ML14170A531
| ML14170A531 | |
| Person / Time | |
|---|---|
| Site: | Robinson, Harris  |
| Issue date: | 11/20/1979 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Jackie Jones CAROLINA POWER & LIGHT CO. |
| References | |
| IEB-79-26, NUDOCS 7911300068 | |
| Download: ML14170A531 (9) | |
Text
S RoUNITED STATES 0
NUCLEAR REGULATORY COMMISSION REGION II 101 MARIETTA ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303
- OV 2 0 1979 In Reply Refer To:
RII:JPO 50-400, 50-401 5Q4-O-50-403 Carolina Power and Light Company ATTN: J. A. Jones Senior Executive Vice President and Chief Operating Officer 411 Fayetteville Street Raleigh, North Carolina 27602 Gentlemen:
The enclosed IE Bulletin No. 79-26 is forwarded.to you for information. No written response is required. If you desire additional information regarding this matter, please contact this office.
Sincerely, James P. O'Reilly Director
Enclosures:
- 1.
- 2.
List of IE Bulletins Issued In The Last Six Months 9113 00
Carolina Power and Light Company
-2 cc w/encl:
R. Parsons, Site Manager Post Office Box 101 New Hill, North Carolina 27562 R. B. Starkey, Jr., Plant Manager Post Office Box 790 Hartsville, South Carolina 29550
UNITED STATES SSINS:
6820 NUCLEAR REGULATORY COMMISSION Accession No.:
OFFICE OF INSPECTION AND ENFORCEMENT 7910250475 WASHINGTON, D.C. 20555 November 20, 1979 IE Bulletin No. 79-26 BORON LOSS FROM BWR CONTROL BLADES Description of Circumstances:
The General Electric Company (GE) has informed us of a failure mode for control blades which can cause a loss of boron poison material. Hot cell examinations of both foreign and domestic blades have revealed cracks near the upper end of stainless steel tubing and loss of boron from the tubes. The cracks and boron loss have so far been confined to locations in the poison tubes with more than 50 percent Boron-10 (B'0 ) local depletion. Observed crack sizes range from a quarter to a half inch in length and from one to two mils in width.
GE has postulated that the cracking is due to stress corrosion induced by solidification of boron carbide (B4C) particles and swelling of the compacted B4C as helium and lithium concentrations grow. Once primary coolant penetrates the cladding (i.e., the cracking has progressed through the cladding wall and the helium-lithium pressures are sufficient to open the crack), boron is leached out of the tube at locations with more than 50 percent B'0 local depletion (local depletion is considered to be twice the average depletion).
It was further found with similar cracking but with less than 50 percent local depletion of B10, that leaching did not occur even though primary coolant had penetrated the cladding.
The cracking and boron loss shorten the design life of the control blade.
According to the GE criteria the end of design life is reached when the reactivity worth of the blade is reduced by 10 percent, which corresponds to 42 percent B10 depletion averaged over the top quarter of the control blade. Because of the leaching mechanism, GE has reduced the allowance for B10 depletion averaged over the top quarter of the control blade from the 42 percent value to 34 percent.
The safety significance of boron loss is its impact on shutdown capability and scram reactivity. Although shutdown capability is demonstrated by shutdown margin tests after refueling, the calculated control blade worths used in the tests are based on the assumption that no boron loss has occurred. Reduction in scram reactivity due to boron loss could increase the severity of Critical Power Ratio (CPR) reductions during the plant transients and could increase the consequences of control rod drop accidents.
Because the locations of limiting Linear Heat Generation Rate (LHGR), CPR, and Average Planar LHGR (APLHGR) are not in controlled cells, local power limit monitoring is not affected by boron loss.
IE Bulletin No. 79-26 November 20, 1979 Page 2 of 4 GE has evaluated the potential effect of boron loss on shutdown capability, CPR reduction and the consequences of control rod drop accidents. GE's evaluation is based on the hot cell result that no boron loss is observed until 50 percent local B10 depletion is attained. For each B4C tube, complete loss of B4C was assumed when the calculated B10 depletion exceeded 50 percent locally. For any blade expected to reach a B10 depletion greater than 34 percent during a cycle, GE assumed a B10 depletion distribution typical of blades at the previously defined end of design life.
Based on these evaluations GE arrived at the following conclusions:
(a) Control rod drop accident consequences are not sufficiently sensitive to small reductions in scram reactivity to be affected by boron loss before the end of design life of the blades involved.
(b) If no more than 26 percent of the control blades have experienced a 10 percent reduction in projected worth taking boron loss into consideration, there is a negligible effect on transient CPR reduction and MCPR limits.
(c) If any control blades have experienced more than 10 percent reduction in projected worth, taking boron loss into consideration, the shutdown margin should be demonstrated to be at least the sum of the shutdown margin required by Technical Specifications plus an increment sufficient to account for the potential for boron loss.
We have examined the bases for GE's conclusions, including the hot cell tests and the calculational assumptions.
The preferred action is to replace all blades expected to have greater than 34 percent B10 depletion averaged over the upper one-fourth of the blade. However, based on our review we believe the relation between boron loss and B10 depletion (i.e., the observations to date show that boron loss does not occur until 50 percent local depletion of B'0 ) is sufficiently understood to justify BWR operation on an interim basis provided the following actions have been taken by licensees.
Action to be Taken by Licensees:
For all BWR power reactor facilities with an operating license:
- 1.
The operating history of the reactor is to be reviewed to establish a record of the current B10 depletion averaged over the upper one-fourth of the blade for every control blade; the record is to be maintained on a continuing basis. This action is required on all reactors whether shut down for refueling or operating.
- 2.
Identify any control blades predicted to have greater that 34 percent Blo depletion averaged over the upper one-fourth of the blade by the next refueling outage.
- a.
Describe your plans for replacement of identified control blades.
IE Bulletin No. 79-26 November 20, 1979 Page 3 of 4
- b.
Describe measures which you plan to take justifying continued operations until the next refueling specifically addressing (1) any blade with greater than 42 percent depletion averaged over the upper one-fourth of the blade; and (2) the condition where you find greater than 26 percent of the control blades calculated to have greater than 34 percent depletion averaged over the upper one-fourth of the blade.
- 3.
At the next cold shutdown or refueling outage, conduct shutdown margin tests to verify that:
- a.
full withdrawal of any control blade from the cold xenon-free core will not result in criticality; and
- b.
compliance with the shutdown margin requirement in a manner that accommodates the boron loss phenomenon (i.e., by including a plant specific increment in the shutdown margin that takes the potential loss of boron from control blades identified from evaluation of Item 1 into consideration).
Perform a destructive examination of the most highly exposed control blade at the end of the next cycle and provide results of the examination within one calender year after removal of the blade. The results to be reported should include:
- a.
Tube number or identification.
- b.
The evaluation of each crack in the tubing.
C.
The calulated B 10 depletion versus elevation for each tube.
- d.
The measured B10 loss versus elevation for each tube.
- e.
The maximum local depletion for tubes having no cracks.
- f.
The maximum local depletion for tubes having no loss of boron.
Alternately, the results of a destructive examination of a blade of similar fabrication and operational history may be provided within one year of the date of issuance of this Bulletin. If the highest local B10 depletion is less than 50 percent, this examination can be deferred until the next refueling.
- 5.
Submit within 45 days of the date of issuance of this Bulletin, a written report of the findings as to Items (1) and (2).
For facilities in a refueling outage, and all other facilities at their next refueling outage, submit the written report on Item (3) within 30 days after plant startup following the outage. A written report on Item (4) is requested within one year after removal of a control blade for destructive examination Reports should be submitted to the Director of the appropriate NRC Regional Office and a copy should be forwarded to the NRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D.C. 20555.
IE Bulletin No. 79-26 November 20, 1979 Page 4 of 4 For all BWR facilities with a construction permit and all other power reactor facilities with an operating license or construction permit, this Bulletin is for information only, no written response is required.
Approved by GAO B180225 (R0072); Clearance expires 7/31/80. Approval was given under a blanket clearance specifically for identified generic problems.
IE Bulletin No. 79-26 Enclosure November 20, 1979 Page 1 of 3 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Bulletin No.
Subject Date Issued Issued To 79-26 Boron Loss from BWR Control 11/20/79 ALL OLs for Action Blades All CPs for Information 79-25 Failures of Westinghouse BFD 11/02/79 All Power Reactor Relays in Safety-Related Systems Facilities with an OL or a CP 79-24 Frozen Lines 9/27/79 All Power Reactor Facilities with an OL or a CP 79-23 Potential Failure of Emergency 9/12/79 All Power Reactor Diesel Generator Field Exciter Facilities with an Transformer OL or a CP 79-22 Possible Leakage of Tubes of 9/5/79 Each Licensee who Tritium Gas Used in Timepieces Receives Tubes of for Luminosity Tritium Gas in Time pieces for Luminosity 79-21 Temperature Effects on Level 8/13/79 All PWR's with an Measurements Operating License 79-20 Packaging Low-Level Radioactive 8/10/79 All Materials Licensees Waste for Transport and Burial who did not receive Bulletin No. 79-19 79-19 Packaging Low-Level Radioactive 8/10/79 All Power and Research Waste for Transport and Burial Reactors with OLs, Fuel Facilities except uranium mills, and certain materials licensees 79-18 Audibility Problems Encountered 8/7/79 All OLs for Action on Evacuation of Personnel from All CPs for Information High-Noise Areas 79-17 Pipe Cracks in Stagnant Borated 10/29/79 All PWRs with (Rev. 1)
Water Systems at PWR Plants Operating License 79-17 Pipe Cracks in Stagnant Borated 7/26/79 All PWRs with Water Systems at PWR Plants Operating License
IE Bulletin No. 79-26 Enclosure November 20, 1979 Page 2 of 3 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Bulletin No.
Subject Date Issued Issued To 79-16 Vital Area Access Controls 7/26/79 All Holders of and applicants for Power Reactor Operating Licenses who Antici page loading fuel prior to 1981 79-15 Deep Draft Pump Deficiencies 7/18/79 All Power Reactor (Supp. 1)
Licensees with a CP and/or OL 79-15 Deep Draft Pump Deficiencies 7/11/79 All Power Reactor Licensees with a CP and/or OL 79-14 Seismic Analyses for As-Built 9/7/79 All Power Reactor (Supp. 2)
Safety-Related Piping System Facilities with an OL or a CP 79-14 Seismic Analyses for As-Built 7/27/79 All Power Reactor (Correction)
Safety-Related Piping System Facilities with an OL or a CP 79-14 Seismic Analyses for As-Built 7/18/79 All Power Reactor (Rev. 1)
Safety-Related Piping System Facilities with an OL or a CP 79-14 Seismic Analyses for As-Built 7/2/79 All Power Reactor Safety-Related Piping System Facilities with an OL or a CP 79-13 Cracking in Feedwater System 10/17/79 All PWR's with an (Rev. 2)
Piping Operating License 79-13 Cracking in Feedwater System 8/30/79 All PWR's with an (Rev. 1)
Piping Operating License 79-13 Cracking in Feedwater System 6/25/79 All PWR's with an Piping OL for action. All BWRs with a CP for information
IE Bulletin No. 79-26 Enclosure November 20, 1979 Page 3 of 3 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Bulletin No.
Subject Date Issued Issued To 79-06C Nuclear Incident at Three Mile 7/26/79 To all PWR Powesr Island - Supplement Reactor Facilities with an OL 79-05C Nuclear Incident at Three 7/26/79 To all PWR Power Mile Island - Supplement Reactor Facilities with an OL 79-02 Pipe Support Base Plate Designs 11/8/79 All Power Reactor (Rev. 2)
Using Concrete Expansion Anchor Facilities with an Bolts OL or a CF 79-02 Pipe Support Base Plate Designs 8/20/79 All Power Reactor (Rev. 1)
Using Concrete Expansion Anchor Facilities with an (Supp. 1)
Bolts OL or a CF 79-02 Pipe Support Base Plate Designs 6/21/79 All Power Reactor (Rev. 1)
Using Concrete Expansion Anchor Facilities with an Bolts OL or a CP 79-01A Environmental Qualification of 6/6/79 All Power Reactor Class 1E Equipment (Deficien-Facilities with an cies in the Environmental OL or a CP Qualification of ASCO Sole noid Valves)