your existing procedures and administrative controls that relate to decay neat removal during reactor shutdown. Analyze these procedures for adequacy of monitoring and responding to events involving lost or degraded decay heat removal.
Special emphasis should be placed on conditions involving icw core recirculation or cooling flow, or when maintenance or refueling activities degrade the decay heat removal capability.
2.
Administrative controls :hould provide the following:
a.
Assure that redundant or diverse decay heat removal methods are available during all modes of plant operation.
(Note: When in a l
refueling mode with water in the refueling cavity and the head removed, an acceptable means could include one decay heat removal I
train and a readily accessible source of water to replenish any l
loss of inventory).
(Note:
Only one power source needs to be operable in order to consider the decay heat removal system operable while in modes 4 and 5).
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b.
For those cases where single failures or other actions result in only one decay heat removal train being available, provide an additional alternate means of decay heat removal or provide an expeditious means for the restoration of the lost train or method.
j c.
Implement administrative controls during periods of low flow or nc l
flow to ensure that the maximum coolant temperature remains below the saturation temoerature.
Consideration should be given to maintaining water level in the reactor vessel sufficiently high to l
enable natural circulation at all times.
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IEC 81-11 July 24, 1981 Page 4 of 4 d.
Require monitoring of the reactor coolant temperature and pressure 3
at a specified frequency.
3.
Any changes needed in the existing procedures or administrative controls as a result of Items 1 and 2 above should be implenented within 120 days of the date of this circular.
No written response to this circular is required.
If you need additional information regarding this subject, please contact the appropriate Regional Office.
Attachment:
Recently issued IE Circul2rs l
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IE Circular 81-11 July 24, 1981 RECENTLY ISSUED IE CIRCULARS Circular Date of No.
Subject Issue Issued to 81-12 Inadequate Periodic Test 7/22/81 All power reactor Procedure of PWR Protection facilities with an System OL or CP 81-10 Steam Voiding in the 7/2/81 All holders of Reactor Coolant System PWR reactor During Decay Heat OL or CP Removal Cooldown 81-09 Containment Effluent 7/10/81 All power reactor Water That Bypasser facilities and and Radioactivity Monitor OL 81-08 Foundation Materials 5/29/81 All holders of a power reactor OL or CP 81-07 Control of Radioactively 5/14/81 All holders of Contaminated Material a power reactor 4
l OL or CP 81-06 Potential Deficiency 4/14/81 All holders of a Affecting Certain Foxboro power reactor 10 to 50 Mil 11 ampere OL or CP Transmitters 81-05 Self Aligning Rod 3/3/81 All holders of a l
End Bushings :or Pipe power reactor Supports OL or CP 81-04 The Role of Shift Technical 4/30/81~
All holders of Advisors and Importance of a power reactor Reporting Operational Events OL I
l 81-03 Inoperable Seismic 3/2/81 All holders of a j
Monitoring Instruments power reactor OL or CP l
1 81-02 Performance of NRC-Licensed 2/9/81 All holders of a Individuals While on Duty power, test, or research reactor OL or CP l
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