ML20033E040
| ML20033E040 | |
| Person / Time | |
|---|---|
| Issue date: | 01/10/1990 |
| From: | Sniezek J Office of Nuclear Reactor Regulation |
| To: | Jordan E Committee To Review Generic Requirements |
| References | |
| IEIN-89-069, IEIN-89-69, NUDOCS 9001170496 | |
| Download: ML20033E040 (11) | |
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JAN 101990 MEMORANDUM FOR: Edward L. Jordan, Chairman
- Committee to Review Generic-Requirements FR0th James H. Sniezek, Deputy Director Office of. Nuclear Reactor Regulation i
SUBJECT:
PROPOSED GENERIC LETTER TO CORRECT A DEFICIENCY IN'THE BWR CRITICAL-POWER RATIO CALCULATION INVOLVING CHANNEL.B0X BOW t
NRR requests that the Comittee to Review Generic Requirements'(CRGR). review the enclosed proposed generic letter-at the-CRGR's. earliest convenience. We -
would like to issue this generic letter soon in order to facilitate the' pro-posed licensee implementation date'.
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t The proposed generic letter would implement existing requirements and. staff positions in that it would correct a known deficiency in-calculating the-critical power ratio (CPR). This deficiency results from greater than expected.
BWR channel box bow effect which reduces the CPR to lower values than predicted by calculations.
If this deficiency is not corrected-in a timely manner, it' could mislead operators about the technical specifications operating limit-margin and, in the worst case, could result in violations of the CPR safety limit with potential fuel failure. The deficiency became known as a result of dryout with fuel failures in a foreign BWR.
This generic letter is addressed to all holders of operating licenses or t
construction permits for BWRs and is sponsored by Ashok Thadani, Director, Division of Systems Technology.
The proposed generic letter and background information required by the CRGR
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charter are enclosed.
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Ja es H. Sniezek, Deputy Director Of ice of-Nuclear Reactor Regulation
Enclosures:
As stated CONTACT: Peter C. Wen, NRR 492-1172 h 764f D KA g
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TO:
ALL HOLDERS OF OPERATING LICENSES OR CONSTRUCTION PERMITS FOR BOILING-WATER. REACTORS (BWRs)
SUBJECT:
CORRECTION OF DEFICIENCY IN BWR CRITICAL POWER RATIO CALCULATION DVE TO CHANNEL BOX BOW i
As a result of information obtained at a meeting on fuel failures caused by j
dryout at a foreign BWR facility and at meetings with BWR fuel vendors..the NRC' issued Information Notice 89-69 on September 29, 1989, to alert addressees to the potential problems associated with excessive channel box bow that could' result in a loss of thennal margin. The generic concern applicable to the U.S.--
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BWRs is that the channel box bow effect is not properly taken into account in the critical power ratio (CPR) calculation. Each BWR fuel vendor plans.to i
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submit a methodology to address this concern. A copy of NRC Information Notice 89-69, " Loss of Thermal Margin Caused by Channel Box Bow," is attached.
Because the possibility of fuel failure as a result of channel box' bow exceed -
ing that assumed in the CPR calculation is remote, the staff has agreed that no j
1 inmediate corrective actions are needed. However, in order to verify compli-ance with the current licensing basis, the staff has concluded that the validi-ty of the CPR calculation used to ensure conformance to the plant technical-specifications should be evaluated and corrected, if necessary,- in a timely Therefore, the staff is requesting that each operating reactor -
manner.
2 licensee ensure that the effects of channe. box bow'on the CPR calculation are properly taken into account in the first fuel reload-scheduled after April 30, j
1990. This may be accomplished by either of the following two options:
1 A.
By ensuring that the procedures for monitoring thermal limits during reactor operation will prevent violation of-the minimum CPR operating limit by the use of an NRC approved rethodology that takes into account the new data on fuel channel bowing, or B.
By imposing on the operating CPR limit (or limiting curve) a channel bow penalty (or S) lattice plants and by removing any channel boxes that areof eith for C being reused after their first bundle lifetime.
All construction permit (CP) holders are requested to complete the above i
actions before the date scheduled for fuel loading.
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' Generic Letter 89-,
To determine if any license or construction permit for facilities covered by this request should be modified, suspended, or revoked, the. staff re l
pursuant to Section 182 of the Atomic Energy Act and 10 CFR 50.54(f) quires.
, that each operating reactor licensee advise the NRC by letter (1) whether it will implement one of the options listed for the first reload scheduled after A3ril 30, 1990, including the.date of the reload, and (2) whether or not c1annel boxes are being used for a second burjdle lifetime.
If the licensee does not intend to adopt either option A or B above, it should include in its response a justification for its position. The NRC staff requires each licensee to provide a response within 60 days of receipt of this letter.
Before fuel loading, CP holders.for BWRs are requested to advise the NRC in writing that the actions requested in this letter have been implemented.
The written response required above shall be addressed to the U.S. Nuclear Regulatory Comission, ATTN:
Document Control Desk, Washington, D.C. 20555, under oath or affirmation under'the provisions of Section 182a, Atomic Energy i
Act of 1954, as amended, and 10 CFR 50.54(f).
In addition, a copy shall be submitted to the appropriate Regional Administrator.
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The actions requested in this generic letter ensure licensees' comp (liance with their technical specifications as required under 10 CFR 50.36(c)(1) 1)(A).
The CPR is a safety limit under this regulation. A known deficiency in calculating the CPR must be corrected to remain in compliance with the licensing basis.
3 If this deficiency is not corrected in a timely-manner, it could mislead operators about the technical specifications operating limit margin and, in the worst case, could result in violations of the CPR safety limit with potential fuel failure. The requested action was evaluated consistent with the provisions of 10 CFR 50.109 and found to be covered by the provisions of paragraph (a)(4)(1).
This request is covered by Office of Management and Budget Clearance Number 3150-0011, which expires January 31, 1991. The estimated average burden hours is 100 person-hours per licensee response, including assessment of the new recomendations, searching data sources, gathering and analyzing the data, and-preparing the required letters.
These estimated average burden hours pertain only to the identified response-related matters and do not include the time for actual implementation of the requested actions. Send comments regarding this-burden estimate or any other aspect of this collection of information, includ-ing suggestions for reducing this burden, to the Information and Records Management Branch, Division of Information Support Services, Office of Informa-tion Resources Management, U.S. Nuclear Regulatory Commission, Washington D.C.
l 20555; and to the Paper Reduction Project (3150-0011), Office of Managemen,t and Budget, Washington, D.C. 20503.
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' Generic Letter 89- ;
If you have any questions about-the actions requested in this generic letter, l
please contact one of the technical contacts listed below or the appropriate NRR project manager.
Sincerely.
James G. Partlow
-Associate Director for Projects Office of Nuclear Reactor Regulation
Enclosures:
1.
NRC Information Notice 89-69) "1.oss of Thennal Margin Caused by Channel Box Bow '
2.
List of Recently Issued NRC Generic Letters Technical Contacts:- Peter C. Wen, NRR (301)492-1172 Daniel B. Fieno, NRR (301)492-3236 l
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l Material Related to-CRGR Meeting No.'179 4
to be made Publicly Available i
1.
Memo dated March-30, 1990 for J. Taylor from E. Jordan, subject-Minutes of CRGR Meeting Number 179, including two enclosures which were-not previously released:
- a., a summary of discussions of proposed. bulletin on~1oss l
of fill-oil in transsitiers manufactured by Rosemount,. including 3 attachments.
- b., a summary of discussions of a proposed generic letter to correct a deficiency in BWR critical power ratio calculations-involving. channel box low, I attachment.
2.
Memo dated January 29, 1990 for E. Jordan from J.-Sniezek forwarding.
review materials on a proposed bulletin of loss of.' fill oil in-transmitters manufactured by-Rosemount.
3.
-Memo dated January
,L1990 for E.. Jordan from J. Sniezek forwarding _
review materials on a proposed generic letter to correct a dpficiency in BWR critical power ratio calculations involving channel box dow.
Sent to PDR on:
4/27/90 1
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Enclosure t
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l CRGR REVIEW PACKAGE j
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.j RESDJNSE 'IO REXXJIRDENTS FOR CONTENT OF PACKAGE SUBTITED FGt OGR REVIBf i
(1)
'Ihe prrpnaai generic requirement or staff position as it is paw to be sent out to licensees.
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'Ihe staff position is provided in the prepn==1 Generic Istter that
'I will be sent to all boiling water reactor (BWR) licensees and applicants. It inform licensees arvi applicants of the loss of thermal margin caused by channel box bow arx1 requires actions to be taken to prevent potential violations of the safety limit mininum critical power ratio (SINCPR).
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l (ii)
Draft staff papers or other urderlying staff h= ants supporting the requirements or staff positions.
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'Ihis section of the regulations requires safety limits to be set for inportant process variables which are found to be m=a_v'y to reasonably protect the integrity of the physical barriers that guard s
against the uinad.wlled release of radioactivity.
Stardard Review Plan Section 4.2 i
Section 4.2.II.A.2.d diaen w the need for a cladding overheating-criterion to prevent fuel failures during normal operation and for anticipated operational transients.
Standard Review Plan Section 4.4 Section 4.4.II reiterates the Standard Review Plan Section 4.2 IW11rements on overheating the cladding.
'Ihe Information Notice informs all BWR licensees and applicants on the loss of thermal margin caused by channel box bow.
(iii)
Each proposed requirement or staff position shall contain the i
sponsoring office's position as to whether the proposal would increase requirements or staff positions, inplement existing
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I requirements or staff positions, or would relax or reduce existing requirements or staff positions.
'Ihe proposed Generic letter would implement existing requirements and staff positions in that it would correct a known deficiency in calculating the critical power ratio (Cm). 'Ihis deficiency is not taking &annel box bow into account in the calculations.
(iv)
'Ibe iaW method of implementation with the canourrence (and any cannants) of OGC on the method proposed.
'Ibe requinments of the pu--:zi Generic Istter would be implemented in the first reload after April 30, 1990 by assuring that the effects of & annel box bow are taken into acocunt in determining the critical power ratio. OGC has no legal &jection to this proposal and their cannants were incorporated.
(v)
Regulatory analyses conforming to the directives and guidance of NURD:/BR-0058 and NURB /CR-3568.
A fornal regulatory analysis is not required because the at--: :1 n
Generic letter does not impose any new positions or requirements.
'Ihe actions pixposed by the Generic letter would maintain aarrant regulatory criteria on cladding overheating. Not taking the actions prtponed by the Generic Istter is likely to result in violation of CPR operating limit technical specifications and ney result in violation of the CPR Safety Limit with potential fuel failures.
(vi)
Identification of the category of reactor plants to Wnich the generic requirement or staff position is to apply.
'Ihis NW Generic Istter is a;plicable to all Boiling Water Reactors.
(vii)
Ibr each such category of reactor plants, an evaluation which h.r.Lcates how the action should be prioritized and scheduled in light of other ongoing regulatory activities. 'Iha evaluation shall haaqt for consideration information available wcuning any of the following factors as ney be appropriate and any other information relevant and material to the pm-: 71 action:
(a) Statement of the specific objectives that the proposed action is designed to achieve;
'Ibe proposed action is being taken to ensne that the SIMCm is not violated so that fuel failures will not occur. 'Ibe pweceed action also ensures that the requirements of 10 CFR 50.36(c)(1)(i) A are net.
(b) General description of the activity that would be required by the licensee or applicant in order to canplete the action; ;
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t The licensees would have to modify the process omputer evaluation of the critical power ratio either by irwrgaating i
& anges to parameters used in the calculation of critical power l
ratio or by ---=%g a bourviing estimate of the magnitude of i
the channel bow effect on the critical power ratio.
j (c)
Potential &ange in the risk to the publio frtan the accidental release of radioactive material; i
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The actions required by the prtponed Generic Intter would i
reduos the risk to the public by reducing the risk of fuel l
failures caused by cladding overheating.
(d) Potential inpact on radiological exposure of facility enployees and other onsite workers, i
The actions required by the proposed Generic Imtter would reduon the risk to facility enployees and onsite workers by reducing the risk of fuel failures causad by cladding j
overheating.
(e) Installation and continuing costa nameiated with the actico, including the oost of facility downtime or the oost of cw duoction delay; J
Each BWR will prtbably require a plant-specific analysis to l
determine if sufficient margin is available to account for fuel dannel box bow in the process omputer determination of the i
l critical power ratio, scune plants will probably require additional analyses for # 7-nt reloads. It is expected r
that s rcdve actions will not be needed for several plants..
The most severe econcanic impact would be for those plants operating near the CPR operating limit which elect to impose a direct CPR penalty on the operating limit.
In those cases, tlw i
estimated potential inpact is yp to $500,000 per 0.01 cm penalty for fuel cycle and power reduction oosts. However, it is expected that a less costly ansdytical geoolution will be selectas fcr those plants.
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r (f) The potential safety inpact of changes in plant or operational e
emplexity, including the relationship of purM and existing regulatory requirements and staff positions; There will be no potential inpact of in plant operational ocuplexity. The actions by the pu g r=4 Generic Istter will require that parameters used in the critical power correlation be modified so that the puc.ss l
ocmputer evaluatice of the critical power ratio will include the effects of channel box bow. Tne modified parameters would be supplied by the fuel vendors as part of the ypdating of Pr=== omputer parameters during refueling outages. '
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(g) The estimated resource burden on the NRC associated with the
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proposed action and the availability of resources; j
The resource kurden to the NRC would involve the review of two topical reports, one from General Electric and one frun l
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Advanced Nuclear Fuels Corporation, on the nothodology develcped to account for the effect of channel box bow on the i
critical power ratio determination. Any alternata proposed by 4
individual licensee would also need review by the staff. Once the Generic Istter is issued, then each affected plant's project manager would monitor:. the inplenantation of apptwed nethodology and gucedares at his plant on the schedule required by the Generic Istter. This effort by the project managers is judged to be minimal. The ramawoes required to rwiew the two topical reports are available in the Reactor Systans BranWDST/NRR (not expected to aw=mi twc m Jhs),
i in addition, pp to sixinan-ncmths NRR resources may be r==imi to review all licensee actions for conformance to this generic letter.
o (h) The potential impact of diffaronoes in facility type, design or -
4 age on the re3evancy and practicality of the proposed action; _
i The gaidance in the proposed Generic Istter is applicable to all BWRs because the only important parameter is the effect of channel bow on the critical power. ratio. Other factors such as facility type, design, or age have no effect on the proposed guidance of the Generic latter.
(i) Whether the proposed action is interim or final, and if interia, the justification for imposing the pwr-ed action on an interim basis.
This is the final staff position in that no further changes are under consideration.
(viii)
For each evaluation conducted pursuant to 10 CFR 50.109, the proposing office Director's determination, together with the rationale for the determination based on the consideration of paragraph (1) and (vii) above, that:
l (a) There is reasonable increase in the overall protection of public health and safety or the comon defense and =mwity to be deriva:1 from the gur.al; and
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(b) '1he direct and iniirect oosts of inplementation, for the l
l facilities affected, are justified in view of this irnd protection.
l Because the guidance in the p..M ceneric tetter corrects a known deficiency in the calcalation of the critical power i
l ratio, there is an increase in the overall protection of public I
bealth ard safety, backfit considerations are not applicable.
'Ibe direct and indirect oosts of implementation are judged to i
be minimal.
l (ix)
For endt evaluation conducted for p --:
' relaxations or decramman i
in current requirements or staff positions,.the proposing office l
Director's determination, together with the rationale for the i
I determination ha d on the considerations of paiw r.[2mi (i) through (vii) above, that:
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(a) h public health-and safety and the ocanon defense and
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security would be adequately protected if the proposed reduction in rwpirements or positions were inplemented; and j
(b) The oost savirgs attributed to the action would be substantial enough to justify taking the action.
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This prh Generic Istter does not relex or decrease current requirements or staff positions.
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UNITED STATES I
NUCLEAR REGULATORY COMMIS$10N 0FFICE OF NUCLEAR REACTOR REGULATION i
WASHINGTON, D.C.
20555 i
September 29, 1989 l
NRC INFORMATION NOTICE NO. 89 69: LOSS OF THERMAL MARGIN CAUSED BY I
CHANNEL BOX. BOW i
Addressees:
All holders of operating licenses or construction permits for boiling-water reactors (BWRs).
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Purpose:
T This information notice is intended to alert addressees to potential problems involving loss of thermal margin caused by excessive bowing of BWR fuel channel boxes.
It is expected that recipients will review the information for appli-cability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice l
do not constitute NRC requirements; therefore, no specific action or written i
response is required.
Description of Circumstances:
During a refueling outage in August 1988, four failed fuel rods in separate assemblies were identified at a foreign BWR facility.
Subsequent evaluation of sipping, visual inspection, gama scan, and hot-cell data led to the con-clusion that these rods failed because the rods were operated under dryout conditions during steady-state operation for an extended period of time (between2and7 days),
The failed fuel rods were located symetrically in the core. The fuel assem-t blies containing the rods that had failed were located adjacent to once-burned fuelassemblieswithhighlyexposedfuelchannels(seeFigure1).
These fuel channels were in their second bundle lifetime and had excessive channel bowing.
I In each assembly with failed fuel, the corner rod facing the adjacent control rod was heavily oxidized and the cladding was penetrated just below the top spacer grid.
In addition, each of the four failed rods had typical secondary internal hydriding damage near the bottom of the fuel rods, resulting in loss of fuel material.
i Discussion:
Dryout of the feel rods in this foreign facility occurred because of modeling errors in the plant process computer, which resulted in nonconservative calcu-latedvaluesoftheminimumcriticalpowerratio(MCPR)ofthecore. These
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l IN 89-69 I
September 29, 1989 1
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modeling errors were caused by neglecting the effects of channel bowing and the geometric variation between the reloaded and once-burned fuel assemblies.
These effects substantially increased the widths of the control rod water gaps i
for the assemblies that contained these four fuel rods beyond that assumed in the plant process computer calculatiors.
The increased neutron moderation as.
sociated with the increased water gap widths led to very high localized power peaking at these four fuel rods. However, these effects were not properly ac-counted for in the MCPR calculations.
For some time, the plant operators were misled by these erroneous MCPR calculations and were operating the plant in steady-state beyond the HCPR safety limit.
The modeling error of generic concern to all BWRs, regardless of the fuel i
i supplier, relates only to the greater-than-expected bowing of fuel channel boxes, which contributed about 15 percent error in the calculated MCPR value l
for this foreign facility. Channel bowing is a manifestation of differences in the channel growth of opposite sides of the channel box and is proportional to channel growth. The information obtained by the NRC indicates that the channel growth shows an accelerated trend at higher burnup exposure, especially when the fuel channels are being reused in their second bundle lifetime.
The effect on core operating MCPR is magnified when fresh fuel is located adjacent to the bowed fuel channels. Core operating limits imposed by technical speci-fications may be exceeded if the reduction in n.argin caused by fuel channel bowing is not properly accounted for in the plant process computer for thermal limits monitoring.
Based on a preliminary evaluation by BWR fuel vendors of U.S. reactors, the impact of the new data on actual versus calculated MCPR values is exsected to range from 0.0 to 0.03 CPR units.
However the impact l
could be muc1 greater (about 15 percent) for any reactors operating with fuel channels being reused in their second bundle lifetime.
This information notice requires no specific action or written response.
If you have any questions about the information in this notice, please contact one of the technical contacts listed below or the appropriate NRR project manager.
bkm' 44 Charles E. Rossi, )irector Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical Contacts: Peter C. Wen, NRR (301)492-1172 Daniel B. Fieno, NRR (301)492-3236 Attachments:
1.
Figure 1, " Channel Bow
- 2.
List of Recently Issued NRC Information Notices
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September 29, 1989 j
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l IN 89-69 September 29, 1989 Page 1 of 1 LIST OF RECENTLY ISSUED NRC INFORMATION NOTICES Information Date of Notice No.
Subject Issuance issued to 89-68 Evaluation of Instrument 9/25/89 All holders of OLs Setpoints During or cps for nuclear Modifications power reactors.
89-67 Loss of Residual Heat 9/13/89 All holders of OLs Removal Caused by or cps for PWRs.
Accumulator Hitro9en Injection 89-66 Qualification Life of 9/11/89 All holders of OLs Solenoid Yalves or cps for nuclear power reactors.
88-46 Licensee Report of 9/11/89 All holders of OLs Supp. 4 Defective Refurbished or cps for nuclear Circuit Breakers power reactors.
89-65 Potential for Stress 9/8/89 All holders of Ols Corrosion Cracking in or cps for PWRs.
Steer. Cenerator Tube Ph gs Supplied by Babock and Wilcox 89-64 Electrical Bus Bar Failures 9/7/89
-All holders of OLs or cps for nuclear l
power reactors.
89-63 Possible Submergence of 9/5/89 All holders of OLs Electrical Circuits Located or cps for nuclear Above the Flood Level Because power reactors, of Water Intrusion and Lack of Drainage 89-62 Malfunction of Borg-Warner 8/31/89 All holders of OLs Pressure Seal Bonnet Check or cps for nuclear Valves Caused By Vertical power reactors.
Misalignment of Disk 89-61 Failure of Borg-Warner Gate 8/30/89 All holders of OLs Valves to Close Against or cps for nuclear Differential Pressure power reactors.
OL = Operating License CP = Construction Permit
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