ML19249B619
| ML19249B619 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 08/17/1979 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Parker W DUKE POWER CO. |
| References | |
| NUDOCS 7909040602 | |
| Download: ML19249B619 (2) | |
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'o, UNITED STATES E'g (f' 'g NUCLE AR REGULATORY COMMISSION y
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101 M ARIE TT A ST., N W., SUITE 3100
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ATL ANTA, GEORGI A 30303
' AUG 17 1979 50-369 50-370 Duke Power Company ATTN:
W.
O. Parker, Jr.
Vice President, Stean Production P. O.
Box 33189 Charlotte, North Carolina 28242 Gentlemen:
The enclosed Bulletin 79-05C and 79-06C is forwarded to you for action. Written responses are required.
If you desire additional inforr.ation regarding this matter, please contact this office.
Sincerely, w
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(q James P. O'Reilly Director
Enclosure:
- 2. '.,isting of IE Bulletins Issued In Last Twelve Months r:_,,.,
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Duke Power Company cc w/ encl:
M. D. McIntosh, Plant Manager Post Office Box 488 Cornelius, North Carolina 28031 J. C. Rcgers, Project Manager Post Office Box 33189 Charlotte, North Carolina 28242 (s ' ', f *, C
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UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 July 26, 1979 IE Bulletin Nos.79-05C & 79-06C NUCLEAR INCIDENT AT THREE MILE ISLAND - SUPPLEMENT Description of Circumstances:
Information has become available to the NRC, subsequent to the issuance of IE Bulletins 79~05,79-05A, 79-05B, 79-06,79-06A, 79-06A (Revison 1) and 79-06B, which requires modification to the " Action To Be Taken By Licensees" portion of IE Bulletins79-05A, 79-06A and 79-06B, for all pressurized water reactors (PWRs).
Item 4.c of Bulletin 79-05A required all holders of operating licenses for Babcock & Wilcox designed PWRs to revise their operating procedures to specify t ha t. in the event of high pressure injection (HPI) initiation with reactor coolant pumps (RCPs) operating, at least one RCP per loop would remain operating.
Similar requirements, applicable to reactors designed by other PWR vendors, were contained in Item 7.c of Bulletin 79-06A (for Westinghouse designed plants) and in Item 6.c of Bulletin 79-06B (for Combustion Engineering designed plants).
Prior to the incident at Three Mile Island Unit 2 (TMI 2), Westinghouse and its licensees generally adopted the position that the operator should promptly trip all operating RCPs in the loss of coolant accident (LOCA) s. uation. This Westinghouse position, has led to a series of meetings between the NRC staff and Westinghouse, as well as with other PWR vendors, to discuss this issue.
In addition, more detailed analyses concerning this matter were requested by the NRC.
Recent preliminary calculatior.s performed by Babcock & Wilcox, Westing-house and Combustion Engineering indicate that, for a certain spectrum of small breaks in the reactor coolant system, continued operation of the RCPs can increase the mass lost through the break and prolong or aggravate the uncover-ing of the reactor core.
The damage to the reactor core at TMI 2 followed tripping of the last operating RCP, when two phase fluid was being pumped through the reactor coolant system.
It is our current understanding that all three of the nuclear steam system suppliers for PWRs now agree that an acceptable action under LOCA symptoms is to trip all operating RCPs immediately, before significant voiding in the reactor coolant system occurs.
Action To Be Taken By Licensees:
In order to alleviate the concern over delayed tripping of the RCPs after a LOCA, all holders of operating licenses for PWR facilities shall take the following actions:
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IE Bulletin Nos.79-05C & 79-06C July 26, 1979 Page 2 of 3 Short-Term Actions 1.
In the interim, until the design change required by the long-term action of this Bulletin has been incorporated, institute the following actions at your facilities:
A.
Upon reactor trip and initiation of HPI caused by low reactor coolant system pressure, immediately trip all operating RCPs.
B.
Provide two licensed operators in the control room at all times during operation to accomplish this action and other immediate and followup actions required during such an occurence. For facilities with dual control rooms, a total of three licensed operators in the dual control room at all times meets the require-ments of this Bulletin.
2.
Perform and submit a report of LOCA analyses for your plants for a range of small break sizes and a range of time lapses between reactor trip and pump trip.
For each pair of values of the parameters, deter-mine the peak cladding temperature (PCT) which results. Tne range of values for each parameter must be wide enough to assure that the maximum PCT or, if appropriate, the region containing PCTs greater than 2200 degrees F is identified.
3.
Based on the analyses done under Item 2 above, develop new guidelines for operator action, for both LOCA and non-LOCA transients, that take into account the i= pact of RCP trip requirements. For Babcock &
Wilcox designed reactors, such guidelines should include appropriate requirements to fill the steam generators to a higher level, following RCP trip, to promote natural circulation flow.
4 Revise emergency procedures and train all licensed reactor operators and senior reactor operators based on the guidlines developed under Item 3 above.
5.
Provide analyses and develop guidelines and procedures related to in-adequate core cooling (as discussed in Section 2.1.9 of NUREG-0578, "TMI 2 Lessons Learned Task Force Status Report and Short-Term Recom-mendations") and define the conditions under which a restart of the RCPs should be attempted.
Long-Term Action 1.
Propose and submit a design which will assure automatic tripping of the operating RCPs under all circumstances in which this action may be needed.
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.c 3.79-05C & 79-06C July 26, 1979 Page 3 of 3 Schedu.e The schedule for the short-term actions of this Bulletin is:
Item ):
Effective upon receipt of this Bulletin, Item 2:
Within 30 days of receipt of this Bulletin, Item 3: Within 30 days of receipt of this Bulletin, Item 4:
Within 45 days of receipt of this Bulletin, Item 5:
October 31, 1979 (as noted in Table B-2 of NUREG-0578, under Item 3).
A schedule for the long-term action required by this Bulletin should be developed and subcitted within 30 days of receipt of this Bulletin.
Reports should be submitted to the Director of the appropriate NRC Regional Office with copies forwarded to the Director, Office of Inspection and Enforcement and the Director, Office of Nuclear Reactor Regulation, Washington, D. C. 2 0555.
Approved by GAO (ROD 72): clearance expires 7/31/80. Approval was given under a blanket cleersace specifically for generic problems.
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IE Bulletin Nos.79-05C & 7 v 06C Enclosure Page 1 of 3 July 26, 1979 LISTING OF IE BULLETINS ISSUED IN LAST WELVE MONTHS Bulletin Subject Date Issued Issued To No.
79-17 Pipe Cracks in Stagnant 7/26/79 All PWR's with Borated Water Systems at operating license PWR l'lant s 79-16 Vital Area Access Controls 7/26/79 All Holders of and applicants for Power Reactor Operating Licenses who anticipate loading fuel prior to 1981 79-15 Deep Draft Pump 7/11/79 All, Power Reactor Licensees with a CP Deficiencies and/or OL 79-14 Seismic Analyses for 6/2/79 All Power Reactor As-Built Safety-Related facilities with an OL or a CP Piping System OL for action. All System Piping BWRs with a CP for information.
79-02 Pipe Support Base Plate 6/21/79 All' Power Reactor Facilities with an (Rev. 1)
Designs Using Concrete OL or a CP Expansion Anchor Bolts 79-12 Short Period Scrans at 5/31/79 All GE BWR Facilities BWR Facilities with an OL 79-11 Faulty Overcurrent Trip 5/22/79 All Power Reactor Device in Circuit Breakers Facilities with an OL or a CP for Engineered Safety Systems 79-10 Requalification Training 5/11/79 All Power Reactor Facilities with an OL Program Statistics 79-09 Failures of GE Type AK-2 4/17/79 All Power Reactor Circuit Breaker in Safety Facilities with an OL or CP Related Systems 79-08 Events Relevant to BWR 4/14/75 All BWR Power Reactor Reactors Identified During Facilities with an OL Three Mile Island Incident
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I IE Balletin Nos.79-05C 4 /9-06C Enclosure July 26, 1979 Page 2 of 3 LISTING OF IE BULLETINS ISSUED IN LAST TWELVE MONTHS Bulletin Subject Date Issued Issued To No.
79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an OL or CP 79-06B Review of Operational 4/14/79 All Combustion Engineer-Errors and System Mis-ing Designed Pressurized alignments Identified Water Power Reactor During the Three Mile Facilities with an Island Incident Operating License 79-06A Review of Operational 4/18/79 All Pressurized Water (Rev 1)
Errors and System Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an OL Island Incident 79-06A Review of Operational 4/14/79 All Pressurized Water Errors and System Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an OL Island Incident 79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Mis-Power Reactors with an alignments Identified OL except B&W facilities During the Three Mile Island Incident 79-05A Nuclear Incident at 4/5/79 All B6W Power Reactor Three Mile Island Facilities with an OL 79-05 Nuclear Incident at 4/2/79 All Power Reactor Three Mile Island Facilities with an OL and CP a ~3
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..t-July 26, 1970 Page 3 of 3 LISTING OF IE BULLETINS LSSUED IN LAST IVELVE MONTHS Bulletin Subj ect Date Issued Issued To No.
79-04 Incorrect Weights for 3/30/79 All Power Recctor Swing Check Valves Facilities with an Manufactured by Velan OL or CP Engineering Corporation 78-12B Atypical Weld Material 3/19/79 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 79-03 Longitudinal Welds Defects 3/12/79 All Power Reactor In ASME SA-312 Type 304 Facilities with an Stainlesc Steel Pipe Spools OL or CP Manufactured by Youngstown Velding an_ Engineering Co.
79-02 Pipe Support Base Plate 3/2/79 All Power Reactor Designs Using Concrete Facilities with an Expansion Anchor Bolts OL or CP 79-01A Environmental Qualification 6/6/79 All Power Reactor of Class 1E Equipment Facilities with an (Deficiencies in the Envi-OL or CP ronmental Qualification of ASCO Solenoid Valves) 79-01 Environmental Qualification 2/8/79 All Power Reactor of Class IE Equipment Facilities with an OL or CP
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