ML17053B244
| ML17053B244 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 11/20/1979 |
| From: | Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Rhode G NIAGARA MOHAWK POWER CORP. |
| References | |
| NUDOCS 7912050261 | |
| Download: ML17053B244 (12) | |
Text
1 ~<'KCy
)'~G~
Ip0 Cy
~g Y/40 'O
++*++
Docket No. 50-410 UNITED STATES NUCLEAR REGULATORY COMMISSION REGION I 631 PARK AVENUE KING OF PRUSSIA, PENNSYLVANIA19406 HOV 20 1979 Niagara Mohawk Power Corporation ATTN:
Mr.
G.
K.
Rhode Vice President System Project Management 300 Erie Boulevard, West
- Syracuse, New York 13202 Gentlemen:,
The enclosed IE Bulletin 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, oyce'.
Grier
~
~
erector
Enclosures:
1.
List of Recently Issued IE Bulletins CONTACT:
D.
L. Caphton (215-337-5266) cc w/encls:
Eugene B. Thomas, Jr.,
Esquire poW ioxa050 2<>
\\
ENCLOSURE 1 UNITED STATES SSINS:
6820 NUCLEAR REGULATORY COMMISSION Accession No.:
OFFICE OF INSPECTION AND ENFORCEMENT 7910250475 WASHINGTON, D. C.
20555 BORON LOSS FROM BWR CONTROL BLADES Description of Circumstances:
IE Bulletin No. 79-26 Date:
November 20, 1979 Page 1 of 4 The General Electric Company (GE) has informed us of a failure mbde 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
) 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 (B C) particles and swelling of the compacted B~C as helium and lithium concentrations grow.
Once primary coolant penetrates tNe cladding (i.e., the cracking has progressed through the cladding wall and the helium-lithium pressures are sufficient to open the crack), (~ron is leached out of the tube at locations with more than 50 percent B
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 8
, 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 with of the blade is reduced by 10 percent, which corresponds to 42 percent B
depletion averaged over the top quarter of the control>)lade.
Because of the leaching mechanism, GE has reduced the allowance for B 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.
Enclosure 1
IE Bulletin No. 79-26 Date:
November 20, 1979 Page 2 of 4 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.
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 base)0on the hot cell result that no boron loss is observed until 50 percent local 8
depletion is attainy).
For each B
C tube, complete loss of 84C was assumed when the calculated B
>fepletion exceeded 50 percent locally.
For any blade expected to ~each a
B depletion greater than 34 percent during a
- cycle, GE assumed a
8 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.
Me have examined the bases for GE's conclusions, including the hot cell tests and the calculational assumptions.
The preferred>~ction is to replace all blades expected to have greater than 34 percent 8
depletion averaged over the upper one-fourth of the blade.
- However, based on our review we believe the relation between boron loss and B
depletion (i.e., the observations to dge show that boron loss does not occur until 50 percent local depletion of B
) is sufficiently understood to justify BMR operation on an interim basis provided the following actions have been taken by licensees.
'I
Enclosure 1
IE Bulletin No. 79-26 Date:
November 20, 1979 Page 3 of 4 Action to be Taken by Licensees:
For all BWR power reactor facilities with an operating license:
2.
The operating history of0the reactor is to be reviewed to establish a
record of the current B
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.
Identify any control blades predicted to have greater that 34 percent B
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.
b.'escribe 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).
4.
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.
b.
C.
d.
e.f.
Tube number or identification.
The evaluation g each crack in the tubing.
The calulated
$0 depletion versus elevation for each tube.
The measured B
loss versus elevation for each tube.
The maximum local depletion for tubes having no cracks.
The maximum local depletion for tubes having no loss of boron.
Enclosure 1
IE Bulletin No. 79-26 Date:
November 20, 1979 Page 4 of 4 Alternately, the results of a destructive examination of a blade of similar fabrication and operational history may be provided within oneyO year of the date of issuance of this Bulletin. If the highest local B
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 (j.) 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.
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.
Bulletin No.
Subject ENCLOSURE 2 RECENTLY ISSUED IE BULLETINS Date Issued IE Bulletin No. 79-26 Date:
November 20, 1979 Page 1 of 1 Issued To 79"22 Possible Leakage of Tubes 9/5/79 of Tritium Gas Used in Timepieces for Luminosity Each Licensee who Receives Tubes of Tritium Gas in Timepieces for Luminosity 79-14 (Supplement No. 2) 79-23 79"24 Same as Title 79-14 Potenti al Fai lure of Emergency Diesel Generator Field Exciter Transformer Frozen Lines 9/7/79 9/12/79 9/27/79 Same as 79-14 All Power Reactor Faci,lities with an OL or CP All Power Reactor Facilities which have either OLs of CPs and are in late stage of construction 79-13 (Rev.
2)
Cracking in Feedwater System 10/17/79 Piping All PWRs with an OL and Designated Applicants (for Action), All Other Power Reactor Facilities with an OL or CP (for Information) 79-17 Pipe Cracks in Stagnant Borated Mater Systems at PMR Plants 10/29/79 All PWRs with an OL (for Action).
All other Power Reactor Facilities with an OL or CP (for Information) 79-25 79-02 (Rev, 2)
Failures of Mestinghouse 11/2/79 BFD Relays in Safety-Related Systems Pipe Base Plate Designs Using ll/8/79 Concrete Expansion Bolts All Power Reactor Facilities with an OL or CP (for Action)
All Power Reactor Facilities with an OL or CP
J 1
C