ML20009E428
| ML20009E428 | |
| Person / Time | |
|---|---|
| Site: | Monticello, Dresden, Perry, Fermi, Duane Arnold, Clinton, Quad Cities, La Crosse, Big Rock Point, LaSalle, Zimmer, Bailly  |
| Issue date: | 07/24/1981 |
| From: | Carroll D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Wright G ILLINOIS, STATE OF |
| References | |
| NUDOCS 8107280196 | |
| Download: ML20009E428 (2) | |
Text
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J Illinois Department of JUL 2 7 S31* b ATT
. Ga N. Wright
" EsE" I2 Deputy Director 1035 Outer Park Drive 9
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Springfield, IL 62704 Gentlemen:
The enclosed IE Circular No. 81-11 titled " Inadequate Decay Heat Removal During Reactor Shutdown" was sent to the licensees listed below on July 24, 1981:
ACTION Commonwealth Edison Company Dresden 1, 2, 3 (50-10, 50-237, 50-249)
Quad-Cities 1, 2 (50-254, 50-265)
Consumers Power Company Big Rock Point (50-155)
Dairyland Power Cooperative LACBWR (50-409)
Iowa Electric Light & Power Company Duane Arnold (50-331)
Northern States Power Company Monticello (50-263)
INFORMATION Cincinnati Gas and Electric Company Zimmer (50-358)
C1 veland Electric Illuminating Company Perry 1, 2 (50-440, 50-441)
Commonwealth Edison Company LaSalle 1, 2 (50-373, 50-374) p
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4 Illinois Department of Nuclear Safety Detroit Edison Company Fermi 2 (50-341)
Illinois Power Company Clinton 1, 2 (50-461, 50-462)
Northern Indiana Public Service Company Bailly (50-367)
Sincerely, t
/. AM&Lb othy E. Carroll, Chief Word Processing and Document Control Section
Enclosure:
IE Circular No. 81-11 cc w/ encl:
Mr. D. W. Kane, Sargent & Lundy J. G. Keppler, RIII Acting Division Directors, RIII P. R. Wohld, RIII l
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SSINS No.:
6830 Accession No.:
8011040256
-IEC'81-11 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 July 24, 1981 IE Circular No. 81-11:
INADEQUATE DECAY HEAT REMOVAL DURING REACTOR SHUTDOWN Eac< ground:
Fol tow)..g several losses of decay heat removal capability at operating pressurized water reactors (PWRs), IE Bulletin 80-12 " Decay Heat Removal System Operability" (issued May 1, 1980) reque;ted 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-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 (BWR) licensees for information with the expectation that the information would be evaluated for applicability and subsequent iction taken as determined necessary.
- However, events involving inadequate cecay heat removal at operating BWRs now indicate the need for BWR licensees to provide additional controls related to decay heat removal.
~
Description of Circumstances:
1.
Brunswick - Temporary Loss of Shutd_own Cooling On December 8, 1980, unplanned heatup of the reactor coolant occurred at Brunswick Unit 2 when the unit was in cold shutdown (<212 F) with all rods inserted. The heatup occurred while the service water cooling for tLe "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 reached and (2) because of the length of 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 control.
The reactor coolant temperature monitored at the CU inlet (from a recirculation loop) indicated <212 F during the repair.
The reactor head vents were reported to be opened during this period. v'th no evidence of steaming.
However, average coolant temperature at the ume of completion of repair approached 212 F with an observed maximum of 217 F.
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 serv:ce water pump discharge check valve.
The primary coolant
IEC 81-11 July 24, 1981 Page 2 of 4 temperature initially was less'than 120 F.
Approximately two hours after the service water systems were secured, the RHR pumps in the A loop were secured to reduce coolant heat input from the pumps.
Repairs took longer than inticipated, and when the conventional and nuclear service water systems werr returned to service, the primary coolant temperature at the vessel sottom 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; hn.ever, there was no evidence of steaming through the open reactor heat vents.
Primary coolant temperature decreased to a normal temperature witnin approximately three hours.
2.
Dresden Unit 3 - Unplanned 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 maintain 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 to maintain vessel temperature when in cold shutdown.
As the decay heat load decreased the unit operators reduced SDC flow until insufficient vessel flow existed to provide mixing of the primary coolant, and accurate temperature measurements by the recirculation pump and SDC pump suction temperature instruments.
Because the operators monitored only the recirculation pump and SDC temperatures, a slow heatup and repressurization of the reactor vessel to 175 psig occurred over a six hour period of time.
Upon discovering the repressurization, SDC flow was increased, and a second SDC loop was placed in se'/ ice to expedite the return to cold 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).
Diring the repressurization event the containment personnel access doors were open, resulting in violation of the Technical Specification limiting condition for operation for primary containment integrity.
Had the Technical Specification 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.
Post event evaluations of the circumstances leading up to the repressur-ization, and the chronology of the event itself, establish that the
IEC 81-11 July 24, 1981 Page 3 of 4 licensee did not evaluate the potential for adverse effects on plant safety resulting from procedure changes removing the vessel 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 ins +eumen;ation was apparently not considered by the licensee, and redundant.nst rumentation which could have provided warning that the event was in 'rogress was not utilized by operations personnel.
The licensees of the above facilities have commi'+ed to make administrative and procedural changes to provide personnel additional guidance when oprrating in the shutdown cooling mode. Additional information regarding these w ents and corrective actions is contained in LERs 2-80-107, 2-80-112 (Brunswick 2), and LER 80-047/017 0 (Dresden 3).
There have been recent events at other BWRs involving the loss of systems providing normal decay heat removal, and appropriate action has been taken by operating peisonnel to put alternete cooling in service.
These events indicate the need for timely operator response and the need to have backup systems available.
Recommended Action for Licensees of BWRs with an Operating License:
1.
Review your existing procedures and administrative controls that relate to decay heat removal during reactor shutdown.
Analyze these procedures for adequacy of monitoring and responding to events involving lost or degraded decay heat remova'.
Special emphasis should be placed on conditions involving low core recirculation or cooling flow, or whe3 maintenance or refueling activities degrade the decay heat removal capability.
2.
Administrative controls should 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 refueling mode with water in the refueling cavity and the head removed, an acceptable means could include one decay heai, removal train and a readily accessible source of water to replenish any 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).
b.
For those cases where single failures or other actions result in only one deca) 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.
Implement administrative controls during periods of low flow or no flow to ensure that the maximum coolant temperature remains below the saturation temperature.
Consideration should be given to maintaining water level in the reactor vessel sufficiently high to l
enable natural circulation at all times.
IEC 81-11 July 24, 1981 Page 4 of 4 d.
Require monitoring of the reactor coolant temperature and pressure at a specified frequ2ncy.
3.
Any changes needed in the existing procedures or administrative controls as a result of Items 1 and 2 above should be implemented 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 Circulars l
,,-.a v
gre a y Attachment IEC 81-11 July 24, 1981 i
RECENTLY ISSUED j
IE CIRCULARS l
Circular Date of No.
Subject Issue Issued to i
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 Reactor 7/2/81 All power reactor Coolant System During Decay facilities with an Heat Removal Cooldown OL or CP 81-09 Containment Effluent Water 7/10/81 All power reactor That Bypasses Radioactivity facilities with an Monitor OL.
81-08 Foundation Materials 5/29/81 All power reactor facilities with an OL or CP j
81-07 Control of Radioactively 5/14/81 All power reactor Contaminated Material facilities with an OL or CP 81-06 Potential Deficiency Affecting 4/14/81 All power reactor Certain Foxboro 20 to 50 facilities with an Milliampere Transmitters OL or CP 81-05 Self-Aligning Rod End Bushings 3/31/81 All power reactor for Pipe Supports facilities with an OL or CP s
81-04 The Role of Shift Technical 4/30/81 All power reactor Advisors and Importance of facilities with an l
Reporting Operational Events OL or near-ter* O L.
l 81-03 Inoperable Seismic Monitoring 3/2/81 All power reactor l
Instrumentation facilities with an j
OL or CP 81-02 Performance of NRC-Licensed 2/9/81 All power recctor Individuals While on Duty
-facilities (research
& test) with an OL or CP OL = Operating Licenses CP = Construction Permit
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