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AUG 151978 6L A MEMORANDUM FOR:
0.B. Vassallo, Assistant Director for Light Water
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FROM:
D.F. Ross, JP., Assistant Director for Rea.c. tor Safety, DSS
SUBJECT:
MIDLAND STAFF POSITIONS AND SECOND ROUND QUESTIONS Plant Name:
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Plant. Units 1 and 2 n-
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Milestone Number:
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Licensing Stage:
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Responsible Branch
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and Project Mana9er:
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Systems Safety Branch Involvedr'~CoM Ferfbnsrn'cM~1rranch
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Description of Review:
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Requested Completion Date: ~~~ August 471!F/8 '
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The Reactor Fuels and ReactoF7hysics Sections of the Core Performance ~'
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Branch have prepared the attached SthTFF6sitTon's ahdMowf Rbeiid'QudsTions on Midland.
We have not receTV'~d"fesfolisWt3"aTT'our ffYst' founa YueTs'-~~ '~ "" -- ~~
e questions.
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Denwood F. Ross, Jr.. Assistant Director
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for Reactor Safety
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SECOND ROUND QUESTIONS FOR MIDLAND REACTOR FUELS SECTION 231.24 a.
From the response to Question 231.2 concerning stress and strain (4.2.1.1) limits, it is not yet clear how the 0.4% elastic strain limit is derived and how it is related to Condition I and II (or III) events.
For example, how is the 0.4% elastic strain limit met in a Condition II event such as " Control Rod Withdrawal"?
b.
The stress and strain limits provided in FSAR section 4.2.t.l.2 appear to be intended for the fuel system in general. What stress limits are proposed for individual components of the fu'el system such as spacer grids, guide tubes, fuel rod cladding, control rods, and other fuel system structural members for Condition I, II, III, and IV events?
231.25 a.
Please show, by means of an exemplar calculation, how combinations (4.2.3.1) of system operating transients are evaluated to ensure that the
" cumulative usage factor is less than 0.9 of the allowable material fatigue life. " Also show, by use of examples (as originally requested in Q 231.3), under what circumstances the O'Donnell and Larger fatigue curve is reduced by a factor of two on stress, rather than 20 on the number of cycles, to derive the design fatigue curve.
b.
The discussion of fretting in FSAR subsection 4.2.3.1.4 alludes to some testing performed in the B&W control rod drive facility and also to some post irradiation examination of fuel assemblies.
In neither case is a report number or other reference provided in support of the rather brief FSAR discussion.
Please provide the desired references.
. 231.25 c.
Recent excessive wear problems have been encountered in some (cont'd)
B&W plants equipped with Mark B fuel assemblies with burnable poison rod assembly (BPRA) and orifice rod assembly " ball-lock" coupling devices.
Please discuss what remedial design changes will be made to preclude similar problems in Midland Units 1 & 2.
Please state the allowable fretting wear and indicate the relationship to the cumulative usage factor.
231.26 The response to Q 231.4 concerning (a) the specification for (4.2.1.1) dryness of the fuel pellets, (b) the statistical sampling technique, and (c) the method of moisture detection is in-adequate. A numerical value for the moisture limit and its relationsh'p to the operating experience referred to in the response to Q 231.4 (FSAR section 4.2.1.1.4) is needed.
Provide typical values for the number of pellets per rod analyzed. Also discuss the method used for moisture detection.
231.27 The last sentence of FSAR section 4.2.1.1.1 " Mechanical Properties" (4.2.1.1) seems to imply that Zircaloy cladding property values provided in Chapter 15 for transient analysis are not the same as those provided in FSAR Chapter 4.
The rationale for such differences, if they exist, should be provided.
It would also be helpful to see a tabulation of the UO2 and Zircaloy properties, along with a list of data sources supporting these values (as requested in Q 231.5).
. 231.28 Please provide the chlorine and fluorine limits used in fuel (4.2.1.3) rod fabrication and compare these to the inpurity levels known to affect cladding performance.
231.29 The stress on the holddown spring (102,800 psi) provided in Table 4.2.7 of FSAR Rev. 9 is close to the stress intensity limit (107,000 psi).
Please show how these values were cal-culated and disucss why the rather small margin is considered adequate for the design.
231.30 The response to question 231.20, concerning pellet / cladding (4.2.3.1) mechanical interaction, implies that cladding mechanical inter-action failures are simply a function of residual ductility (see FSAR subsection 4.2.3.3.1).
If the cladding is in perfect condition, having absolutely no defects, then this assertion is essentially true. After the fuel has undergone significant burnup, however, it has defects. Moreover, unless exactly the same stress conditions and environment exist during a laboratory test as exist in the fuel element in the reactor,-the test results are not directly applicable, but must be subject to interpreta-tion. Therefore, the FSAR statement that "recent irradiated cladding ductility data... indicate that current pruduction cladding operating in the temperature range typical of a PWR (650F or greater) retains a ductility during irradiation in excess of that which would lead to a PCI concern" is inaccurate and misleading and should be modified or eliminated.
.. - - - -. - 231.30 First round question 231.20 has also not been answered satis-(cont'd) factorily in terms of the request for identification of the operating conditions that would lead to a PCI failure concern (even if the referenced test data were accepted as directly applicable).
We are studying pellet / cladding interaction generically. There is evidence that PCI is a concern not only during normal operation but for transients and accidents as well, whereas the discussion of pellet / cladding interaction (PCI) of the kind initiated by stress corrosion cracking, provided in FSAR subsection 4.2.3.1.3, addresses normal operation experience only--no discussion is pro-vided concerning potential PCI under transient ard accident con-ditions, including ATWS.
Particularly for transient overpower or reactivity insertion events, where the fuel pellets may over-heat and expand against the cladding, the potential exists for fuel failure due to PCI.
Because our review of the consequences of PCI failures has so far net resulted in the identification of specific safety problems, we have not imposed any operating restrictions.
If any safety issues are identified in the future, however, appropriate restrictions will be implemented.
231.31 The response to 1st-round question 231.21 in FSAR subsection 4.2.3.4.2 (4.2.3.4) is very general in nature and lacks detail.
For example, although a brief reference is made to analyses of frictional forces between fuel rods and. spacer grids, no detail is provided concerning the magnitude of these forces (as compared to hydraulic forces) or the
231.31 limits to irradiation growth of fuel rods and spacer grids.
(cont'd)
Nor is any detail provided regarding the post-irradiation examinations that are asserted to support the conclusion that the frictional forces are adequate throughout life.
Please provide enough detail to establish a credible case.
231.32 The response to ist round question 231.23 in FSAR subsection (4.2.3.5) 4.2.3.5.3 should have included some numerical values for measured fuel handling and shipping loads for comparison with design loads.
Please provide the measured and design values so that an assessment of margin can be made.
231.33 Although the FSAR contains a fairly detailed description of the (4.2.1.3) models in the TAC 0 fuel performance code, the FSAR is not clear about what TAC 0 is to be used for. The FSAR should clearly indicate that TACO will be used for all safety analyses (including ECCS analysis) related to reactor licensing; previously used codes like TAFY should not be used.
SECOND ROUND QUESTIONS FOR MIDLAND REACTOR PHYSICS SECTION 232.
Reactor Physics Section 232.12 Your response to Question 232.2 is inadequate. Topical report BAW-10028 indicates that certain misloadings in equilibrium cycles could lead to violation of fuel limits (DNBR).
Does this conclusion apply to the Midland first cycles and, if so, could such misloadings be detected by the incore instrumentation? Provide a quantitat'ive discussion including the effect of the misloading on the misloaded assembly and on neighboring assemblies containing incore detectors.
Provide conclusions regarding the probability of detection of assembly loading errors and, if appropriate, discuss the consequences of operating with an undetected misloading.
232.13 Topical Report BAW-10122 " Normal Operating Controls" has not been received for review.
Please delete the reference to this report and present a short summary of it or provide the report in time for a sufficient review to permit the licensing of Midland.