ML19275H341
| ML19275H341 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 03/01/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555, FOIA-81-236 NUDOCS 8006240705 | |
| Download: ML19275H341 (7) | |
Text
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Midland FS 1st-Round Fuels Guestions
( FSAR Section 4.2) 231.1 The design bases listed in the Midland FSAR are correlated (4.2.1) with plant conditions such as normal operation (Condition I),
upsets (Condition II), emergencies (Condition III) and faulted conditions ( Condition IV). This categorization differs from that used in B-SAR-205.
'n the latter case the design bases are separated into only three categories in tenns of fuel assembly loading conditions, one of which, viz. shipping and handling conditions, is not addressed in the Midland grouping.
Please explain the ra'tionale for the change in design bases categorization and the de'etion of shipping and handling considerations.
231.2 The FSAR listing of stress and strain limits concludes with the (4.1.1.1) statement that "those limits are consistant with current practice." This section should also contain a brief statement regarding the origin, evolutionary history, and rationale for each limit.
The limits should also be correlated with the design bases (Condition I thru IV events where applicable. For example, the relationship (if any) of the 1% inelastic + 0.4%
elastic strain limit to Condition I, II and III events should be indicated; e.g., how do these limits preclude fuel failure during Condition II transients or limit failure to a small calculable fraction of rods during Conditon III accidents?
231.3 The " cumulative fatigue damage factor" mentioned in the discussion (4.2.1.1) of " Vibration and Fatigue." FSAR section 4.2.1.1.3 should be defined. Please explain also the rationale for limiting the cumd ai.ive fatigue damage factor to 90; of the allowable material fatigue life, i.e., why not some other fraction of fatigue life--
for example, 80%? Why are all Condition I and I! events to be encompassed by this rule, but only one Condition III event?
Please show, by use of examples, when the O'Donnell-and-Langer curve is modified by a factor of two on stress amplitude and under what conditions a factor of 20 on the number of cycles is, instead, used to get a properly conservative design basis.
Discuss the design features used to ensure that flow-induced vibrations do not lead to excessive fuei rod and guide tube fretting, and refere.qce the test data and operational experience that supports this conclusion.
231.4 Please discuss the specifications for dryness of the pellets and (4.2.1.1) cladding as a Zircaloy hydriding preventive measure. Di scuss the statistical sampling technique and the method of moisture detection used to ensure that the moisture has been removed.
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231.5 Tablulate the U0 and Zircaloy thermal-physical properties 9
(4.2.1.2) used in the fuel ~ design analysis and reference the data supporting these values.
231.6 The discussion of UO chemical properties should list the 3
(4.2.1.2) major impurities known to impair the performance of the fuel or cladding, the level of impurity known to cause a problem, and the impurity limits used in fabrication.
231.7 The CROV code used in the creep collapse analysis was re-(4.2.1.3) ceived and accepted for use in safety analysis related to licensing subject to the following conditions:
- 1) The creep-related material properties used in the analysis should be similar to those characteristic of current B&W cladding.
- 2) The initial ovality input to CROV should both bound the as-fabricatedd cladding and be not less than 0.0005 inches.
(Ovality = OD max. - 00 min)
- 3) The result", of long-term, inreactor confirmatory tests will continue to be favorable.
Please indicate how these conditions were met in the use of the CROV code to analyze creep collapse in Midland fuel.
231.8 In our review of the "TACC" thermal analysis and fuel performance
( '. 2.1. 3 )
code, we determined that several modifications to the code were needed, including a revisicn of the fission gas release model to account for enharced release at high exposure.
The Midland FSAR does not reference the approved. version of TAC 0, viz.
BAW-10087A, Rev. 1, August 1977 Please reference and use the approved version of TAC 0 for the Midland Fuel Thermal Analysis and modify the FSAR as recuired.
231.9 The FSAR discussion of fuel assembly sturctural design contains (4.2.1.5) the following statement:
The Fuel Assembly is designed to ensure safe operation for Condition I and II events."
One can interpret this statement to have, by implication, one or more of the following meanings:
a)
Condition III and IV events need not be handled safely, b)
Condition III and IV events need not be considered in the design analysis.
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3 c)
Condition I and II events constitute a definition of the term " operation," but III and.IV do not.
Please indicate which, if any, of these interpretations is correct and explain the rationale for the interpretation.
Show how the stress intensity limits used for stainless steel and zirconium - based alloys bound the calculated maximum expected loadings for Condition I thrL IV.
Li st, in tabular form, numerical values for these limits and loadings and show how the loadings were calculated.
231.10 Please list the numerical values and equations, along with (4.2.1.6) their reference sources, for the pertinent thermal-physical properties used in the design of the Ag-In-Cd and A10 -B C 23 4 absorber materiais. A minimum list of pertinent properties should include melting point, swelling, thermal conductivity, thermal expansion and gas release. Also list the calculated expected values for end-of-life swelling and gas release ai.d compare these to the naximum allowable design values under normal and off-normal conditions; i.e., Conditons I thru IV.
231.11 Discuss the fabrication specificatior.: of the A1 0 -B C 23 4 (4.2.1.6) lumped burnable poison pellets; in particular discuss the specifications to ensure that residual moisture levels are below those which could result in hydriding and parforation of the Zircaloy-4 claddinq.
If the poison rod cladding were perforated, the B C would react with primary coolant water 4
to form H 803, which would then be leached into the coolant.
3 Please discuss the potential safety implications of the reactivity insertion resulting from the loss of B-10 from the burnable poison rods via this mechanism. Would resulting power changes be detected?
231.12 List and discuss the testing performed to show that the absorber (4.2.1.6) materials are compatible with their respective cladding materials i.e., Ag-In-Cd alloy with stainless steel and Al 0 -0 C with 23 4 Zircaloy-4.
Show the respective rate equations and the amount of attack over the design life.
231.13 Please demonstrate that the use of minimum unirradiated strength (4.2.1.6) values for the control rod and burnable poison rod cladding alloys is conservative under all postulated reactor conditions; e.g.
demonstrate that the increased strength due to irradiation is not af fected by a decrease in ductility.
Please discuss the bases for the 1% and 3% strain limits for 304SS and Zircaloy-4 cladding, respectively.
Show how these limits are consistent with analytical and test results, as stated in FSAR section 4.2.1.6.3.
Pl ease list and briefly describe the control component examinations mentioned in FSAR section 4.2.1.6.4.
_4, 231.14 The discussion of fuel surveillance, which includes a post-(4.2.1.7) irradiation examination (PIE) program, requires considerable ampl i fication.
Please provide a listing of the PIE tests' performed and in progress, discuss the results to date and anticipated, and show how the results provide verification of the adequacy of the fuel design.
231.15 The discussion of flow-induced vibration and fretting requires (4.2.3.1) extensive amplication in regard to the treatment of the out-of-core testing said to have demonstrated that flow-induced vibration amplitudes of the fuel rods do not cause fretting for PWR operating conditions. The discussion in FSAR section 4.2.3.1.4 does not provide enough quantitative information on the fretting and wear test program on either the fuel rods or control rod guide tubes.
Please cite published references for the tests mentioned in this FSAR section. The statement, in FSAR section 4.2.3.1.1 regarding the confirmation of these results by PIE of production B&W fuel also requires support with specific documented examples.
'31.16 Please provide numerical values for fuel rod stresses caused (4.2.3.1) by ( a) pressure differential (b) ovality bending, (c) the rmal,
and (d) grid loads for the worst case Condition I thru IV events.
Provide numerical evidence to support the assertion that differ-ential fuel rod growth and flow-induced vibration stresses do not affect these worst case stresses.
231.17 Please provide the coolant chemistry spccifications said to (4.2.3.1) coatrol the oxidation of the fuel rod cladding and provide data to support this assertion. What provisions are made to control cladding oxidation and crud deposition? Please provide further support, in the fonn of referable data for the Statement that "the majority of the SCC ( strass corrossion cracking) experience has been reported for conditions not repre-sentative of B&W operating conditions of current design."
231.18 The rod bowing correlation presented in FSAR section 4.2.3.1.8 (4.2.3.1) has not been approved by URC.
The currently acceptable correlation for thermal hydriding calculations for BSW Mark 815x15 fuel rods is as follows:
( AC/ C)95 = 0.065 + 0.00145 M where BU is mwd / tu and ( AC/ C)95 is a hot, fractional, 95/95 closure.
These coefficients meet the statistical requirements necessary for use in directly assessing DNBR penalties.
Pl ease correct the cited FSAR section and re-do the rod bowing analysis using the corrected correlation.
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231.19 Show the relationship betueen.he fuel rod cladding swelling (4.2.3.1) design curve and rod growth data refered to in FSAR section
- 4. 2. 3. 1. 10.
List the data sources.
231.20 Please reference or provide the "recent irradiated cladding (4.2.3.3) ductility data" asserted to indicate that " current production cladding under typical operating conditions retains a ductility with irradiation in excess of that which would lead to a PCI concern."
What "aty7ical" or off-normal operating condition would be expected to cause a PCI problem, based on these ductility data and analyses ( also see Q231.16).
231.21 Please discuss the test data and analyses which support the (4.2.3.4) assertion that frictional contact between the spacer grids (4.2.3.4) and fuel rods is adequate to maintain rod position while not leading to undesirable forces due to differential fuel rod growth.
231.22 Please provide the dimensions and spring constants for the upper and lower plenum springs and show quantitatively that the resistance to creep and relaxation of the spring alloy is suf ficient to withstand the worst postulated flux, temper-ature, and stress conditions.
231.23 Please describe the extent to which the fuel handling and shipping design loads have been confirmed experimentally.
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Reactor *.ysics Section
'12.0 Reactor Physics
- 12. 1 The topical report, BAW-10123, referenced on this page (P 4.3-9) has not been submitted.
Please provide this report for review or provide a summary of its contents on t'!.e Midland docket.
z32.2 The incore instrumentation is cat able of detecting gross (P 4.3-12) distortions in radial power distributions.
However, it may not be capable of detecting localized perturbations (e.g., interchange of Batch 1 and 2 assemblies near core-center). Show that any fuel loading error that is not detectable with incore instrumentation produces pertur-bations which do not violate thermal limits when operating at 102% FP.
232.3 The topical report, BAW-10120, referenced on this page (P 4.3-19) has not been submitted for review.
Please provide this report or provide a summary of its contents on the Midland docket.
212.4 Has a compiere analysis been performed to identify all 15.0-2) maloperations or failures 1n the ICS or ICS control
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functions which might produce more serious consequences in transients?
'32.5 Have safety-related systems also been analyzed witu e 15.0-2) regard to failure of passive components?
232.6 Comment on the consequences of the Startup Accident as
- P 15.4-2) a function of initial core reactivity and indicate the reason for the choice of 1% ak/k subcritical.
232.7 In view of the f act that scram occurs at powers signifi-(15.4.1.3) cantly less than full power when a pressure trip occurs and the delay time of 0.7 sec. for the pressute trip is at the extreme of the sensitivity analysis, please comment on the suitability of the full power scram insertion curve for this transient.
What effects compensate for the fact that, at.ow power, the f ractional Initial reactivity insertion is lower than that shown in Figure 15.0-3 I
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- 32.8 Explain the statement (P 15.4-B) that the positive (15.4.3.1 reactivity increase due to single rod withdrawal will cause the inlet temperature to increase in view of the fact that the ICS acts to reduce inlet temperature with increasing power above 15% FP.
232.9 What is the significance cif the location of the point (rig. 15.4-20) labelled ICS compensation on this figure?
In particular why is it plotted at an initial power of 105%?
,12.10 Please give a short description of the techniques (15.4.8.4) used, assumptions made, and results obtained which support the conclusion that no serious core damage or additional loss of :oolant system integrity results f rom the rod ejection accident.
232.11 Describe the features of the CREM and reactor coolant (15.4.8.2) boundary design that prohibit or render very unlikely the ejection of a second control rod as a result of the first ejection.
r?.12 The full power rod ejection accident is performed for
.2.4.8.2) an assumed rod worth of 0.65%.
What is the expected worth of a rod which might be ejected at full power.
What scenario is employed to produce a rod of 0.65%
wcrth? What would be the worth of a second rod which might be ejectedi 6e *
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