ML20203G573
| ML20203G573 | |
| Person / Time | |
|---|---|
| Issue date: | 02/17/1999 |
| From: | Wen P NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| PROJECT-694 NUDOCS 9902190369 | |
| Download: ML20203G573 (8) | |
Text
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m February 17, 1999 l
MEMORANDUM TO: FILE l
FROM:
Peter C. Wen, Project Manager original signed by:
Generic issues and Environmental Projects Branch i
Division of Regulatory improvement Programs
SUBJECT:
DISCUSSION TOPICS FOR FEBRUARY 24,1999 MEETING WITH l
WESTINGHOUSE OWNERS GROUP REGARDING WCAP-14696, i
" CORE DAMAGE ASSESSMENT GUIDANCE" j
The attached message was faxed today to Robert Lutz of the Westinghouse Owners Group (WOG). The sole purpose of the message is to prepare WOG personnel for an NRC/WOG meeting to be held on February 24,1999. The message itself does not constitute a formal request for information or represent a formal staff position; the message will be included in the I
meeting summary.
Project No. 694
Attachment:
As stated cc w/att: See next page l
DISTRIBUTION:
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r.p, Westinghouse Owners Group Project No. 694 cc:
Mr. Nicholas Liparulo. Manager Equipment Design and Regulatory Engineering Westinghouse Electnc Corporation Mail Stop ECE 4-15 P.O. Box 355 Pittsburgh, PA 15230-0355 Mr. Andrew Drake, Project Manager Westinghouse Owners Group Westinghouse Electric Corporation Mail Stop ECE 5-16 P.O. Box 355 Pittsburgh, PA 15230-0355 Mr. Jack Bastin, Director Regulatory Affairs Westinghouse Electric Corporation 11921 Rockville Pike Suite 107 Rockville. MD 20852 Mr. Hank Sepp, Manager Regulatory and Licensing Engineering Westinghouse Electric Corporation PO Box 355 Pittsburgh, PA 15230-0355 l
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1 ATTACHMENT NRC Comments / Questions on WCAP-14696. " Westinghouse Owners Group Core Damage Assessment Guidance" to be discussed on 2/24/99 NRC/WOG Meeting Comments For Discussion P4 As discussed in Section 1.3, some utilities use the core damage assessment methodology (and/or the post-accident sample system) to select the source term to be used in offsite dose assessment. For those plants, the results of the core damage assessment guideline (percent core damage) would need to be translated into a source term estimate. However, the same fission product (FP) behavior that complicates core damage assessment (e.g., holdup in the reactor coolant system, containment, and sumps) would need to be accounted for in this translation. The methodology should discuss how the core damage assessment results would be used to determine source terms at these plants, especially if information presently provided by the post-accident sample system (PASS) is no longer available.
P15, para 3 The document states that an indicated temperature of 12000F can be translated to a peak cladding temperature of 14000F. Does this imply that an indicated temperature of 15000F can be translated to a peak cladding temperature of 17000F7 What if the indicated temperature was 19000F7 Do factors such as system pressure, or extended periods of core uncovery result in a different indicated-to-actual temperature relationship? How will the operator or TSC Core Damage Assessment Team know when an instrument has failed?
P21, para 5 The document states that "the NUREG-0737 requirements for a core damage assessment based on samples of plant fluids does not have a valid basis using the current understanding of fission product behavior and the progression of core damage accidents." In an ideal situation, this may be true. But in reality many of the on line instruments may not be reliable and can not be the only source to obtain the status of the core. The grab samples are critical even though the results may not be available within a short time.
P33, para 4 Describe how the PASS systems meet design specifications of NUREG-0737 and Regulatory Guide 1.97 in view of the deficiencies identified.
P37, para 3 Describe what evidence exists to demonstrate that the RCS pressure / temperature relationship provides a reliable basis for judging whether clad rupture will occur. Justify that the codes do not co1tain conservatisms that may tend to over-predict clad failure. Should address this through a validation activity.
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P45, para 2 The WCAP states that for containment radiation levels corresponding to rupture of 60% of rods, little or no FP release from fuel pellet matrix should be occurring.
The basis for the 60% value is not clear. Previous discussion and figures center on a 0.5% pellet release and show that radiation levels for a 0.5% pellet release l
would correspond to about a 20% gap release. However, the guideline and the value for CRM2 seems to be based on a 1% pellet over-temperature release.
Clarify whether this is the cause of the disparity and make the text and guideline consistent.
1 P50, item 2 We disagree with the statements that "the results of analyses of samples of plant fluids would not provide any clarification of the type and degree of core damage compared to that obtained using fixed in-plant instrumentation. Thus, the need for post accident sampling of plant fluids for the determination of the type and degree of core damage is not required by the new WOG Core Damage Assessment Guideline." Although the results from sampling may not be available in a short time for any decisions, they would be very helpful for confirmation of the readings given by the in-plant instrumentation. The results of the grab samples can be used for the determination of the type and degree of core damage.
P2of10 It seems that the revised guideline is relevant only while the event is continuing to deteriorate (since CETs and RCS pressures are different then at the time of core damage, and no longer indicative of what had occurred earlier), and that following recovery, some other type of assessment (i.e., PASS or grab samples) would be needed. it is strongly recommended that the scope of the guideline be expanded to: (1) include core damage assessment during both the core degradation and accident recovery phases, and (2) address the respective roles of CDAM and PASS / grab samples during each phase. As such, the guideline would represent an integral approach for assessing the state of the core during and followina core degradation.
P3of10 It is not clear that evaluation of clad rupture based on pressure / temperature criteria is valid in practice. Some type of validation of this concept, as well as the recommended values for RCP2, CET3, and CET4, appears necessary.
P18of27 Recommend retaining the CH1 setpoint for ice condensers since this hydrogen concentration is well below lower flammability limit and would not be impacted by igniter operation. Also, rather than deleting the CH2 setpoint for ice condensers, the TSC should consider whether igniters are actuated, and whether there is evidence of a burn.
P190f27 in establishing CH2, CH3, CH4, and CH5, the WCAP recommends certain assumptions regarding the amount of metal-water reaction and fraction of hydrogen released to containment. The validity of the recommended values should be illustrated by comparing these values with the results from best estimate code calculations (MAAP, MELCOR, SCDAP) for representatbe severe accident sequences.
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e-Additional Comments (to be discussed selectively as time permits)
P2, para 4 Although there is no specific regulation for core damage assessment,10 CFR 50.47(b)(9) requires licensees to have " adequate methods, systems... " Part of this assessment process may include core damage assessment if it is used as part of the licensees emergency plan to make protective action recommendations or to classify events requires emergency action levels (EALs).
P3, para 6.
Incorrect statement regarding activity levels used as basis for EAL for RCS clad barrier (it is based on normal coolant activity)
P4, para 1 incorrect statement regarding the amount of clad damage (5-10% should be 2 -
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5%)
P5, para 4 Should be updated to reflect latest RTM (96) information.
I P7 Table 1 should have another column with " Indicated Core Exit Temperature".
P9, parai Should define which FP species are considered "non-volatile". The last sentence is not true if non-coolable geometry forms or molten pool is retained due to external reactor vessel ecoling.
P9, para 2 Next to last sentence seems *; say that any tellurium (Te) or small portions of non-volatiles would indicate that the core is ex-vessel. This is not true, as evidenced by the early in-vessel source term in NUREG-1465, and should be clarified.
P11, para 4 Provide basis for following statement: From the perspective of potential offsite releases of fission products and the need to recommend offsite protective actions, there are only three levels of core damage that are important: no damage, fuel rod cladding damage and fuel over-temperature damage.... Thus,
the diagnosis of core melting as specified in reference 4 will not be developed as part of this core damage assessment..
P12, para 1 Gravitational settling is said to be primary removal mechanism. Is this based on IDCOR or most recent work? If the former, confirm that this statement is supported by results of the more recent work.
j P13, para 4 The document states that analysis of samples of reactor coolant, containment sump water and containment atmosphere for specific radionuclides could be useful in estimating the core damage. This statement is in direct f.onflict with the j
final conclusion given on page 50.
P18, para 3 The statement that there is no information related to the delay time for the hydrogen reading to stabilize appeart, incorrect. Explain why this type of information isn't already available trom hydrogen monitor calibration testing.
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P18, para 4 Should include some discussion of how auto-ignition, random ignition, or diffusion flames in large dry containments may also result in low hydrogen concentrations not indicative of the degree of core camage.
P19, para 3 The requirement is for sampling and analyzing within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of the decision to j
do so (not accident initiation). We are not aware of any plant who has a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i
samph nquirement.
P22, para 2 The document indicates that a 0.1% clad damage is used for EAL classification.
Picase describe how this level of clad damage would generally be detected.
P22 para 3 Should note that dose projection should not just take into account CDA but also PASS results.
P24, para 2,3 The discussion seems to over-emphasize the amount of information that can be inferred from core exit thermocouples (CETs). This level of discussion seems unnecessary since the operators will not be able to discern the differences between sequences based on the CET data (given that it is only available up to about 2100F), and since the guidelines do not include such guidance / instructions.
1 P25-28 Figures 2 and 3 present " core exit thermocot:ple indications" values which would not be possible for the existing instruments. Should indicate that these are
" predicted peak fuel temperatures" rather than " core exit thermocouple indications."
l P29, para 1 The discussion seems to imply that operators wc aid assign a different level of reliability to CET readings based on RCS pressure. This level of discussion j
seems unnecessary since the guidelines do not include such j
guidance / instructions.
l P29 Should include some discussion of hydrogen production from radiolysis and corrosion, and how it compares to: (1) production from clad over-temperature, and (2) the 1% hydrogen concentration value discussed in Section 6.1.
PS2, para 1 The amount of cesium hydroxide in the containment atmosphere is said to be much higher in sequences with sprays or fans. Confirm that this is correct.
P39, para 3 The discussion seems to overstate the impact of instrument accuracy on low end readings. Isn't accuracy given as a function of measured value? The message could be incorrectly interpreted to mean that hydrogen concentration values less than 1% are not high enough to be taken seriously and used in the assessment.
P2of10 in the event of an SGTR or ISLOCA, containment radiation monitors will not provide usefulinformation. The assessment of plant status would then rely l
primarily on CETs. Explain why use of other radiation monitors (e.g., in the steam line or stack) is not suggested to provide confirmation of core damage.
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l P4of10 It is not clear whether agreement within 50% can realistically be achieved. This should be evaluated through some type of validation.
P6of10,82 The equation in Section B2.b estimates the extent of over-temperature based on the number of CETs exceeding CET3. However, CET3 is used to denote clad rupture from pressure effects. Explain why a temperature value associated with over-temperature damage (e.g., CET2) is not recommended instead.
F' P4of27 The recommended value for CET2 (2000F) seems too high to represent a stable condition. If you reach this value, PCT and pellet temp would be several hundred degrees higher and CETs would rapidly increase beyond 2000F.
(Discussions in Section 2.5 of WCAP-14986 suggest that a PCT above 1800F would result in significant hydrogen generation and core melting.) Explain why the value selected for CET2 is not one that can be sustained without proceeding to core melt. Should address this through a validation activity.
P5of27 The recommended value for CET3 appears inconsistent with the values discussed on P37. Should address this through a validation activity.
P6of27 Use of a value for CET4 based on conservative analyses, as recommended, will j
tend to over-predict the extent of clad damage. This may or may not be important depending on the particular sequence. Explain why this value is not
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based on best estimais analyses, similar to approach for determining CET3.
l P8of27 The recommended value for RCP2 appears inconsistent with the values discussed on P37. Should address this through a validation activity.
- P10of27 Explain how sensitive the recommended RTD temperature is to the specific scenario. Provide an estimate of RTD temperature that would be seen at the
!!me the CETs reach 1200F in several reprasentative sequences, e.g., (1) LOCA in hot leg and in cold leg, (2) transient, (3) SGTR.
P12of27 Describe why RCS pressure isn't a consideration in using reactor vessel level instrumentation for confirmation, since it would affect swol!sn level. Confirm what pressure was assumed in determining the recommended value, and whether this value bounds the RCS pressure effect.
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P90f24 Bullets 3 and 4 indicate that the core should be partially uncovered if clad l
damage has occurred. This is true if core damage is occurring at that moment, l
but the core may have uncovered previously, and then been recovered. The l
discussion should indicate that the reactor vessel level and SRM histories need l
to be considered rather than the instantaneous values. (Same comment on P18 l
of 24).
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Editorial Comments P2, para 2 Change " Criteria ll.B.3" to " Item II.B.3" P37, para 3 Change "if the RCS temperature is less than 1050 psig" to "if the RCS pressure is less than 1050 psig."
P44 Figures 5 and 6 should be replotted on a scale that would permit interpolation P46, Table 7 The words " fuel rod over-temperature" and " fuel over-temperature" seem to be used interchangeably, as are the words " core over-temperature" and " fuel over-temperature". Please clarify in the document whether these words are intended to be synonymous.
P7of24, para 6 The words CET4 and CET3 are switched in the second sentence.
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