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DISTRIBUTION:
DOCKET FILES: 50-424/425 00T 12 isN.8:r/f CP L RUBENSTEIN L. PHILLIPS R. LOBEL D. FIENO MEMORANDlM FOR:
T. M. Novak, Assistant Director H. RICHINGS for Licensing, DL H. BALUKJIAN M. DUNENFELD FROM:
.L. S. Rubenstein, Assistant Director for fore and Plant Systems DSI
SUBJECT:
V0GTLE UNITS 1 AND 2 DRAFT SAFETY EVALUATION REPORT Plant Name:
Vogtle Units 1 and 2 Docket Number:
50-424/425 Licensing Stage:
Operating License Responsible Branch:
Licensing Branch No. 4 Project Manager:
M. Miller (X-24259)
DSI Review Branch:
Core Perfomance Branch Review Status:
Six Confirmatory Issues Nine Open Issues The Core Performance Branch has prepared the enclosed DSER input for Sections 4.2, 4.3, 4.4,15.4.1,15.4.2,15.4.3,15.4.7 and 15.4.8 of the Vogtle Units 1 and 2 FSAR covering up to amendment 9 of the FSAR.
The confirmatory and open issues are identified as follows:
Confirmatory Issues:
1.
Confirmation of the correct references for the cladding rupture and cladding ballooning and flow blockage models for the large break LOCA (see Paragraphs 4.2.3.2(6) and 4.2.3.3(3) in this report).
A ceramitwent to use the on-line fuel failure detection methods (see 2.
Paragraph 4.2.4.2 in this report).
3.
Confirmation that the rod bowing analysis has been perfomed with the latest approved model (see Paragraph 4.2.3.1(6) in this report).
I 4.
Confirmation of the correct value for the peak linear power for nomal operation (in the FSAR, Table 4.1-1 shows 12.6 kW/ft but Table 4.4-1 shows 12.5 kW/ft).
l Contacts:
H. Richings H. Balukjian M. Dunenfeld X-29418 X-29422 X-28097 1
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Confirmation that the analysis of the dropped control rod event meets DN8 limits (see Section 15.4.3, second paragraph).
Open Issues 1.
A statement about the sensitivity of the chemical volume and control system letdown monitor for detecting fuel rod failures (see Paragraph 4.2.4.2 in this report).
2.
A commitment in the postirradiation fuel surveillance program to perform additional surveillance if unusual behavior is noted in the visual examination or if plant instrumentation indicates gross fuel failures (see Paragraph 4.2.4.3 in this report).
3.
A stataaent in the postirradiation fuel sruveillance program that addresses the disposition of failed fuel (see Paragraph 4.2.4.3 in thisreport).
4.
Supply information regarding flow measurement capability with crud build-up (see Section 4.4.4.2).
5.
Mdress the concern on ther.'.al-hydraulic design compariso' '(see Section n
4.4.5).
6.
Moress the concern regarding N-1 loop operation (see Section 4.4.6).
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7.
Mdress the concerns regarding the Loose Parts Monitoring System (see Section 4.4.7).
i 8.
Supply the information for Item II.F.2 of NUREG-0737 (see Section 4.4.8).
i 9.
Mdress the concern regarding DNBR for a steam line break (see Section 4.4.9).
Original s!gned by 1
L. S. Rubenstein L. S. Ilubenstein Assistant Director for Core and Plant Systems, DSI
Enclosure:
As stated cc:
R. Bernero D. Eisenhut E. Mensam M. Miller
- M. Wigdor R. Capa
- SEE PREVIOUS CONCURRENCE ***
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5 V0GTLE-1 AND -2 DRAFT SAFETY EVALUATION' REPORT 4.2 Fuel Design f
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The fuel assembly described in the FSAR for Vogtle-1 and -2 is a 17 x 17 array of fuel rods having a diameter of 0.374 in. This design will be referred to as the Standard Fuel Assembly (SFA) in the following paragraphs.
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Section 4.2 of the FSAR presents the design bases for the SFA.
For the
,I Westinghouse 2 ) analysis, plant design conditions are divided into four categories of operation that are consistent with traditional industry classi-fication (ANSI Standards N18.2-1973 and N-1212-1974): Condition 1 is Normal i
Operation, Condition 2 is Incidents of Moderate Frequency, Condition 3 is Infrequent Incidents, and Condition 4 is Limiting Faults.
Fuel damage is then related to these conditions of operation, which are coupled to the fuel design bases and design limits. The subsections of the design bases section address topics such as (a) cladding, (b) fuel material, (c) fuel rod performance.
(d) spacer grids, (e) fuel assembly structural design, (f) in-core control components, and (g) surveillance program. Thus, as part of the discussion of l
the cladding design bases, cladding mechanical properties, stress-strain limits, vibration and fatigue, and cladding chemical properties are also pre-sented. A similar approach is taken for the other major subtopics.
l The review and safety evaluation will follow SRP Section 4.2 (NUREG-0800,
[,
Rev. 2). The objectives of this fuel system safety review are to provide assurance that (a) the fuel system is not damaged as a result of normal opera-f tion and anticipated operational occurrences, (b) fuel system damage is never so severe as to prevent control rod insertion when it is required, (c) the number of fuel rod failures is not underestimated for postulated accidents, and U
(d)coolabilityisalwaysmaintained.
"Not damaged" is defined as meaning that
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f fuel-rods do not fail, that fuel system dimensions remain within operational 09/21/84 4
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O tolerances, and that functional capabilities are not reduced below those assumed in the safety analysis. This objective implements General Design Cri-
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terion (GDC) 10 of 10 CFR Part 50, Appendix A (" General Design Criteria for y
Nuclear Power Plants") and the design limits that accomplish this are called Specified Acceptable Fuel Design Limits (SAFDLs).
" Fuel rod failure" means e
that the fuel rod leaks and that the first fission product barrier (the clad-ding)has,therefore,beynbreached. Fuel rod failures must be accounted for in the dose analysis required by 10 CFR Part 100 (" Reactor Site Criteria") for postulated accidents.
"Coolability," which is sometimes termed "coolable i
j geometry," means, in general, that the fuel assembly retains its rod-bundle 1
geometrical configuration with adequate coolant channels to permit removal of residual heat after a severe accident. The general requirements to maintain l
control' rod insertability and core coolability appear repeatedly in the General Design Criteria (GDC 27 and 35). Specific coolability requirements for the h
l loss-of-coolant accidents are given in 10 CFR Part 50.46 (" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors").
I h
To meet the above-stated objectives of the fuel system review, the following f
specific areas are critically examined:
(a) design bases, (b) description and designdrawings,(c)designevaluation,and(d) testing, inspection,andsur-veillance plans.
In assessing the adequacy of the design, several items I
involving operating experience, prototype testing, and analytical predictions are weighed in terms of specific acceptance criteria for fuel system damage, fuel rod failure, and fuel coolability. Recently, Westinghouse developed the Optimized Fuel Assembly (0FA), which is described in WCAP-9500. WCAP-9500 is mentioned in the last paragraph (p. 4.2-2) of Section 4.2 of the FSAR for Vogtle-1 and -2 but is not included in the reference list for that section.
WCAP-9500 was approved by the NRC (Rubenstein, May 15, 1981, and Tedesco, May 22, 1981). The OFA design also consists of a 17 x 17 array of fuel rods but with the rods having a diameter of 0.360 in., which is somewhat smaller than the rod diameter in the SFA. Because the fonnat of WCAP-9500 followed Regulatory Guide 1.'/0, some of the fuel design bases and design limits for the OFA W se not presented in WCAP-9500 in a form that pennitted cross-checking by the NRC with the acceptance criteria provided in Section 4.2 of the SRP.
Therefore, several questions were issued (Rubenstein, August 8, 1980) to 09/21/84 5
V0GTLE 1&2 SER INPUT PR.
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i clarify the design bases and limits. Responses to those questions are con-f tained in letters from Westinghouse (Anderson, January 12, 1981, and April 21, i
1981). These responses are applicable to the Standard Fuel Assembly to be used i
in Vogtle-1 and -2 as well (Petrick, September 9,1981)'. References to these l
questions and answers will be made at several places in the review that f
follows.
4.2.1 Design Bases Design bases for the safety analysis address fuel system damage mechanisms and y
suggest limiting values for important parameters such that damage will be limited to acceptable levels. For convenience, acceptance criteria for these design limits are grouped into three categories in the Standard Review Plan:
l (a) fuel system damage criteria, which are most applicable to nomal operation 2 plant Condition 1), including anticipated operational occurrences 2 plant
.QoJndition 2), (b) fuel rod failure criteria, which apply to normal operation 2 plant Condition 1), anticipated operational occurrences 3 plant Condition 2),
i and postulated accidents 2 plant Conditions 3 and 4), and (c) fuel coolability criteria, which apply to postulated accidents 2 plant Conditions 3 and 4).
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4.2.1.1 Fuel System Damage Criteria The following paragraphs discuss the evaluation of the design bases and corresponding design limits for the damage mechanisms listed in the SRP.
These design limits along with certain criteria.that define failure (see Sec-tion 4.2.1.2 of this report) constitute the SAFDLs required by GDC 10. The design limits in this section should not be exceeded during normal operation including anticipated operational occurrences.
(1) Cladding Design Stress In Section 4.2.1.1 of the {SAR, it is indicated that the cladding stresses under Conditions 1 and 2 are less than the Iircaloy yield stress, with due consideration of temperature and irradiation effects. The design basis for fuel rod cladding stress as given in the response (Anderson, January 12, 1981, 09/21/84 6
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to excessive fuel rod cladding stresses. The design limit for fuel rod clad-ding stress under Condition 1 and 2 modes of operation is that the volume-y averaged effective stress calculated with the von Mises' equation, considering j
interference due to uniform cylindrical pellet-to-cladding contact (caused by N
pellet themal' expansion and swelling, unifom cladding creep, and fuel rod /
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coolant system pressure pifferences), is less than the Zircaloy 0.2 percent j
offset yield stress. as affected by temperature and irradiation. This is a j
traditional limit consistent with previous Westinghouse design practice, but
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with credit being taken by Westinghouse for irradiation-induced strengthen-ing.
I Wehaveapproved(Rubenstein, June 6,1983)WCAP-9179, Revision 1,which l
includes approval for taking such credit.
r (2) Cladding Design Strain E
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With regard to cladding strain, a design limit for fuel rod cladding plastic h
tensile creep (due to uniform cladding creep and uniform cylindrical fuel pellet swelling and themal expansion) of less than 1 percent from the unir-
,d radiated condition is given in the response (Anderson, January 12, 1981, and April 21, 1981) to Q231.2 and in Section 4.2.1.1 of the FSAR. Furthermore, the f
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total tensile strain transient limit (due to uniform cylindrical pellet thermal i
expansion during the transient) is stated to be less than 1 percent from the pretransient value. The applicant indicates in Section 4.2.1.1 of the FSAR i
that this limit is consistent with proven practice (WCAP-8183). While the supporting data for nomal operation (Condition 1) has not been explicitly reviewed, that value appears to be consistent with past practice (no numerical value for normal operation cladding strain is provided as an acceptance cri-terion in the Standard Review Plan), and thus there is reasonable assurance that 1 percent total plastic creep strain is an acceptable design limit for nomal operation, including Condition 1 power changes (load following). For transient-induced defomation, the Standard Review Plan indicates that 1 per-cent uniform cladding strain is an acceptable damage limit that should preclude
- All questions and responses referred to in this manner were part of the review of WCAP-9500, and the first application of the SFA, on the Shearon Harris docket. References to the FSAR refer to the Vogtle-1 and -2 FSAR.
09/21/84 7
V0GTLE 1&2 SER INPUT l
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c some types of pellet / cladding interaction (PCI) failures. Such a limit, hew-ever, while consistent with past practice, should not be construed to be a broadly applicable PCI damage limit because there is ample evidence (Tokar, i
November 14,1979) that PCI failures can' occur at less than 1 percent eniform cladding strain.. Westinghouse has indicated in its response (Anderson, l
January 12, 1981, and April 21,1981)toQ231.24that1percentplasticstrain j
fromthepretransientva}ueisnotmeanttoserveasabroadlyapplicablePCI l
criterion. Nevertheless, the 1 percent cladding transient plastic strain cri-terion appears to be an acceptable design limit for the type of application j
indicated in SRP Section 4.2.
For fuel assembly structural design, Westing-l house set design limits on stresses and deformations due to various nonopera-
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tional, operational, and accident loads. As indicated in Section 4.2.1.5 of
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the FSAR, the stress categories and strength theory presented in Section III of
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the ASME Code are used as a general guide. This is consistent with acceptance criterionII.A.1(a)ofSRPSection4.2andisacceptable.
1 (3) Strain Fatigue According to Section 4.2.1.1(C,1) of the FSAR, the cumulative strain fatigue
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cycles are less than the design strain fatigue life, which is consistent with provenpractice(WCAP-8183). The strain fatigue criteria given in the-response (Anderson, January 12, 1981, and April 21,1981)toQ231.2andin Section 4.2.3.3.1 of the FSAR are the same as those described in SRP Section 4.2, viz., a safety factor of 2 on stress amplitude or of 20 on the number of cycles and are, therefore, acceptable.
(4) Fretting Wear h
While the Standard Reyiew Plan does not provide numerical bounding-value f
acceptance criteria for fretting wear, it does stipulate that the allowable
[l fretting wear should be stated in the safety analysis' report and that the stress and fatigue limits ghould presume the existence of this wear.
i InSections4.2.1.1,C,2and4.4d.7oftheFSAR,itisindicatedthatpotential fretting wear due to vibration is prevented, assuring that the stress-strain I
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V0GTLE 182 SER INPUT
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limits are not exceeded during the design life. From the response (Anderson,
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January 12, 1981, and April 21, 1981) to Q231.5 it can also be seen that the j
Westinghouse design basis for fretting wear is that fuel rods shall not fail ly during Condition 1 and 2 events. Furthermore, Westinghouse does not use an j
explicit fretting wear limit in their stress and fatigue analysis for fuel I-rods. However, Westinghouse does use a value (proprietary) of wall thickness
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as a general guide in evgluating cladding imperfections, including fretting
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Cladding imperfec'tions including fretting wear are thus considered in wear.
the stress and fatigue analysis, albeit in a qualitative manner.
In view of I
the apparently small effects of these defects and large stress and fatigue margins (see Section 4.2.3.1(4) of this report), this design method is i
acceptable.
The design basis for guide thimble tubes [see response (Anderson, January 12,
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1981, and April 21, 1981) to Q231.41] is that the thinning of the guide thimble I
tube walls should not result in the failure of the fuel assembly structural integrity or functionability of the guide thimble tubes. We find this to be an acceptable design basis.
With regard to a design limit for guide thimble tube wear, Westinghouse has
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determined from stress analyses that the most limiting load on the fuel assem-l bly structure is that which might occur during a fuel handling accident.
For j
the analysis of this accident, Westinghouse uses a design criterion of 6 g, as l
noted in Section 4.2.1.5 of the FSAR. This design limit is therefore used for degraded guide thimble tubes and has been previously accepted for Westinghouse fuels.
i, (5) 0xidation and Crud Buildup The SFA design basis for cladding oxidation and crud buildup is that the increase in cladding temperature due to cladding oxidation and crud buildup is notexcessive(seeSection4.2.1.2(3),below).
l Section 4.2 of the Standard Review plan identifies cladding oxidation, hydrid-ing, and crud buildup as potential fuel system damage mechanisms. Hydriding is 09/21/84 9
V0GTLE 1&2 SER INPUT W..
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discussed in Section 4.2.1.2(1), below. Because of the increased thermal f
resistance of these layers, there is an increased potential for elevated tem-perature within the fuel as well as the cladding. Because the effect of oxida-
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tion and crud layers on fuel and cladding temperature is a function of several h
different parameters (such as heat flux and themal-hydraulic boundary condi-
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tions), a design limit on oxide or crud layer thickness does not, per se, pre-v clude fuel damage as a result of these layers. Rather, it is necessary that h,
these layers be appropriately considered in other temperature-related fuel sys-tem damage and failure analyses. Thisapproach(e.g.,seeSections4.4.2.9.1, 4.4.2.11, and 4.4.2.11.5 of the FSAR) taken by Westinghouse in the design of l
the Standard Fuel Assembly is found by the NRC staff to be acceptable.
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(6) Rod Bowing
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Fuel rod bowing is a phenomenon that alters the pitch dimensions between 4
11
- y adjacent fuel rods. Bowing effects local nuclear power peaking and the local heat transfer to the coolant. Rather than placing design limits on the amount p
of bowing that is permitted, the effects of bowing are included in the safety analysis (see Sections 4.2.3.1,0 and 4.2.3.3.5 of the FSAR). This is consis-tent with the Standard Review Plan and is acceptable. The methods used for predicting the degree of rod bowing are evaluated in Section 4.2.3.1(6) of
.this report, and the impact of the resulting bow magnitude is evaluated in Sections 4.3 and 4.4 of this report.
(7) Axial Growth In the SFA design, the core components requiring axial-dimensional analyses are the control rods, neutron source rods, burnable poison rods, fuel rods, and fuel assemblies (thimble plugging rods are omitted because they are short and notaxial-growthlimited). The axial growth of the first three of these com-ponents is primarily dependent upon the behavior of poison, source, or spacer pellets and their Type 304; stainless-steel cladding. The growth of the latter two is mainly governed by the behavior of fuel pellets, Zircaloy-4 cladding, and Zircaloy-4 guide thimble tubes.
09/21/84 10 V0GTLE 1&2 SER INPUT
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The Westinghouse design bases for core component rods are that (a) dimensional i
stability and cladding integrity are maintained during Condition 1 and 2 events l
and (b) these components do not interfere with shutdown during Condition 3 and 4 events.
7 i
Westinghouse does not, per se, have design limits on the axial growth of their j
control, source, and burgable poison rods. However, allowances are made to f
accomodate (a) pellet swelling due to gas production and (b) relative thennal expansion between the stainless-steel cladding and the encapsulated material.
Westinghouse does not account for irradiation growth of the stainless-steel claddingandhascitedexperiments(FosterandStrain, October 1974)asjusti-fication for the insignificance of irradiation growth of stainless-steel at PWR operating conditions.
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For the Zircaloy cladding and fuel assembly components, the axial-dimensional 1,
behavior is governed by creep (due to mechanical or hydraulic loading) and irradiation growth. The critical tolerances that require controlling are (a) the spacing between the fuel rods and the fuel assembly (shoulder gap) and (b)thespacingbetweenthefuelassembliesandthecoreinternals. Failure to adequately design for the former may result in fuel rod bowing, and for the latter may result in collapse of the holddown springs. With regard to inade-quately designed shoulder gaps, problems have been reported (Schenk, October 1973; Kuffer and Lutz,1973; FSAR of R. E. Ginna Unit 1,1972; Clark, July 24, 1983; Rubenstein, June 17, 1983; and Nerses, April 28,1983) in foreign (Obrigheim and Beznau) and domestic (Arkansas-2, Ginna, and St. Lucie-2) plants
(
that have necessitated predischarge modifications to fuel assemblies.
t With regard to a design basis for shoulder gap spacing, it is indicated in J
Sections 4.4.2 and 4.2.3.5.1 of the FSAR and it is stated by Westinghouse in I
the responses (Anderson, January 12, 1981, and April 21,1981)toQ231.2, Q231.8, Q231.25, and Q231.40 that interference is precluded by having clearance between the fuel rod end and the top and bottom nozzles. The design clearance accomodates the differences in growth, fabrication tolerances, and the differ-ences in thermal expansion between the fuel cladding and the thimble tubes.
r 09/21/84 11 V0GTLE 1&2 SER INPUT Q
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Westinghouse does not have specific limits on growth, but does provide a gap
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spacing that is equal to or greater than a percentage of the fuel rod length.
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With regard to fuel assembly growth, Westinghouse has a design basis that there
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shall be no axial interference between the fuel assembly and upper and lower L
core plates caused by temperature or irradiation. As a design limit, Westing-house provides a minimum, gap, which is a fraction of the fuel assembly length,
]
between the fuel assembly and the reactor internals.
t The above design bases and limits dealing with axial growth are acceptable.
(8) Fuel Rod and Nonfuel Rod Pressures For Condition 1 and 2 events, the mechanical design basis for core component l
rods described in the FSAR is that dimensional stability and cladding integrity
]
are maintained. A necessary corollary of this design basis is that the driving force, rod internal pressure, is never so great as to result in loss of dimen-sional stability and cladding integrity.
Section 4.2 of the Standard Review Plan identifies rod internal pressure as a potential fuel system damage mechanism.
In this sense, damage is defined as an increased potential for elevated temperatures within the rod as well as an increased potential for cladding failure. Although the Standard Review Plan mentions only fuel and burnable poison rods, the mechanism also applies to control rods, neutron source rods, and other core component rods. Because rod internal pressure is a driving force for, rather than a direct mechanism of, fuel system damage, it is not necessary that a damage limit be specified.
It is only necessary that the phenomenon be appropriately considered in other fuel system damage and fuel failure analyses. In other words, rod internal pressure must be considered in calculating the temperature of the rod internals, clad-ding deformation, and cladding bursting.
In order to simplify the akalysis of fuel system damage due to excessive rod i
internal pressure, the Standard Review Plan states that rod internal gas pres-sure should remain below the nominal system pressure during nonnal operation l
l 09/21/84 12 V0GTLE 1&2 SER INPUT pg.:(
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I unless otherwise justified. Westinghouse has elected to justify limits other i
than that provided in the Standard Review Plan.
For the fuel rods, revised internal rod pressure criteria as described in
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WCAP-8963,anapproved(Stolz,May 19,1978) topical report, were used in the e
FSAR. Briefly stated, these criteria (Section 4.2.1.3,8 of the FSAR) allow the fuelrodinternalpressupetoexceedthesystempressureundercertaincondi-tions:
(a) the internal pressure is limited such that the fuel-to-cladding gap does not increase during steady-state operation, and (b) extensive departure E
from nucleate boiling (DNB) propagation does not occur during normal operation and any accident event. These criteria have been previously approved and remain acceptable.
It is stated in Section 4.2.3.1,8 of the FSAR that the burnup-dependent fission gas release model in WCAP-8720, which has been approved (NUREG-0390) by the NRC, was used in the FSAR. Addendum No. I to WCAP-8720 has also been approved (NUREG-0390) by the NRC.
For the nonfuel rods, the rod internal pressure is limited such that the mechanical design limits, discussed in Section 4.2.1.5 of the FSAR, are not exceeded for Condition I and 2 events. This implies a stress limit of 2/3 of the material yield stress and a strain limit of 1 percent. These limits are unchanged from previously approved Westinghouse fuel designs and remain accept-able for this FSAR.
(9) Assembly Liftoff The Standard Review Plan calls for the fuel assembly holddown capability (gravity and springs), to exceed worst-case hydraulic loads for nonnal opera-tion, which includes anticipated operational occurrences. The SFA design basis provides for positive holddown for Condition 1, but allows momentary liftoff during one Condition 2 event (see Section 4.4.2.6.2 of the FSAR). This design basis is acceptable provided that it can be shown that the affected fuel assem-blies will reseat properly without damage and without other adverse effects 09/21/84 13 V0GTLE 182 SER INPUT
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during the event. The ability of the affected fuel assemblies to satisfy this i
provision will be discussed in Section 4.2.3.1, below.
,I (10) Control Material Leaching i
'i The Standard Review Plan.and General Design Criteria require that control rod il reactivity be maintained, Control rod reactivity can sometimes be lost by Il leaching of certain poison materials if the control rod cladding has been t
f' breached. The mechanical design basis for the control rods is stated in
).
Section 4.2.1.6 of the FSAR to be consistent with the loading conditions of i
Section III of the ASME Code. Thus, the design basis for the SFA control rods h
is to maintain cladding integrity; because cladding integrity would ensure that
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reactivity is maintained, this design basis might appear to be acceptable.
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However, under some circumstances, unexpected breaches might go undetected, so d
the NRC does not normally accept control rod cladding integrity as a sufficient h
design basis. A discussion will be presented under Section 4.2.3.1, below, d
that shows that adequate surveillance will be provided to ensure maintenance of reactivity.
4.2.1.2 Fuel Rod Failure Criteria 5
The evaluation of fuel rod failure thresholds for the failure mechanisms listed in the SRP is presented in the following paragraphs. When these failure H
thresholds are applied to normal or transient operation, they are used as j
limits (the specified acceptable fuel design limits of GDC 10), since fuel (l
failures under those conditions should not occur (according to the traditional
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conservativeinterpretationofGDC10). When these thresholds are applied to I
accident analyses, the number of fuel failures must be determined for input to.
the radiological dose calculations required by 10 CFR 100. The basis or reason for establishing these failure thresholds is thus predetermined, and only the threshold values are reviewed below.
i 09/21/84 14 V0GTLE 1&2 SER INPUT w
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h (1) Internal Hydriding h
Hydriding as a cladding failure mechanism is precluded by controlling the level
!j of moisture and other hydrogenous impurities during fabrication. As described II in the revised response (Anderson, January 12, 1981, and April 21,1981)to h
Q231.6, the moisture levels in the uranium dioxide fuel are limited by Westing-j house to less than or equal to 20 ppm. This specification is compatible with
-the ASTM specification for sintered uranium dioxide pellets, which allows two micrograms of hydrogen per gram of uranium (2 ppm). These are the same limits 3
provided in the Standard Review Plan and are therefore acceptable.
(2) Cladding Collapse If axial gaps in the fuel pellet column were to occur due to densification, the i
cladding would have the potential of collapsing into a gap (flattening).
j Because of the large local strains that would result from collapse, the clad-
]
ding is assumed to fail. As indicated in Section 4.2.1.3 of the FSAR and in responses (Anderson, January 12, 1981, and April 21,1981)toQ231.?,Q231.9, and Q231.34 it is a Westinghouse design basis that cladding collapse is pre-g cluded during the fuel rod design lifetime. This design basis is t.ie same as h
that in the Standard Review Plan and is therefore acceptable.
b f
(3) Overheating of Cladding r
The design basis as given in Section 4.4.1.1 of the FSAR for the prevention of fuel failures due to overheati g is that there will be at least 95 percent j
probability that departure from' nucleate boiling (DNB) will not occur on the limiting fuel rods during nonnal operation or any transient conditions arising from faults of moderate frequency (Condition 1 and 2 events) at a 95 percent confidence level. This design basis is consistent with the thermal margin criterion of SRP Section 4.4 and is, thus, acceptable. The specific DNBR limits and methods of analysis are reviewed in Section 4.4 of this report.
09/21/84 15 V0GTLE 1&2 SER INPUT.
Q
.w : :
a :-
=
.-m=w
..s s
e i
I l
(4) Overheating of Fuel Pellets Cj As a second method of avoiding cladding failure due to overheating, Westing-i house avoids centerline fuel pellet melting as a design basis. This design f
basis is the same as given in the Standard Review Plau and fs thus acceptable.
h I
The design limit (Sectiops 4.2.1.2 and 4.4.1.2.1 of the FSAR) corresponding to the design basis given above is that., during modes of operation associated with
[
Condition 1 and Condition 2 events, there is at least a 95 percent probability J
that the peak kW/ft fuel rod will rot exceed the UO melting temperature.
2 This design limit is an acceptable representation of the design basis given previously.
p (5) Pellet / Cladding Interaction As indicated in SRP Section 4.2, there are.w generally applicable criteria for PCI failure. However, two acceptance criteria of limited application are pre-sented in the SRP for PCI:
(a)lessthan1percenttransient-inducedcladding strain and (b) no centerline fuel melting. The response (Anderson, January 12, 1981, and April 21,1981) to Q231.2 indicates that the 1 percent cladding plas-tic strain limit is met for the SFA design, and as stated in Section 4.2.1.2 of the FSAR, the SFA design ensures that UO2 centerline melting will not occur through selection of a calculated fuel centerline temperature of 4700*F as an overpower limit. Thus the SFA design basis and limits agree with the only existing licensing criteria for PCI.
(6) Cladding Rupture In the LOCA analysis for SFA-designed plants, an empirical model is used to predict the occurrence of cladding rupture. The failure temerature is expressed as a function of differential pressure across the clading wall.
There are no specific design limits associated with cladding rupture, and the rupture model is a portion of the ECCS evaluation model, which is documented in Revision 1 of WCAP-9220-P-A and WCAP-9221-A.
a 09/21/84 16 V0GTLE 1&2 SER INPUT
~
_ = -
- - " ~ ~ ~ " - " " - " "
7r 1.'
f
...i 4.2.1.3 Fuel Coolability Criteria
[
For major accidents in which severe fuel damage might occur, core coolability I
must be maintained as required by several GDCs (e.g., GDC 27 and 35). The fol-l lowing paragraphs discuss the evaluation of limits that will assure that coola-h bility is maintained for the severe damage mechanisms listed in Section 4.2 of
(
the SRP.
1
[
(1) Fragmentation of Embrittled Cladding I
For LOCA analysis (Section 15.6.5.1 in the FSAR), Westinghouse uses the accep-1 tance criteria of 2200*F on peak cladding temperature and 17 percent on maximum i
cladding oxidation as prescribed by 10 CFR 50.46. For events other than the l
LOCA, we do not have separately established temperature or oxidation criteria.
E Yet it is clear that for short-term events such as locked rotor, the 2200*F peak cladding temperature and 17 percent oxidation LOCA criteria are not t
really meaningful, because the temperature history'for such an event is much shorter than that of a LOCA. For events such as locked rotor, therefore, Westinghouseusesauniquepeak-cladding-temperature (PCT)criterionof2700*F (e.g.,seeSection15.3.3.2andSection 15.4.8.1.2 of the FSAR).
The Westinghouse 2700*F PCT limit was selected taking into consideration the short time (a few seconds) that the fuel is calculated to be in DNB for a locked-rotor type event and the fact that the PCT and total metal-water reac-tion at the fuel hot spot would not be expected to impact fuel coolable geom-etry. While this limit has been used by Westinghouse for several years, the basis for the limit has only recently been reviewed. However, an assessment by us (Van Houten, February 23,1981) of the available experimental information indicates that fuel rod cladding will, indeed, retain its rod-like geometry after exposure to short-tenn (a few seconds) peak cladding temperature of 2700*F. That conclusion is based on four Japanese reports (Shfozawa, March 1979; Hoshi, May 1980; JAERI-M-9011, September 1980; and Fukishiro, October 1980) that describe experimental results for reactor test programs reported k
)
09/21/84 17 V0GTLE 1&2 SER INPUT
_ _ _ =
t-li
.o f
}
i.
since 1979. We, therefore, concludes that there is reasonable assurance that l
f the 2700'F PCT limit for short-term events such as locked rotor is an accep-
- l table coolability limit for the Westinghouse SFA design.
It should be noted that NRC acceptance of the 2700'F PCT limit for fuel rod coolability is currently restricted to undercooling events such as locked fl rotor. For overpower evtynts such as control rod ejection, which involve a pellet-to-cladding mechanical interaction, the NRC has not determined the p
applicability of a PCT limit and currently uses a fuel rod enthalpy criterion of 280 cal /g for coolability of a rod-ejection accident.
t, (2) Violent Expulsion of Fuel Material 2
The design bases that there should be little or no possibility of fuel dis-
[
persal in the coolant, gross lattice distortion, or severe shock waves are given in Section 15.4.8.1.2 of the FSAR and are equivalent to those in the f
Standard Review Plan.
The design limits given in the FSAR are:
[
(a) Average fuel pellet enthalpy at the hot spot will be below 225 cal /g for
[
unirradiated fuel and 200 cal /g for irradiated fuel.
1 (b) Average cladding temperature at the hot spot will be below the temperature at which cladding embrittlement may be expected (2700*F).
(c) Peak reactor coolant pressure will be less than that which could cause pressures to exceed the faulted condition stress limits.
(d) Fuel melting will be limited to less than 10 percent of the fuel volume at the hot spot even if the average fuel pellet enthalpy is below the limits in(a),above.
09/21/84 18 V0GTLE 182 SER INPUT
,l _
~_
- p..
l-e I
These limits are more conservative than the single 280 cal /g limit given in i,
Regulatory Guide 1.77, they have been previously approved in the review of
(,
WCAP-7588, and they remain acceptable.
q (3) Cladding-Ballooning and Flow Blockage I!
i
{.
In the LOCA analyses for,SFA-designed plants, empirical models are used to y
predict the degree of cladding circumferential strain and assembly flow block-f age at the time of hot-rod and hot-assembly burst. These models are each expressed as functions of differential pressure across the cladding wall.
- i l
There are no specific design limits associated with ballooning and blockage.
i and the ballooning and blockage models are portions of the ECCS evaluation h
model, which is documented in Revision 1 of WCAP-9220-P-A and WCAP-9221-A.
1:
(4) Structural Damage from External Forces t;
il
{
Section 4.2.3.5 of the FSAR states that the fuel assembly will maintain a j
geometry that is capable of being cooled under the worst case accident Condi-t f
tion 4 event and that no interference between control rods and thimble tubes will occur during a safe shutdown earthquake. This is equivalent to the design basis as presented in the Standard Review Plan and is therefore acceptable.
4.2.2 Description and Design Drawings The description of fuel system components, including fuel rods, bottom and top nozzles, guide and instrument thimbles, grid assemblies, rod cluster control assemblies, burnable poison assemblies, neutron source assemblies, and thimble plug assemblies, is contained in Section 4.2.2 of the FSAR.
In addition.
Tables 4.1-1 and 4.3-1 of the FSAR provide numerical values for various core component parameters. While each parameter listed in SRP subsection 4.2.2 is not provided in the FSAR, enough information is provided in sufficient detail to provide a reasonably accurate representation of the SFA design and this infonnation is thus acceptable. However, the number of fuel rods in Table 4.3-1 of the FSAR should be 50,952 and not 50,592 (see Table 4.1-1 in I
theFSAR).
I l
09/21/84 19 V0GTLE 1&2 SER INPUT t
- 4 h h
y
_z_,._
s _ _h
= - -
='
)
- f.
3 i-1-
I 4.2.3 Design Evaluation i
}
Design' bases and limits were presented and discussed in Section 4.2.1, above.
[
In this section, Westinghouse methods of demonstrating that the SFA fuel design l-meets the design criteria that have been established are reviewed.
[
[
Thissectionwill,therefore,correspondpointbypointtoSection4.2.1,
[
above. The methods of demonstrating that the design criteria have been met
[
include operating experience, prototype testing, and analytical predictions.
4.2.3.1 Fuel System Damage Evaluation i
The following paragraphs discuss the evaluation of the ability of the SFA fuel to meet the fuel system damage criteria described in Section 4.2.1.1, above.
i Those criteria apply only to normal operations and anticipated transients, l,
(1) Cladding Design Stress As indicated in the response (Anderson, January 12, 1981, ard April 21,1981) to Q231.2, Westinghouse used its Performance-Analysis and Design (PAD) code, WCAP-8720, to analyze cladding stress. That code has been reviewed and found t
acceptable (Stolz, February 9,1979, and Rubenstein, June 30,1982). Typical calculated design values for cladding effective stress provided in the response (Anderson, January 12, 1981, and April 21, 1981) to Q231.2 are stated to be considerably below the 0.2 percent offset yield stress design limit.
(2) Cladding Design Strain The NRC-approved Westinghouse fuel performance code (PAD) was used in the strain analysis, as indicated in the response (Anderson, January 12, 1981, and April 21, 1981) to Q231.2. Typical design values of steady-state and transient creep strain, as calculate ( by that code, are found to be below the 1 percent strain criterion. Hence, we conclude that the SFA cladding strain design
~
limits have been met.
09/21/84 20 V0GTLE 1&2 SER INPUT h.-
~~
p.
.c - -
_ _ =
_=
._7
$[.
i; i
h (3) Strain Fatigue r
As indicated in the response (Anderson, January 12, 1981, and April 21,1981) y to Q231.2, Westinghouse used their approved PAD code for the strain range and l!
strain fatigue. life usage analysis. Experimental data obtained from Westing-
[
housetestingprograms(seeSection4.2.3.3.1oftheFSAR)wereusedby f
Westinghouse to derive the Zircaloy fatigue design curve, according to the response (Anderson, January 12, 1981, and April 21,1981)toQ231.4. For a given strain range, the number of fatigue cycles is less than that required for failure, considering a minimum safety factor of 2 on stress anplitude or a p
minimum safety factor of 20 on the number of cycles, (the fatigue usage Octor islessthan1.0). The computations were performed with an approved code.
It is concluded that the SFA fatigue design basis has been met.
(4) Fretting Wear With regard to the Westinghouse fretting analysis of the fuel cladding, the conclusions of the review are the following:
(a) Cladding fretting and fuel vibration have been experimentally investi:-
l gated,asshowninWCAP-8278(andnonproprietaryversionWCAP-8279)and noted in Section 4.2.3.1,A of the FSAR. WCAP-8278 (and WCAP-8279) has been approved by us (Rubenstein, March 19,1981).
(b) The out-of-pile flow tests and analyses (WCAP-9401) to determine the mag-nitude of fretting wear that is anticipated for the OFA design have been previously reviewed and found acceptable (Rubenstein, April 23,1981).
These analyses are also acceptably conservative for SFA applications.
(c) LWR operating experience demonstrates that the number of fretting-induced fuel failures is insignificant.
i (d) There should be only a small dependence of cladding stresses on fretting wIoar because this type of wear is local at grid-contact locations and relatively shallow in depth.
09/21/84 21 V0GTLE 1&2 SER INPUT m.
ci
[
(e) The built-in conservatisms (that is, safety factors of 2 on the stress amplitudes and 20 on the number of cycles) in the strain fatigue analys'is f
as well as the calculated margin to fatigue life limit adequately offset i
the effect of fretting wear degradation.
Therefore, it is concluded that the SFA fuel rods will perfom adequately with a
j respecttofrettingwearI b
}
Fretting wear has also been observed on the inner surfaces of guide thimble tubes where the fully withdrawn control rods reside. Significant wear is limited to the relati,vely soft Zircaloy-4 guide thimble tubes because the Inconel or stainless steel control rod claddings are relatively wear-resistant. The extent of the wear is both time-dependent and plant-dependent and has, in some non-Westinghouse cases, extended completely through the guide thimble tube wall.
Westinghouse has predicted that an SFA can operate under a rod cluster control assembly (RCCA) for a period of ti:ne that exceeds the amount of rodded time expected with current 3-cycle fuel schemes before fretting wear degradation would result in exceeding the present margin to the 6-g load criterion for the fuel handling accident. However, we required several applicants to perform a surveillance program because of the uncertainties in predicting wear rates for the standard 17x17 fuel assembly design. The objective of this program was to demonstrate that there was no occurrence of hole formation in rodded guide thimble tubes, thus providing some confidence that scramability is ensured.
These applicants fomed an owners' group, which has submitted a generic report (Leasburg, March 1,1982)thatprovidespostirradiationexaminationresults on guide thimble tube wear in the Westinghouse 17x17 fuel assembly des'ign.
Based on this report, we have concluded (Rubenstein, April 19,1982) that the Westinghouse 17x17 fuel assembly design is resistant to guide thimble tube wear.
y,
?
09/21/84 22 V0GTLE 1&2 SER INPUT 9
b -
s
..... ~... -...... _.....
g g e_g
- - - - " ~ "
f}i
'h 7.
e (5) Oxidation and Crud Buildup i
In the FSAR, there is no explicit discussion of cladding oxidation, hydriding, h
and crud buildup. Crud and oxide are mentioned in Sections 4.4.2.9.1, 4.4.2.11, 4.4;2.11.5, and 4.4.4.5.2 of the FSAR. The applicable models for f
cladding oxidation and crud buildup are discussed in the supporting documen-S tation (Salvatori, January 4,1973) for the Westinghouse fuel perfonnance code j
PAD-3.1. These models were previously approved by us. A new temperature-
!l dependent cladding oxidation model is also presented in WCAP-9179. Because the temperature-independent model in PAD-3.1 is conservative with respect to the
]
approved n,odel in WCAP-9179, the NRC continues to find the older models appli-cable. These models affect the cladding-to-coolant heat transfer coefficient J
and the temperature drop across the cladding wall. Mechanical properties and j
analyses of the cladding are not significantly impacted by oxide and crud i
buildup. On the basis of the Westinghouse discussion (Anderson, January 12, l
1981) of the impact of cladding hydriding on fuel performance, and on our t
review of the oxidation and crud buildup models, we conclude that these effects have been adequately accounted for in the Standard fuel Design.
(6) Rod Bowing In Section 4.2.3.1,0 of the FSAR, it is indicated that the model used fnr evaluation of fuel rod bowing is in WCAP-8691 (nonproprietary version is WCAP-8692), which was withdrawn (see NUREG-0390). Revision I to WCAP-8691 was submittedbyWestinghouseandhasbeenapprovedbytheNRC(Rubenstein, October 25,1982). The applicant needs to use Revision 1 to WCAP-8691 as the reference for the fuel rod bowing model and to confirm that the rod bowing analysis for Vogtle-1 and -2 fuel has been performed with this model.
(7) Axial Growth Relativetothediscussion,inSection4.2.1.1(7),above,onstainlesssteel growth, the NRC is aware of supporting information (Bloom, April 1972, and Appleb', April 1972) that was not cited by Westinghouse, but which also implies y
that irradiation growth of stainless steel should not be significant at the 09/21/84 23 V0GTLE 1&2 SER INPUT 1
Im
~-
-. n =+u=a n
-w
- -,..~-
~
[.
V t
f temperatures and fluences that are associated with PWR operation.
Furthermore.
[
because we are unaware of any operating experience that indicates axial-growth-related problems in Westinghouse NSSS plants, we conclude that Westinghouse has I
made sufficient accommodations for control, source, and burnable poison rod axial rod growth in their NSSS designs.
{
TheWestinghouseanalysigofshouldergapspacingfortheSFAhasfoundthat
(
interference will not occur until achieving burnups beyond traditional values.
l The NRC, therefore, finds that the required shoulder gap spacing has been rea-
}
sonably acconnodated. However, for extended burnup applications, the adequacy j-of the spacing should be reverified. Furthermore, because stress-free irradia-tion growth of zirconium-bearing alloys is sensitive to texture (preferred
- [
cystallographicorientation)andretainedcoldwork,which, inturn,are p!
strongly dependent on the specific fabrication techniques that are employed l
during component production, reverification of the design shoulder gap should be performed if Westinghouse current fabrication specifications are signifi-cantly altered.
Finally, we find the Westinghou e analysis of fuel assembly growth to be acceptable. However, as stated in the above discussion on shoulder gap spac-ing, reverification of the fuel assembly growth should be perfonned if signifi-i cant changes are made in the Westinghouse cur' rent fabrication techniques.
(8) Fuel Rod and Nonfuel Rod Pressures As noted in Section 4.2.1.3.B of the FSAR, the analysis of fuel rod internal pressurefortheStandardFuelDesignisdescribedinanapproved(Stolz, May 19, 1978) top,ical report, WCAP-8963. The evaluation relies on the Westinghouse PAD-3.3 fuel performance code, which has also been approved (Stolz, February 9, 1979).
The analysis of the internal pressure of nonfueled rods for the SFA is gen-erallybasedonSectionIII,SubsectionNG-3000,oftheASMECode(see Section 4.2.1.6 of the FSAR). Absorber rod, burnable poison rod, and neutron source rod cladding is cold-worked Type 304 stainless steel, which is not l
l 09/21/84 24 V0GTLE 1&2 SER INPUT i
i h.
_m___.m....-
__._. 3
'l..
li;.
ll covered by the ASME Code. Westinghouse therefore defines as the stress limit
{.
an intensity value Sm equal to 2/3 of the material yield stress. The yield
[
stress for this material is approximately 62,000 psi. A strain limit of I percent also applies to the cladding. Predicted maximum values of rod
[
internal pressure have been provided in the response (Anderson, January 12, 1981, and April N
21,1981)toQ231.2andtheyarewellbelowthoseimposedby I
the cladding stress and strain limits.
1 l-We conclude that there is adequate assurance that nonfueled core component rods
]
can operate safely during Conditions 1 and 2 because appropriate stress and strain limits are met even though the maximum internal rod pressure may exceed system pressure.
sq (9) Assembly Liftoff a
a h
In response to the NRC's question on this topic. Westinghouse has confirmed
{
that momentary liftoff will occur only during a turbine overspeed transient (thisisalsostatedinSection4.4.2.6.2oftheFSAR). Westinghouse has
~
further found that (a) proper reseatt.g will occur after momentary liftoff, ll (b) damage to adjacent assemblies will not occur even if one assembly is fully lifted and the adjacent ones remain seated, and (c) no ill consequences of momentary liftoff are expected. We conclude, therefore, that fuel assembly liftoff has been adequately addressed for the SFA design.
(10) Control Material Leaching While the design basis for the SFA control rods is to maintain cladding integ-rity, and while.the pro,bability of control rod cladding failures appears to be quite low, we have considered the corrosion behavior of the Vogtle-1 and -2 control material and burnable poison and conclude that a breach in the cladding should not result in serious consequences because the Ag-In-Cd or hafnium absorbermaterialandthepoisonmaterial(borosilicateglass)arerelatively inert.
i i
09/21/84 25 V0GTLE 182 SER INPUT g.
= --
- -- y
- v..
7 1
i' 4.2.3.2 Fuel Rod Failure Evaluation The following paragraphs discuss the evaluation of (a) the ability of the SFA
}
fuel to operate without failure during normal operation and anticipated tran-j sients, and (b).the accounting for fuel rod failures in the applicant's acci-J.
dent analysis. ~ The fuel rod failure criteria described in Section 4.2.1.2, above,wereusedforthigevaluation, j
j (1)
Internal Hydridino d
i Westinghouse has used moisture and hydrogen control limits in the manufacture j
of earlier fuel types and has found that typical end-of-life cladding hydrogen levels are less than'100 ppm--a level below which hydride blister formation is l
not anticipated in fuel cladding. We therefore conclude that reasonable I
evidence has been provided that hydriding as a fuel failure mechanism will not
)
be significant in the SFA.
u (2) Claddino Collapse
~
In calculating the time at which cladding col 1 apse will occur, Westinghouse uses the generic methods described in WCAP-8377, which is approved (Stello, January 14,1975) for licensing applications.
Inputs to the analysis include cladding ovality, helium propressurization, free volume of the fuel rod, and limiting power histories.
The applicant has confirmed (FSAR Amendment No. 9. August 1984) that the cal-culated cladding collapse time for Vogtle-1 and -2 using WCAP-8377 methods is more than the expected lifetime of the fuel. Consequently, it is concluded that the criterion for cladding collapse is satisfied.
l 09/21/84 26 V0GTLE 1&2 SER INPUT
,J gn u.
- ..: = =.:.u.
z:
l
-- ~
- u y e 1
_. u.. a... s _ m y,q
!I.
t.:.-
= [
[
(3) Overheating of Cladding 1
As stated in SRP Section 4.2, adequate cooling is assumed to exist when the thermalmargincriteriontolimitthedeparturefromnuclearboiling(DNB)or 3
boiling transition in the core-is satisfied. The method employed to meet the
{
DN8 design basis is reviewed in Section 4.4 of this report.
' t h
(4) Overheating of Fuel ' Pellets i
t
]
j 1The design evaluation of the fuel centerline melt limit is performed with the 8.
Westinghouse fuel performance code, PAD-3.3 (WCAP-8720). This code, which has been approved (Stolz. February 9,1979, and Rubenstein, Jur.e 30,1982), is also used to calculate initial conditions for transierds and accidents described in s
[
' Chapter 15oftheStandardReviewPlan(seeSection4.2.3.3(1),below,for i.
further consents on PAD-3.3).
l In applying the PAD-3.3 code to the centerline melting analysis, the melting
'\\
temperature of the U0 is assumed to be 5080*F unirradiated and is decreased 2
by 58'F par 10,000 mwd /t. This relation has been almost universally adopted by the industry and has been accepted by us in the pas't. The expressions for thermal conductivity and gap conductance, described in Section 4.4.2.11 of the FSAR, are unchanged from that originally described in,the PAD code. We con-sider it unnecessary,to further review these models.
g.
The peak linear heat rating resulting from overpower transients / operator errors I
(assuming a maximum overpower of 118 percent) for Vogtle-1 and -2 is 18.0 kW/ft.
AsnotedinSec{on.4.4.2.11.6 of the FSAR, the centerline temperature at this peak linear hest rating is below that required to produce fuel melting.
~
Consequently, it is concluded that' the criterioh~for the prevention of fuel Jcenterline melting is satisfied.
6 i
L V.
f i
Og/21/84 27 ' '
V0GTLE 182 SER INPUT g p
=. -...... _.- __.._
z.m
=
n _.
-n I
}
(5) Pellet / Cladding Interaction The only two PCI criteria in current use in licensing (1 percent cladding strain and no fuel melting), while not broadly applicable, are easily satis-fled. As noted.in the discussion of the cladding stress and strain evaluation.
{
Westinghouseusesanapprovedcode(PAD)tocalculatecreepstrain,andthe g
values calculated by that code are found to be below the 1 percent strain criterion. And, as indicated in the discussion on overheating failures in l
Section 4.4.2.11.6 of the FSAR, the no-centerline-melt criterion is satisfied.
Therefore, the two existing licensing criteria for PCI have been satisfied.
(
)
In addition to the SRP-type treatment of PCI, however, the response (Anderson, j
January 12, 1981, and April 21, 1981) to Q231.23 and FSAR Section 4.2.3.3(a) address PCI from the standpoint of its effect on fatigue life. PCI produces cyclic stresses and strains that can affect fatigue life of the cladding.
Fur-(
thermore, gradual compressive creep of the cladding onto the fuel pellet occurs due to the differential pressure exerted on the fuel rod by the coolant.
Westinghcuse contends that by using prepressurized fuel rods the rate of clad-ding creep is reduced, thus delaying the time at which fuel-to-cladding contact j
first cccurs. We agree that fuel rod prepressurization should improve PCI 1,
resistance, albeit in a presently unquantified amount.
In conclusion, Westinghouse has used approved methods to demonstrate that the present PCI acceptance criteria have been met.
(6) Cladding Rupture In the LOCA analysis for SFA-designed plants, an empirical model is used to predict the occurrence of cladding rupture. The rupture nodel utilized for the k
large break analysis is stated in Section 15.6.5.3.1.1 of the FSAR to be the 1981 version of the LOCA Evaluation Model; however, the references (8 and 13) stated for that rodel in the last paragraph of that section of the FSAR are incorrect. Tn. c ~ rect refereren (11, which is Revision 1 of WCAP-9220-P-A
]
[trf% CAP-9221-A)andhis5 6 preyed by the NRC) is ir the reference list but is not used in the text of that section of the FSAR.
l I
09/21/84 28 V0GTLE 182 SER INPUT s"n
~ _.
w_1 *
.c 2
-r-
--a---
,. --uxw
^
1 p'.
k-The rupture model utilized for the small break analysis was the approved I!.
October 1975versionoftheECCSevaluationmodel(seeSection 15.6.5.3.1.2 of
[
the FSAR). This model has been found acceptable for this analysis.
s 5'
The appropriate, references for the large break LOCA analysis need to be con-
' l.
firmed. The o'verall impact of cladding rupture on the response of the SFA design to the loss-of-coo,lant accident is evaluated in Section 15.6.5 of this report and is not! reviewed further in this section.
4.2.3.3 Fuel Coolability Evaluation y
The following paragraphs discuss the evaluation of the ability of the SFA fuel to meet the fuel coolability criteria described in Section 4.2.1.3, above.
Those criteria apply to postulated accidents.
(1) Fragmentation of Embrittled Cladding The primary degrading effect of a significant degree of cladding oxidation is 5
embrittlement of the cladding. Such embrittled cladding will have a reduced ductility and resistance to fragmentation. The most severe occurrence of such embrittlement is during a LOCA.- The overall effects of cladding embrittle-ment on the SFA design for the loss-of-coolant accident are analyzed in Sec-tion 15.6.5 of this report and are not reviewed further in this section.
One of the most significant analytical methods that is used to provide input to the analysis in Section 15.6.5 of this report is the steady-state fuel perfor-mance code, which is reviewed in Section 4.2.
This code provides fuel pellet temperatures (stored energy) and fuel rod gas inventories for the ECCS evalua-tion model as prescribed by Appendix K to 10 CFR 50. The code accounts for i
fuel thermal. conductivity, fuel densification, gap conductance, fuel swelling, cladding creep, and other phenomena that affect the initial stored energy.
For
{
.this purpose Westinghouse.uses a veintively new fuel performance code called PAD-3.3,(WCAp-8720). This' code was approved by our safety evaluation (Stolz, j
February 9, 1979, and Rubenstein, June 30,1982),
2 t
~
I 09/21/84 29 V0GTLE 182 SER INPUT e
- ~
~ * * " ' "
~
~ *
'~*
,_ = n =- x
=- _
p.-...
e h
i,
[
For non-LOCA events, the locked rotor accident (one-pump seizure with four
[
loops operating) is the most severe undercooling event that is analyzed. This i
event is analyzed in Section 15.3.3 of the FSAR where it is found that the peak cladding temperature is 1835'F which is well below the 2700'F design limit. The analysis of this event is reviewed in Section 15.3.3 of this
)
report, but it is clear that the SFA meets the non-LOCA peak cladding tem-perature design limit.
(2) Violent Expulsion of Fuel Material i
The analysis that demonstrates that the design limits are met for this event for the SFA is presented in Section 15.4.8 of the FSAR and is reviewed in that section of this report.
- f.(
(3) Cladding Ballooning and Flow Blockage
]
I~
The cladding ballooning and f' low blockage models for the large break LOCA are I
integral parts of the Westinghouse ECCS evaluation model. Consequently, the concern expressed in Section 4.2.3.2(6) of this report as to the appropriate references for the Westinghouse ECCS model used for the large break LOCA anal-ysis needs to be addressed before this analysis can be approved.
4 I.
The cladding ballooning and flow blockage analysis for the small break LOCA was performed with correlations from the approved October 1975 ECCS evaluation model (see Section 15.6.5.3.1.2) of the FSAR). This model has been found to be acceptable for this analysis.
The appropriate. references for the large break LOCA analysis need to be con-firmed. The overall impact of cladding ballooning and assembly flow blockage models on the response of the SFA design to the loss-of-coolant accident is evaluated in Section 15.6.5 and is not reviewed further in this section.
s.
09/21/84 30 V0GTLE 182 SER INPUT t
7 e-a-
..., n.;
-,.._n-_--.-
t E
(4) Structural Damage from External Forces It is stated in Section 4.2.3.4 of the FSAR that Westinghouse has perfomed r
}
theseanalysesutilizingmodelsdescribedinWCAP-8236(andWCAP-8288)and L
WCAP-9401(andWCAP-9402). WCAP-9401 essentially augments the information presented in WCAP-8236 because both WCAP reports apply to similar assemblies.
f WCAP-9401hasbeenrevieyedandapproved(Rubenstein, April 23,1981);there-(
fore, these models are acceptable for these analysis.
(
l[
The maximum grid and non-grid component loads from the safe shutdown earthquake i
,(SSE)andLOCAeventswerecalculatedusingtheaboveapprovedmodels. The maximum impact loads ;from these two events were combined using the square-root-
[,
of-sum-of-squares (SRSS) method, as per SRP 4.2, Appendix A, for each component I
and found to be less than the allowable stresses for each of these Vogtle-1 and
[
-2 components. Consequently, these analyses are found to be acceptable.
1.
I 4.2.4 Testing Inspection, and Surveillance Plans 4.2.4.1 Testing and Inspection of New Fuel
[
As required by SRP Section 4.2, testing and inspection plans for new fuel should include verification of significant fuel design parameters. While details of the manufacturer's testing and inspection programs should be docu-mented in quality control reports, the programs for onsite inspection of new fuel and control assemblies after they have been delivered to the plant should 4
also be described in the FSAR.
l The Westinghouse quality control program that will be applied to Vogtle-1 and
-2 fuel is discussed in Section 4.2.4 of the FSAR and addresses fuel system components and parts, pellets, rod inspection, assemblies, other inspections.
b and process control.
Fuel system component inspection depends on the component parts and includes dimensiqns, visual appearance, audits of test reports, mate-rial certification, and nondestructive examinations. Pellet inspections, for example, are perfomed for dimensional characteristics such as diameter, den-sity, length, and squareness of ends. Fuel rod, control rodlet, burnable i
09/21/84 31 V0GTLE 1&2 SER INPUT t... -,....
- .. -. = -
~
~., _. n.,,-.,.,,.-.
l...
L t
l poison rod, and source rod inspections reportedly consist of nondestructive f
examination techniques such as leak testing, weld inspection, and dimensional measurements. Process control procedures are described in detail.
In-core
,I control component testing and inspection is described in See U'n 4.2.4.4 of
'l the FSAR. In-addition, the applicant states in Section 4.2.4.; of the FSAR I
that if any tests and inspections are to be performed by others on behalf of Westinghouse, Westinghous,e will review and approve the quality control pro-cedures, inspection plans, and so forth, to ensure that they are equivalent to f
the description provided in Sections 4.2.4.1 through 4.2.4.4 of the FSAR and
'l are performed properly to meet all Westinghouse requirements.
I Based on the information provided in Section 4.2.4 of the FSAR and the com-l mitment by Westinghouse to ensure the acceptability of any tests and inspec-tions perfor=cd by others on behalf of Westinghouse, we conclude that the fuel testing and inspection program for new fuel is acceptable.
l 4.2.4.2 On-Line Fuel Failure Monitoring In Section 11.5.2.3 and Table 11.5.2-1 of the FSAR, the applicant has provided a description of the chemical volume and control system (CVCS) letdown monitor for on-line fuel rod failure detection. A definite comitment to use the fuel failure detection instruments is required to meet the guidelines of II.D.2 of SRP Section 4.2.
The applicant has indicated (see response to Q490.4-1 in Amendment No. 9, dated August 1984, to the FSAR) that this information will be provided in Amendment No. 10 to the FSAR.
Section 11.5.2.3 and Table 11.5.2-1 of the FSAR do not include information about the sensitivity of the CVCS letdown monitor for detecting fuel rod fail-ures. The sensitivity of the monitor needs to be confirmed, as stipulated in II.D.2 of SRP Section 4.2.
4.2.4.3 Postirradiation Syrveillance Westirighouse has extensive experience with the use of 17 x 17 standard fuel assemblies in other operating plants. As noted in Section 4.2.3.3.2 of the 09/21/84 32 V0GTLE 1&2 SER INPUT x_ _
.~.~-_ =
=
g.
y~..
ji i
FSAR, this experience is summarized in WCAP-8183, which is periodically updated N
to provide the most recent information on operating plants. Additional test U
assembly and test rod experience is given in Sections 8.and 23 of WCAP-8768 Revision 2.
n t
[
Section 4.2.4.6 of the FSAR indicates that it is currently anticipated by the applicantthatpostirradptionpoolsidesurveillanceoftheVogtle-1and-2
't fuel assemblies will not exceed a qualitative visual examination of some dis-I charged fuel assemblies from each refueling, which satisfies part of II.D.3 of SRP Section 4.2.
To satisfy the remaining part of II.D.3, the applicant must I
(a) make a commitment in the surveillance program to perfom additional sur-veillance if unusual behavior is noticed in the visual examination or if plant instrumentation indicates gross fuel failures and (b) address the dispos'ition of failed fuel in the postirradiation fuel surveillance program.
n 4.2.5 Evaluation Findings The following have not yet been provided by the applicant:
(1) Confimation of the correct references for the cladding rupture and clad-ding ballooning and flow blockage models for the large break LOCA (see l
Paragraphs 4.2.3.2(6) and 4.2.3.3(3) in this report).
\\
(2) A comitment to use the on-line fuel failure detection methods (see Paragraph 4.2.4.2 in this report).
(3) A statement about the sensitivity of the chemical volume and control system letdown monitor for detecting fuel rod failures (see Para-graph 4.2.4.2 in this report).
(4) A commitment in the postirradiation fuel surveillance program to perform additional surveillance if unusual behavior is not6d in the visual exami-nation or if plant instrumentation indicates gross fuel failures (see Paragraph 4.2.4.3 in this report).
09/21/84 33 V0GTLE 1&2 SER INPUT W
.-=.-=.=m
-t
-;.-----=
a -. a w -
= p..-;:w.,+.w - ~.c. n i:
[...
.j:
I 7
(5)
A' statement in the postirradiation fuel surveillance program that addresses the disposition of failed fuel (see Paragraph 4.2.4.3 in L
this report),
t,. '
I, g
(6) Confirmation that the rod bowing analysis has been perfomed with the latest approved mod,el (see Paragraph 4.2.3.1(6) in this report).
t (7) Confirmation of the correct value for the peak linear power for nomal operation (in the FSAR, Table 4.1-1 shows 12.6 kW/ft but Table 4.4-1 shows
[
12.5kW/ft).
.I l l Whentheabovearephovided,wewillconcludethattheVogtle-1and-2 fuel has been designed so that (a) the fuel system will not be damaged as a result of normal opertion and anticipated operational ocurrences, (b) fuel damage f-during postulated accidents would not be severe enought to prevent control rod h
insertion when it is required, and (c) core coolability will always be main-r tained, event after severe postulated accidents, and thereby meets the related requirements of 10 CFR Part 50.46; 10 CFR Part 50, Appendix A; GDC 10, 27, and 35; 10 CFR Part 50, Appendix X; and 10 CFR Part 100. This conclusion is based on the following.
(1) The applicant has provided sufficient evidence that these design objectives will be met based on operating experience, prototype testing, and analytical predictions. ThosepredictionsgNQith'structur response, control rod ejection, and fuel densification have been performed in accordance with (1) the guidelines of Regulatory Guide 1.77, and methods that the staff has reviewed and found to be acceptable alternatives to Regulatory Guides 1.60 and 1.126, and (b) the guidelines for " Evaluation of Fuel Assembly Structural Response to Externally Applied Forces" in Appendix A to SRP Section 4.2.
1, (2) The applicant has provided for testing and inspection of the fuel to ensure that it is within design tolerances at the time of core loadings.
L 09/21/84 34 V0GTLE 1&2 SER INPUT K1;.
.. = = -.. =
.. =
},,.
wa- -
of j
I:
c, r,
t*
e O
9 e
e it I[a
[f
/
N,a 6
A A'
h 4
4 J
b.
e 4
e 09/21/84 35 V0GTLE 1&2 SER INPUT pr.z
.. = ::-
-. - - - =
" - =
,a. :.=m
, m,, ;. m _ u m _.
- +.
---_;.=
?
^
I The applicant has made a comitment to perfom on-line fuel failure moni-f toring and postirradiation surveillance to detect anomalies or confirm j
that the fuel has performed as expected.
?
r I
We conclude that the applicant has described methods of adequately predicting fuel rod failures during postulated accidents so that radioactivity releases
{
are not underestimated aqd thereby meets the related requirements of 10 CFR
{
Part 100.
Inmeetingtheserequirements,theapplicanthas(a)usedthefis-sion product release assumptions of Regulatory Guides 1.4,1.25, and 1.77, and
)
(b) performed the analysis for fuel rod failures for the rod ejection accident k
in accordance with the guidelines of Regulatory Guide 1.77.
On the b' asis of the review, it is concluded that the applicant's fuel system design has met all the requirements of the applicable regulations, regulatory guides, and current regulatory positions.
i 4.2.6 References Anderson, T. M., Westinghouse Letter to J. R. Miller, USNRC, " Responses to i L Request Number 3 for Additional infomation on WCAP-9500, NRC letter from J. R. Miller to T. M. Anderson, August 15, 1980," dated January 12, 1981.
1 Anderson, T. M., Westinghouse, Letter to J. R. Miller, USNRC, " Responses to Questions on WCAP-9500 Section 4.2, Resulting from NRC/ Westinghouse Meeting on March 18, 1981," dated April 21, 1981.
ANSI N-212-1974, American National Standard Nuclear Safety Criteria for the Design of Stationary Boiling Water Reactor Plant _s_, American Nuclear Society, LaGrange Park Illinois.
ANSI N18.2-1973, American National Standard Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants, American Nuclear Society, LaGrange Park, Illinois.
i F
i 09/21/84 36 V0GTLE 1&2 SER INPUT h
_ _ _ _.. ~,. _
gs,#.m
. %%_ m _ ~.
m uc__
a,.
k Appleby, W. K., et al., " Fluence and Temperature Dependence of Void Formation l.
in Highly Irradiated Stainless Steels," p. 156. Radiation-Induced Voids in Metals, USAEC Proceedings of the 1977 International Conference held at Albany, j
New' York, June 9-11, 1971 (April 1972).
I I
ASME Boiler and Pressure Vessel Code,Section III, " Rules for Construction of f
Nuclear Power Plant Components," American Society of Mechanical Engineers, New i-York, New York.
Bloom, E. E., " Nucleation and Growth of Voids in Stainless Steels During Fast-
)
?!eutron Irradiation," p.1, Radiation-Induced Voids in Metals, USAEC Proceedings of the 19.77 International Conference held at Albany, New York, June 9-11, 1971 (April 1972).
(
I Clark, R. A., USNRC, Letter to J. M. Griffin (Arkansas Power and Light Com-pany), "ANO-2 Fuel Inspection Results," July 24,1983(DocketNo.50-368).
Foster, J. P., and R. V. Strain, " Empirical Swelling Equations for Solution-
~ Annealed Type 304 Stainless Steel," Nuclear Technology, Vol. 24, p. 93 (October 1974).
Final Safety Analysis Report of Robert Emmett Ginna Nuclear Power Plant Unit 1, Rochester Gas and Electric Corporation, USNRC Docket Number 50-244, 103, dated 1972.
Fukishiro', T., et al., The Influence of Coolant Flow on' Fuel Behavior Under Reactivity Initiated Accident Corditions, JAERI-M-9104, Japanese Atomic Energy Research Institute, October 1980.
1 Hoshi, T., et al., Fuel Failure Behavior of PCI-Remedy Fuels Under Reactivity
' Initiated Accident Conditions, JAERI-M-8836, Japanese Atomic Energy Research Institute, May 1980.
s i
JAERI-M-9011, Semiannual Program' Report on the NSRR Experiments - July to December 1979 Japanese Atomic Energy Research Institute, September 1980.
09/21/84 37 V0GTLE 1&2 SER INPUT
=m iy....
j0 N
Kuffer, K., and H. R. Lutz, " Experience of Commercial Power Plant Operation in
.[
Switzerland, Fifth Foratom Conference. Florence Italy,1973.
' $!p Leasburg, R. H., VEPCO, Letter to H. R. Denton, USNRC, " Fuel Assembly Guide
'll Thimble Tube Wear Examination Report," dated March 1, 1982.
}lD i,
Nerses, V., USNRC, Letter to Florida Power and Light Co. (FP&L), "Sumary of
. Meeting held in Bethesda, MD with FP&L on March 3, 1983 Regarding Axial Growth j
and High Burnup Fission Gas Release," April 28, 1983.
i!
NUREG-0390, Vol. 7, No.1, Topical Report Review Status, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Comission, July 15, 1983.
fl NUREG-0800, Revision 2, Standard Review Plan, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Comission, July 1981 (Fomerly NUREG-75/087).
Petrick, N., SNUPPS, Letter to H. Denton, USNRC, dated August 31 and September 9, 1981.
1 Rubenstein, L. S., USNRC, Memorandum to R. L. Tedesco, USNRC, " Review of Westinghouse Optimized Fuel Assembly Topical Report (TAR-5254) " dated August 8, 1980.
Rubenstein, L. S., USNRC, Memorandum to R. L. Tedesco, USNRC, " Review of Topical Report WCAP-8278," dated March 19, 1981.
[
Rubenstein, L. S., USNRC, Memorandum to R. L. Tedesco, USNRC, " Review of j-Topical Report WCAP-9401," dated April 23, 1981.
I-i-
Rubenstein, L. S., USNRC, Memorandum to R. L. Tedesco, USNRC, " Safety
-Evaluation Report on WCAP-9500," dated May 15, 1981.
Rubenstein, L. S., USNRC, Memorandum to T. M. Novak USNRC, " Resolution of the Westinghouse Guide Thimble Tube Wear Issue," dated April 19, 1982.
09/21/84 38 V0GTLE 1&2 SER INPUT
% ww.e
~ -
~~
j!
l i
Yi Rubenstein, L. S., USNRC, Memorandum to R. L. Tedesco, USNRC, " Safety M
Evaluation of WCAP-8720 Addendum 1," da'nd June 30, 1982.
?!tiji Rubenstein, L. S., USNRC, Memorandum to T. M. Novak, USNPC, "SERs for b
Westinghouse. Combustion Engineering, Babcock & Wilcox, and Exxon Fuel Rod l'l Bowing Topical Reports,". dated October 25, 1982.
~
r Rubenstein, L. S., USNRC Memorandum to F. J. Miragalia, USNRC, "SER On Supplemental Information to WCAP-9179, Revision 1 " dated June 6, 1983.
Rubenstein, L. S., USNRC, Memorandum to G. C.*Lainas. USNRC, "ANO-2 Fuel I
Inspection Results," dated June 17, 1983.
i Salvatori, R., Westinghouse Letter to D. Knuth, USNRC, " Core Coolant and Rod Surface Temperature," dated January 4,1973.
i lj Schenk. H., " Experience from Fuel Performance at KWO," SM-178-15. International Atomic Energy Agency, dated October 1973.
h Shiozawa, S., et al., " Evaluation on Oxidation of Zircaloy-4 Cladding During j
Rapid Transient in NSRR Experiments," JAERI-M-8187, Japanese Atomic Energy t.
l Research Institute, March 1979.
j Stello, V., USNRC, Memorandum to R. DeYoung, " Evaluation of Westinghouse Report i
WCAP-8377, Revised Clad Flattening Model," dated January 14, 1975.
Stolz, J. F., USNRC, Letter to T. M. Anderson, Westinghouse, " Safety Evaluation p
of WCAP-8963," dated May 19, 1978.
t Stolz, J. F., USNRC, Letter to T. M. Anderson, Westinghouse, " Safety Evaluation of WCAP-8720," dated February 9,1979.
Tedesco, R. L., USNRC, Let,ter o T. M. Anderson, Westinghouse, " Acceptance for Referencing of Licensing Topical Report WCAP-9500," dated May 22, 1981.
09/21/84 39 V0GTLE 1&2 SER INPUT x v= =...
c.= = --
~. --.
~
. ~..
Ih
.{'.4 7
Tokar, M., USNRC, " Report to ACRS Concerning NRR Efforts on Pellet / Cladding Interaction," dated November 14, 1979.
Van Houten, R., USNRC, Memorandum to M. Tokar. USNRC, " Demonstrated Surviva-bility of Zircaloy Cladding After Brief Exposure to Temperatures of 1775*K and j
Above," dated February 23. 1981.
i r
!f
-WCAP-7588. An Evaluation of the Rod Ejection Accident in Westinghouse Pres-surized Water Reactors Using Special Kinetics Methods, Rev.1 (Nonproprietary).
- }
[ Approved by USNRC on January 7, 1975: see NUREG-0390, 7(1):II-134, July 15, 1983.]
j WCAP-8183 (revised annually).
Iorii, J. A. and Skaritka, J., Operational Experience with Westinghouse Cores.
I WCAP-8236 (Proprietary) and WCAP-8288 (Nonproprietary). Gesinski, L. and,
t Chiang. D., Safety Analysis of the 17x17 Fuel Assembly for Combined Seismic and k
Loss-of-Coolant Accident, 1973.
WCAP-8278 (Proprietary) and WCAP-8279 (Nonproprietary). Demario, E. E.,
Hydraulic Flow Test of the 17x17 Fuel Assembly, 1974.
1 s
WCAP-8377 (Proprietary) and WCAP-8381 (Nonproprietary). George, R. A.;
Lee, Y. C.; and Eng, G. H., Revised Clad Flattening Model,1974.
WCAP-8691 (Proprietary) and WCAP-8692 (Nonproprietary). Reavis, J.
R., et al.,
Fuel Rod Bowing,1975.
WCAP-8691, Revision 1 (Proprietary) and WCAP-8692, Revision 1 (Nonproprietary).
Skaritka, J., et al., Fuel Rod Bow Evaluation, July 1979.
WCAP-8720(Proprietary)an(WCAP-8785(Nonproprietary). Miller, J. V., ed.,
Improved Analytical Models Used in Westinghouse Fuel Rod Design Computations, t
1976.
t
?
5 09/21/84 40 V0GTLE 182 SER INPUT m u
_-., _ ;.__ =
_=._1-_
,j gg,_ -
t.
Jl.-
j l'
l
[.
WCAP-8768 Revision 2.
Eggleston, F. T., Safety-Related Research and Devel-E opment for Westinghouse Pressurized Water Reactors, Program Sumaries - Winter j
1976 - Sumer 1978, October 1977.
f WCAP-8963(Proprietary)andWCAP-8964(Nonproprietary). Risher, D. H., et al.,
[
Safety Analysis for the Revised Fuel Rod Internal Pressure Design Basis,
~
November 1976 and August 1977, respectively.
I h
WCAP-9179, Revision 1(Proprietary)andWCAP-9224(Nonproprietary). Beaumont, M. D., et al., Properties of Fuel and Core Component Materials,1978.
i WCAP-9220-P-A (Proprietary) and WCAP-9221-A (Nonproprietary), Revision 1.
Eicheidinger, C., Westinghouse ECCS Evaluation Model, 1981.
t WCAP-9401-P-A (Proprietary) and WCAP-9402-A (Nonproprietary). Beaumont, M. D.,
et al., Verification Testing and Analysis of the 17x17 Optimized Fuel Assembly, t
August 1981.
WCAP-9500. Reference Core Report 17x17 Jptimized Fuel tssembly, July 1979.
i 09/21/84 41 V0GTLE 1&2 SER INPUT x
_ ~,
a I
l..
f 4.3 Nuclear Design
)
The Vogtle Units 1 and 2 power plants each have a reactor core consisting of
[
193 fuel assemblies of the Westinghouse standard 17x17 design. The core has I
a design heat output of 3411 thermal Megawatts and is similar in most respects
,l to the Callaway reactor and other recent Westinghouse 4 loop reactors.
We l
have reviewed'the nuclear design of the Vogtle reactors. Our review was j
conducted in accordance, with the guidelines provided by the Standard Review Plan, Section 4.3, and 'was based on information contained in the Final Safety Analysis Report, amendments thereto, and the referenced Topical Reports.
4.3.1 Design Bases Design bases are pr,esented which comply with the applicable General Design Criteria.
Acceptable fuel design limits are specified (GDC 10), a negative g
prompt feedback coefficient is specified (GDC 11) and tendency toward i
divergent operation (power oscillation) is not permitted (GDC 12). Design
, bases are presented which require a control and monitoring system (GDC 13) 3 which automatically initiates a rapid reactivity insertion to prevent j
exceeding fuel design limits in normal operation or anticipated transients I
(GDC20). The control system is required to be designed so that a single malfunction or single operator error will cause no violation of fuel design limits (GDC25). A reactor coolant boration systs is provided which is capable of bringing the tractor to cold shutdown conditions (GDC 26) and the control system is required to control reactivity changes during accident conditions when combined with the engineered safety features (GDC 27).
Reactivity accident conditions are required to be limited so that no damage to the reactor coolant system boundary occurs (GDC 28).
i
^
j We find the design bases presented in the FSAR to be acceptable.
]
4.3.2 Design Description j
The FSAR contains the description of the first cycle fuel loading which consists of three different enrichments and has a first cycle length of i
approximately one and a half years. The enrichment distribution, burnable il
~
l I
w i3 gp. ::= -. =
.- :... _ =
_ = _
.
..m-_
]
i poison distribution, soluble poison concentration and higher isotope j.
(actinide) content as a function of core exposure are presented.
Values presented for the delayed neutron fraction and prompt neutron lifetime
[
at beginning and end of cycle are consistent with those normally used l-and are acceptable.
\\a-t Power Distribution
'f The design bases affecting power distribution are:
I
{
The peaking factor'in the core will not be greater than 2.30 during normal operation of full power in order to meet the initial conditions assumed in the loss of coolant accident analysis.
I l' l l-Under normal conditions (including maximum overpower) the peak fuel power will not ' produce fuel centerline melting.
I il The core will not operate during normal operation or anticipated d
operational occurrences, with a power distribution that will cause h
the departure from nucleate boiling ratio to fall below 1.3 (W-3 l
correlation with modified spacer factor).
1' The 2.30 F peaking factor is determined and maintained via calculations of g
l extren:es of allowed transient power distributions and periodically measured radial power distributions and radial peaking factors F and F These a H.
also provide maximum initial conditions for events described in Section 15 which assure that peak full power does not cause center line fuel melting or result in departure f.om nucleate boiling during anticipated operational occurrences.
The applicant has descrioed the manner in which the core will be operated and power distribution monitored so as to assure that these limits are met. The core will be operated in the Constant Axial Offset Control (CAOC) mode q
which has been shown to result in peaking factors less than 2.30 for both constant power and load following operation. The applicant has elected to use an improved load follow package, developed by Westinghouse, in Vogtle Units 1 and 2.
I 1
2
m___
. =
m- -
~..
i
[
i CAOC is described in WCAP-8385 (Proprietary) and WCAP-8403 (non-Proprietary),
" Power Distribution Control and Load Following Procedures." This report contains methodology for operation with and without p2rt length control rods.
The former mode allows better return to power capability than the latter.
Use of part length rods has been withdrawn from Westinghouse reactors.
The improved load follow strategy provides a return to power capability during
)
operation without part 1.ength rods comparable to the level previously I
obtainable fraa operation with part length rods.
I The improved load follow strategy involves a redesignated control rod bank and modified overlap that allows greater reactivity insertion than the former design bank within the constraints of a widened, asymetric CAOC band.
The control bank has been changed from eight to four rods.
The four rods removed from the control ban' k have been reassigned as a shutdonw bank, thus main-
)
taining shutdown margins.
The CAOC band has been changed from 15 to +3, -12 delta flux difference. The greater inserted reactivity is available for return to power capability upon control rod withdrawal.
Another element in
.the load follow strategy is the use of moderator temperature reductions to augment return to power capability.
The temperature reduction adds reactivity during rapid return to power through the inherently negative moderator temperature coefficient.
4 The analysis used to calculate the maximum peaking factor which can occur using the improved strategy expands the set in the CA0C topical report to 18 calculational cases. However, with the reassigned control bank, maneuvers resulting in greater control rod insertion for a longer duration become operationally practical but tend to become slightly more limiting in terms of total peaking factors.
Therefore, simulated load follow maneuvers which return the flux difference to the target value (and thereby reduce control rod insertion) have been replaced by load follow strategies which maintain the deeper rod insertion. As a result of our evaluation, we agree with Westinghouse's conclusion that substitution of these more conservative cases will maintain the {hniting nature of the 18 case load following analysis.
l l
L r.
_.. w The analysis performed by Westinghopse indicated that the peaking factor limit could not be met at BOL of Cycle 1 due to the wide flux difference band.
This resulted in limiting the width of the band for the first 20% of the cycle typically, and until 3,000 MWD /MTU burnup for Vogtle Units 1 and 2 to the value of 25%.
This 15% is the value previously justified by the CAOC analysis.
These features will be incorporated in the Vogtle Units 1 and 2 Technical Specifications.
l We conclude, for the r asons stated above, that the improved load follow package will continue to prevent the 2.30 peaking factor limit from being exceeded in normal operation of the power plant, and therefore is acceptable.
5 The two types of instrumentation systems are nonnally provided to monitor core t
l power distribution.
Excore detectors with two axial sections are used to monitor core power, axial offset and azimuthal tilt for the 2.30 F limit, and q
movable incore detectors permit detailed power distributions to be measured.
These systems are used in operating reactors supplied by Westinghouse and we find their use acceptable for Vogtle Units 1 and 2 when a 2.30 limit is the minimum requirement (or possible lower when cycle specific 18 case or equivalent analyses so indicate).
Reactivity Coefficients The reactivity coefficients are expressions of the effect on core reactivity of changes in such core conditions as power, fuel and moderator temperature, moderator density, and baron concentration. These coefficients vary with fuel burnup and power level.
The applicant has presented values of the coefficients in the FSAR and has evaluated the uncertainties of these values.
We have reviewed the calculated values of reactivity coefficients and have concluded that they adequately represent the full range of expected values.
We have reviewed the reactivity coefficients used in the transient and accident analyses and conclude that they conservatively bound the expected values, including uncertainties.
Further, moderator and power Doppler coefficients along with horon worth are measured as part of the startup phy. sics testing to assure'that actual values are within those used in these anal;yses.
4
=
=.
m g..
l' l..
i l i
control i
To allow for changes in reactivity due to reactor heatup, load following, and I
fuel burnup with consequent fission product buildup, a significant amount of i
excess reactivity is built into the core. The excess reactivity is controlled by a combination of full length control rods and soluble boron.
Soluble boron
]
is used to control changes due to:
I i
t f
Moderator density and temperature changes from ambient to operating i
tarperatures.
l Equilibruim xenon and samarium buildup.
I Fuel depletion and fission product buildup - that portion not control, led j
by lumped burnable poison.
Transient xenon resulting from load following.
Control rods are used to control reactivity change due to:
Moderator reactivity changes from hot zero to full power.
Fuel temperature changes (Doppler reactivity changes).
Burnable poison rods placed in some fuel assemblies are used for radial flux shaping and to control part of the reactivity change due to fuel depletion and fission product buildup.
The applicant has provided data to show that adequate control exists to satisfy the above requirements with enough additional control rod worth to provide a hot shutdown effective multipliation factor less than the design basis value of 0.987 during initial equilibrium fuel cycles with the most reactive control rod stuck out the core.
In addition, the chemical and volume control systen, will ha capable of shutting down the reactor by adding soluble boron and maintianing it shut down in the colo, xenon free condition at any time in core life.
These two systems satisfy the require-ments of General Design Criterion 26.
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I Comparisons have been made between calculated and measured control rod bank i
worth in operating reactors and in critical experiments.
These comparisons
)
lead to the conclusion that bank worths may be calculated to within approxi-mately ten percent.
In addition bank worth measurements are performed as part of the startup test program to assure that conservative values have been g
used in safety analyses.
t l
Based on these comparis,ons, we conclude that the applicant has made suitably f
conservative assessments of reactivity control requirements and that 9
l adequate control rod worths have been provided to assure shutdown capability.
}
Control Rod patterns and Reactivity Worths The control rods are divided into two categories - shutdown rods and regulating rods.
The shutdown rods are always completely out of the core when the reactor
?
is at operating conditions. Core power changes are made with regulating rods i
which are nearly out of the core when it is operating at full power.
Regulating rod insertion will be controlled by power-dependent insertion limits required in the Technical Specifications to assure that:
There is sufficient negative reactivity available to permit rapid shutdown of the reactor with adequate margin.
The worth of a control rod that might be ejected is not greater than that which has been shown to have acceptable consequences in the safety analyses.
We have reviewed the calculated rod worths and the uncertainties in these worths, and conclude that rapid shutdown capability exists at all times in core life assuming the most reactive control rod assembly is stuck out of the core.
Stability The stability of the Voggle Units 1 and 2 cores to xenon induced spatial oscillations is discussed in the FSAR. The overall negative reactivity (power) coefficient provides,.ajisurance that the reactor will be stable 6
m
_ w.
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I against total power oscillation. The applicant also concluded that sustained
[
radial or azimuthal xenon oscillations are not possible. This conclusion is based on measurements on an operating reactor of the same dimensions which
[
showed stability against these oscillations.
We concur with this conclusion.
l This core is predicted to be unstable with respect to axial xenon oscillations 8
after about 12000 Megawatt days per ton of exposure. The applicant has accept-ably shown that axial oscillations may be controlled by the regulating rods 4
to prevent reaching any fuel damage limits.
!f Criticality of Fuel Assemblies I
Criticality of fuel assemblies outside the reactor is precluded by adequate design of fuel transfer and storage facilities.
The applicant presents infomation on calculational techniques and assumptions used to assure that jy criticality is avoided.
We have reviewed this information and the criteria l{
which will be employed and find them to be acceptatie.
Vessel Irradiation Values are presented for the neutron flux in various energy ranges at mid-height of the pressure vessel inner boundary.
Core flux shapes calculated by standard design ' methods are input to a transport theory calculation (Sn) 10 which results in a neutron flux of 2.1 x 10 neutrons per square centimeter 6
per second having energy greater than 10 electron-volts at the inner vessel 19
]
boundary.
This results in a fluence of 2.2 x 10 neutrons per square
+
centimeter for a forty year vessel life with an 80 percent use factor. The methods used for these calculations are state of the art, and we conclude that acceptable analytical procedures have been used to calculate the vessel fluence. The Materials Engineering Branch will review the require-i ments for surveillance programs and the pressure-temperature limits for i
j{
operation.
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4.3.3 Analytical Methods g
The applicant has descripd the computer programs and' calculational techniques f
used to obtain the nuclear characteristics of the reactor design.
The calcu-l{
lations consist of three distinct types, which are performed in sequence:
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- determination of effective fuel temperatures, generation of macroscopic few-3 i
group parameters, and space-dependent few-group diffusion calculations. The f
programs used (e.g., LASER, TWINKLE, LEOPARD, TURTLE and PANDA) have been applied as part of the applications for most earlier Westinghouse designed f
nuclear plant facilities and the predicted results have been compared with measured characteristics obtained during many startup tests for first cycle
(
and reload cores. These' results have validated the ability of these i
methods to predict expe,rimental results.
We, therefore, conclude that these
[
methods are acceptable for use in calculating the nuclear characteristics of Vogtle Units 1 and 2.
4.3.4 Summary of Evaluation Findings The Vogtle Units 1 and 2 nuclear design was reviewed according to Section 4.3 of the St'andard Review Plan (NUREG-0800). All areas of review and review procedures fraa that section have been followed either for this reactor or for previous similar reactors (e.g., Callaway) or for Topical Report reviews.
The applicant has described the computer programs and calculational techniques used to predict the nuclear characteristics of the reactor design and has provided examples to demonstrate the ability of the analyses to predi<:t reactivity and physics characteristics of Vogtle Units 1 and 2.
To allow for changes of reactivity due to reactor heatup, changes in operating conditions, fuel burnup, and fission product buildup, a significant amount of excess reactivity is designed into the core.
The applicant has provided sub-stantial information relating to core reactivity balances for the first cycle and has shown that means have been incorporated into the design to control excess reactivity at all times. The applicant has shown that sufficient control rod worth is available to make the reactor subcritical with an 1
effective multiplication factor no greater than 0.987 in the hot condition at any time during the cycle with the most reactive control rod stuck in the fully withdrawn position. On the basis of our review, we conclude that the applicant's assessmen$ of reactivity control requirements over the first core cycle is suitably conservative, and that adequate negative worth has been "provided by the control system to assure shutdown capability.
Reactivity 8
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control requirements will be reviewed for additional cycles as this information
{
becomes available. We also conclude that nuclear design bases, features, and limits have been established in conformance with the requirements of General o
Design Criteria 10,11,12,13, 20, 25, 26, 27 and 28.
b This conclusion is based on the following:
1.
The applicant has, met the requirements of GDC 11 with respect to prompt inherent nuclear feedback characteristics in the power operating range by:
Calculating a negative Doppler coefficient of reactivity, and a.
b.
Using calculational methods that have been found acceptable.
i The staff has reviewed the Doppler reactivity coefficients in this case and found this to be suitably conservative.
2.
The applicant has met the requirements of GDC 12 with respect to power oscillations which could result in conditions exceeding specified acceptable fuel design limits by:
Showing that such power oscillations are not possible and/or a.
can be easily detected and thereby remedied, and b.
Using calculational methods that have been found acceptable.
3.
The applicant has met the requirements of GDC 13 with respect to provisions of instrumentation and controls to monitor variables and systems that can affect the fission process by:
l a.
Providing instrumentation and systems to monitor the core I
power distribugion, control rod positions and patterns, and other process variables such as temperature and pressure, rad
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Providing suitable alarms and/or control room indications for I
these monitored variables.
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4.
The applicant has met the requirements of GDC 26 with respect to provision for two independent reactivity control systems of different f
designs by:
f Having a system that can reliably control anticipated operational a..
occurrences, i
b.
Having a system tha't can hold the core subcritical under cold conditions, and Having a system that can control planned, normal power changes.
c.
/
5.
The applicant has met the requirements of GDC 27 with respect to reactivity control systems that have a combined capability in conjunction with poison addition by the emergency core cooling systen of reliably controlling reactivity changes under postulated accident conditions by:
a.
Providing a movable control rod systen and a liquid poison system, and L
b.
Performing calculations to demonstrate that the core has sufficient shutdown margin with the highest-worth stuck rod.
6.
The applicant has met the requirements of GDC 28 with respect to postulated reactivity accidents by (reviewed under Section 15.4.8):
a.
Meeting the regulatory position in Regulatory Guide 1.77, b.
Meeting the criteria on the capability to cool the core, and t.
Using calculational methods that have been found acceptable for reactivity insertion accidents.
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The applicant has met the requirments of GDC 10, 20, and 25 with respect to specified acceptable fuel design limits by providing analyses demonstrating:
That normal operation, including the effects of anticipated a.
operational occurrences, have met fuel design criteria.
/
b.
That the automatic initiation of the reactivity control system assures that fuel design criteria are not exceeded as a result of anticipated operational occurrences and assures the auto-matic operation of systems and components important to safety under accident conditions, and That no single malfunction of the reactivity control system causes c.
violation of the fuel design limits.
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REFERENCES f
L WCAP-8385, T. Morita, et al., " Power Distribution Control and Load Follow Procedure", Septenber 1974.
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V0GTLE UNITS 1 AND 2 i
Safety Evaluation Report 4.4 Thermal-Hydraulic Design f
4.4.1 Performance ar.d Safety Criteria The performance and safety criteria for the Vogtle Units 1 and 2 core design as stated in Section 4.4.1 of the FSAR are:
(1)
" Fuel damage (defined as' penetration of the fission product barrier, i
1.e., the fuel rod clad) is not expected during normal operation and operational transients (Condition I) or any transient conditions ji arising from faults of moderate frequency (Condition II).
It is not f
possible, however, to preclude a very small number of rod failures.
These will be within the capability of the plant cleanup system and are consistent with the plant design bases."
(2) "The reactor can be brought to a safe state following a Condition III event with only a small fraction of fuel rods damaged (see above definition) although sufficient fuel damage might occur to preclude immediate resumption of operation."
(3) "The reactor can be brought to a safe state and the core can be kept subcritical with acceptable heat transfer geometry following transients arising from Condition IV events."
4.4.2 Design Bases The performance and safety criteria listed above are implemented through the following design bases.
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I 4.4.2.1 Departure from Nucleate Boiling l
The margin to departure fran nucleate boiling at any point in the core is expressed in terms of the departure from nucleate boiling ratio (DNBR).
The DNBR is defined as the ratio of the heat flux required to produce departure from nucleate boiling et the calculated local coolant conditions
{
to the actual local heat flux.
The thermal-hydraulic design basis,sas stated in Section 4.4.1.1 of the l
Vogtle FSAR for the prevention of departure from nucleate boiling is as follows:
"There will be a 95 percent probabi ity that departure from nucleate boiling (DNB) will not' occur on the limiting' fuel rods during normal opeiation and j
operational transient; and any transient arising from faults of moderate frequency (Condition I and II events)'at a 95 percent confidence level."
~ ~
4.4.2.2 Fuel ' Tem hrature t
- The fuel temperature design basis givea in Section 4.4.1.2 is:
l "During modes of operation associated with Condition I and Condition II events, I
l there is at least a 95 percenc probability that the peak kW/ft fuel rods will not exceed the U0 melting te perature at the 95 percent confidence level."
2 i
This design basis is evaluated in the Safety Evaluation Report on Section 4.2
" Fuel System Design."
f 4.4.2.3 Core Flow Section 4.4.1.3 of the FSAR has the following core flow design basis.
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"A minimum of 95.5 percent of the primary coolant flow will pass through the fuel rod region of the core and be effective for fuel rod cooling."
I-4.4.2.4 Hydrodynamic Stability The hydrodynamic stability design basis given in Section 4.4.1.4 is as follows.
" Modes of operation associated with Condition I and II events shall not lead to hydrodynamic instability."
4.4.3 Thermal Hydraulic Design Methodology 4.4.3.1 Departure from Nucleate Boiling The thermal-hydraulic design analysis was performed using the W-3 Critical Her.t Flux (CHF) correlation in conjunction with the THINC-IV analysis.
THINC-IV is an. open channel computer code which determines the coolant density, mass velocity, enthalpy, vapor void, static p: essure, and DNBR distribution along parallel flow channels within a reactor core.
The W-3 correlation was developed from data obtained frcm experiments conducted with fluid flowing inside single heated tubes. As test procedures progressed to the use of rod bundles instead of tubes, the correlation was modified to include the effects of "R" and "L" mixing vane grids and axially non-unifom power distributions.
A correlation factor is developed to adopt the W-3 correlation to 17x17 fuel assemblies with top split mixing vane grids (R grid).
This correlation factor, termed the " modified spacer factor," was developed as a multiplier on the W-3 correlation.
A description of the 17x17 fuel assembly test program and a summary of the results are described in the NRC approved WCAP-8298-P-A and WCAP-8299-A.
The test program predicted heat flux includes a 0.88 multiplier which is part of the 17x17 modified spacer factor. However, a multiplier of 0.86 has been conservatively applied for all DNB analyses. The test,results 3
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indicated that a reactor core using this geometry may operate with a minimum DNBR of 1.28 and satisfy the design criterion. However, a minimum DNBR of 1
1.30 is conservatively used for this plant.
The applicant has proposed this minimum departure from nucleate boiling ratio of 1.30 to ensure that there is a 95 percent probability at a 95 percent confidence that critical. heat flux will not occur on the limiting fuel rod.
The use of the W-3 correlation with a minimum DNBR of 1.30 has been previously approved by the staff. '/
i A description of the THINC-IV computer code is given in hCAP-7956, "THINC-IV L
An Improved Program For Thermal-Hydraulic Analysis of Rod Bundle Cores." The design application o.f the THINC-IV program is given in detail in WCAP-8054,
[
" Application of the THINC-IV Program to PWR Design." Both WCAP-7956 and i
WCAP-8054 have been reviewed and approved by the staff.
The staff has previously reviewed under a different docket, a November 2,1977 t
letter from C. Eiche1dinger (Westinghouse) to J. Stolz (NRC) which described THINC-IV analyses using a cosine upper plenum radial pressure gradient with a maximum value of 5 psi at the core center and 0 psi at the periphery. The
^
results of these analysAs showed that the effects of a core pressure distri -
)
bution on the minim'um DMR are negligible. The staff conducted a similar sensitivity study using ' COBRA-IV.
Our results also showed that the effects are small (NUREG-0847). Based on these analyses, the staff concludes that the use of a nonuniform exit pressure gradient in the Vogtle Units 1 and 2 thermal-hydraulic design is acceptable.
The design calculational procedure, using THINC-IV, is to perform a' core wide analysis followed by a hot assembly and hot subchannel analysis.
[
,i For the hot assenbly and hot subchannel analyses, a set of hat.' channel factors are used to account for deviations due to manufacturing toleranchs. A reload review of a pressurized water reactor, not of Westinghouse design, showed that the hot channel factors used in the thermal-hydraulic analysis of the initial core did not bound future cycles (i.e., beyond the first cycle). The staff 4
l
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questioned th,e applicant to determine if their methods appropriately bound ll future cycles. The applicant responded that the safety analysis is intended
[l to be valid for all plant cycles and the values of the input parameters used j
in the safety analysis are selected to bound the values expected in all
!:h ~
subsequent cycles. The applicant further stated that when all of the reload j
related parameters for a given accident are bounded, the reference safety
! ?
analysis is valid; however, if a parameter is not bounded, further evaluation l$
is necessary.
This fur,ther evaluation is to confinn that the margin of safety defined in the ' basis for any Technical Specification is not reduced.
lf Based on the information given above, the staff concludes that the applicant l
has adequately addressed our concerns on future cycle considerations.
ly The staff also requested that the applicant provide information as to I
whether there are p'lans to use the Westinghouse optimized fuel assembly or
)
Westinghouse improved thermal margin procedure as described in WCAP-8567 I
for Vootle. The applicant indicated that Vogtle currently has no intention of using the optimized fuel assemblies or WCAP-8567 in the initial core.
However, advanced fuels will be covered in subsequent refuelings and they will then amend the FSAR as necessary.
I Based on our finding that the CHF correlation and the thermal-hydraulic computer code used by the applicant have been previously approved by the staff, the applicant has appropriately bounded future cycles in his safety analyses, and the use of a unifonn core exit pressure gradient has been adequately justified, the staff concludes that the DNB design methodology used in the design of the Vogtle Units is acceptable.
4.4.3.2 Core Flow The core flow design basis requires that the minimum flow which will pass through the fuel rod region and be effective for fuel rod cooling is 95.5 l
l percent of the primary coolant flow rate. The remainder of the flow, p.
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called bypass flow, will be ineffective for cooling since it will take the
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following bypass paths:
4
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(1) flow through the spray nozzles into the upper head; i
(2) flow into the rod cluster control rod guide thimbles; (3) leakage from the vessel inlet nozzle directly to the vessel outlet nozzle; (4) flow between the, baffle and barrel; and (5) flow in the gaps 6etween the fuel assemblies.
The amount of bypass flow (4.5%) is determined by a series of hydraulic resistance calculations on the core and vessel internals and verified by model flow tests. Since the amount of bypass is consistent with approved plants of similar design, the staff concludes that th'e core flow given in the Vogtle FSAR, 95.5 percent, is acceptable.
4.4.3.3 Hydrodynamic Stability For steady-state, two-phase heated flow in parallel channels, the potential for hydrodynamic instability exists.
4 The applicant stated that the core design is stable since Westinghouse reactors will not experience any Ledinegg instability over Condition I and II operational ranges and open channel configurations, which are a feature of Westinghouse PWRs, are more stable than closed channel configurations. This was shown by flow stability tests which were conducted at pressures up to 2200 psia. The results-showed that for flow and power levels typical of power reactor conditions, no flow oscillations could be induced above 1200 psia.
n Also, a method developed by Ishii (Saha, et al,1976) for evaluating density wave stability in parallel closed channel systems was used to assess the stability of typical Westinghouse reactor designs. The results indicate that
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l a large margin to density wave instability exists. Finally, data from numerous rod bundle tests which were performed over wide ranges of
[
operational conditions show no evidence of premature'DNB or of inconsistent j
data which might be indicative of flow instabilities in the rod bundles.
i The staff is' conducting a generic study of the hydrodynamic stability of j
light water reactors.
Limitations to the thermal-hydraulic design
{
resulting from the staff study will be compensated for by appropriate operating restrictions; however, none are anticipated.
/
In the interim, the staff concludes that past operating experience, flow i
stability experiments and the inherent thermal-hydraulic characteristics of Westinghouse pressurized water reactors serve as a basis for accepting the Vogtle stability evaluation for issuance of an operating license.
4.4.4 Operating Abnormalities 4.4.4.1 Fuel Rod Bowing A significant parameter which affects the thermal-hydraulic design of the core is rod-to-rod bowing within fuel assemblies. The Westinghouse methods for predicting the effects of rod bow on DNB, WCAP-8691, Revision 1 " Fuel Rod Bow Evaluation," have been approved by the staff.
In response to a question, the applicant stated that Westinghouse-designed plants do not consider the effects of rod bow for an assembly average burnup greater than 33,000 MWD /MTU because beyond this burnup, the burndown effects preclude the fuel from achieving the limiting value of F4 H*
M For the worst case, at an assembly average burnup of 33,000 MWD /MTU, the calculated penalty is less than 3%.
However, sufficient margin (9.1%) is maintained to sustain full and low flow DNBR penalties.
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On the basis of its review of the infomation in the FSAR response to questions, and the fact that the methods used have been previously approved by the staff, t
{
the staff has concluded that the proposed rod bow calculations are acceptable.
The available thermal margin used to offset the rod bow penalty is required to be put in the bases of the Technical Specifications.
(
For plants designed by Westinghouse, the staff has approved the following generic margins, (" Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing on I
[
Thermal Margin Calculations for Light Water Reactors," December 1976), which may f
be used to offset the reduction in DNBR due to rod bowing as shown in Table 4.1.
e Table 4.2 Generic Margins o
L
% Reduction in j
Margin Rod Bow Penalties i
The use of a design minimum DNBR of 1.30 1.6 instead of the 95/95 DNBR limit'of 1.28.
A reduction in fuel rod pitch for the hot-1.7 channel analysis.
l The use of a thermal diffusion coefficient 1.2 p
(TDC) of 0.038 instead of a TDC of 0.051.
The addition of an extra grid in the design 2.9 of the Westinghouse 17x17 fuel assembly relative to the 15x15 fuel design.
The use of a 0.88 multiplier on the modified 1.7 spacer factor (Fs) of the W-3 correlation instead of a 0.865 multiplier.
[
Maximum generic margin which ndy be claimed.
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i 4.4.4.2 Crud Deposition s
Crud deposition in the core and an associated change in core-pressure drop and flow have been observed in scme PWRs not of Westinghouse design.
The applicant has stated that:
(1) operating experience on Westinghouse reactors 5
indicates that a flow resistance allowance for crud deposition is not required;
[
and (2) the effects of c.rud enter into the calculations by the use of a surface r l roughness factor three times greater than those obtained from operating
]
In' response to a staff question the applicant stated that t
the reactor coolant system flow rate will be measured prior to initial I
criticality, at 50 percent power, and at least once per 18 months. A I
surveillance test procedure will give guidelines for reactor coolant system flow measurement each 18 months and will utilize the same method of measure-f ment used in the startup test procedures.
In the event of a lower than l
design flowrate, the action to perform an engineering analysis and justify f
continued operation is acceptable.
The applicant should also provide
]
a description of the means of detecting reduced flow in between the 18 month I
flow measurement test periods as it has not been verified that venturi fouling will be adequately accounted for in determining the core flow rate.
This issue must be adequately addressed before the staff can approve the applicant's capability to measure core flow.
4.4.5 Thermal-Hydraulic Comparison i
The thermal-hydraulic design parameters for Vogtle Units 1 and 2 are listed in Table 4.2 and cmpared to the values of these parameters for the SNUPPS plants as presented on the Vogtle FSAR.
Vogtle Units 1 and 2 are designed to operate at the same thermal power as the SNUPPS plants. The W-3 CHF correlation and THINC-IV computer program were used in the design of both the Vogtle and SNUPPS plants.
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l Tablo 4.2 - Reactor Design Comparison f
Vogtle
(
Units 1 and 2 SNUPPS 1
Performance Characteristics Reactor core. heat output (MWt) 3411 3411
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System pressu' e, psia 2250 2250 r
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Departure from nucleate g
boiling ratio
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r Typical cell 2.07 2.07 Thimble cell 1.73 1.73 Minimum DNBR y.
1.30 1.30 i
Critical heat flux correlation W-3 W-3 f
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Coolant Flow 6
}
Total flowrate (10 lb/hr) 140.3 140.3 I
Effective flowrate for heat 6
transfer (10 lb/hr) 134.0 134.0 Average velocity along fuel rods (fps) 16.7 16.7 Average mass velocity 6
2 (10 lb/hr-ft )
2.62 2.62 Coolant Temperature, *F Nominal reactor iniet 558.4 558.4 Average rise in core 60.1 60.1 Heat, Transfer.100% Power Active heat transfer surface area, (ft )
5970b 59700 2
2 Average heat flux (BTU /hr-ft )
189800 189800 2
Maximum heat flux (BTU /hr-ft )*
436500 440300 Average linear heat rate (kw/ft) 5.44 5.44 Maximum thermal output (kw/ft) 12.5 12.6 This limit is associated with the value of F = 2.30 for Vogtle and q
F = 2.32 for SNUPPS.
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l[4 The differences between the Vogtl'e and SNUPPS plants as shown in Table 4.2 are slightly lower values for the Vogtle maximum heat flux and maximum linear thermal output. These differences between the thermal-hydraulic designs of the Vogtle and SNUPPS plants are negligible. The SNUPPS plant has been previously eviewed and approved by the staff. Although the Vogtle FSAR shows the thermal-hydraulic values of the Vogtle are very similar to those of SNUPPS, the current SNUPPS FSAR shows some other differences (e.g., coohnt flow and temperatures).
The staff requested infomation on these differences and will report on our findings in a supplement to this SER.
4.4.6 N-1 Loop Operation N-1 loop operation is when one of the reactors coolant loops is out of service.
Thus, only three coolant loops are available to supply coolant to the reactor core.
In response to a staff question, the apolicant stated that as a limiting condition for operation all reactor coolant loops shall be in operation during modes 1 and 2.
The staff will require that the Technical Specification include appropriate provision to ensure that N-1 type of operation is prohibited. The applicant should also state wh' ether or not there is any intention of operating in the N-1 mode in the future.
The staff will report on this in a supplement to the SER.
4.4.7 Loose Parts Monitoring System The applicant has provided a description of the Loose Parts Monitoring System (LPMS) which will be used by Vogtle Units 1 and 2.
This system is called a Metal Impact Monitoring System (MIMS).
The design will consist of twelve active instrumentation channels, each comprising a piezoelectric accelerometer (sensor), signal conditioning equipment and diagnostic equipment. Two redundant sensors are fastened mechanically to the reactor coolant system 11 ge
q f
-(RCS) at each of the following potential loose parts collection regions:
1.
Reactor pressure vessel-upper head region.
2.
Reactor pressure vessel-lower head region.
3.
Each steam generator-reactor coolant inlet region.
j The system will be capable of detecting a metallic loose part that weighs from O.25 to 0.30 pounds impacting within 3 feet of a sensor and having a kinetic energy of 0.5 foot pounds on the inside surface of the RCS pressure boundary.
The systen is designed to remain functional for a seismic event up to and including the operating basis earthquake (0BE).
The applicants response was incomplete and also took exception to some items in Regulatory Guide 1.133 for the LPMS. Therefore we requested more information (question 492.3). A response has not been provided.
We will require the licensee to provide a LPMS consistent with the provisions of Regulatory Guide g
1.133 and to commit to provide, prior to power operation, a final design report which contains the following:
(a) an evaluation of the LPMS for conform-ance to Regulatory Guide 1.133; (b) a desciiption of the system hardware, operation, and impiementat : an of the loose parts detection progra.d, including plans for startup testing, acquisition of baseline data, and alam settings; (c) a description and evaluation of diagnostic procedures used to confirm the presence of a loose part; and (d) a description of the operator training program.
The staff will report our findings in a future SER supplement.
4.4.8 Instrumentation for Detection of Inadequate Core Cooling The applicant's response to Question 492.3 with respect to the design require-ments stated in NUREG-0737. Item II.F.2 " Instrumentation for Detection of Inadequate Core Cooling (GCI)" is incomplete. Therefore,-we will require the 12 x.
,. = ~
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t
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applicant to provide the documentation required by Item II.F.2 of NUREG-0737 V
The ICC instrumentation consists of subcooling margin monitors, core exit thermocouple system, and reactor vessel level instrumenation system.
The
{
response should include evaluation of each ICCI component against the design f
requirements. The staff will report our findings in a future SER supplement.
(
- 4. 4.9 DNBR for Steam.Line Break-t
[
The RCS pressure during' the steam line break (SLB) accident presented in
[
subsection 15.1.5 of the FSAR shows that the pressure drops below the range 4
of pressure (1000 to 2300 psi) for which the W-3 correlation was originally developed.
The staff requested more information (question 492.9) to justify the W-3 correlation for the SLB. The applicants response was not extensive enough to resolve our concern.
The staff will report on the resolution of
(
this issue in a future supplement.
6 4.4 10 Conclusion c
The thermal-hydraulic design of Vogtle Units 1 and 2 was reviewed. The acceptance criteria used as the basis for our evaluation are set forth in the Standard Review Plan (SRP), NUREG-0800 in Section 4.4, Part II, " Thermal and Hydraulic Design Acceptance Criteria." The scope of the review included the design criteria, core design, and the steady-state analysis of the core thermal-hydraulic perfonnance.
The review concentrated on the differences between the proposed core design and those designs which have been previously reviewed and found acceptable by the staff.
It was found that all such differences were acceptable except as noted below.
The applicant's thermal-hydraulic design analyses were performed using analytical methods and correlations that have been previously reviewed by the staff and found acceptable.
The staff concludes that the initial core has been designed with appropriate margin to ensure that acceptable fuel design limits are not exceeded during steady-state operation and, anticipated operational occdrrences. The thermal-13 N *, *,......, -.
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a hydraulic design of the initial core, therefore, meets the requirements of 1
General Design Criterion 10,10 CFR Part 50, and is acceptable.
This a
conclusion is based on the applicant's analyses of the core thermal-hydraulic
[
performance which were reviewed by the staff and found to be acceptable. The applicant has committed to a pre-operational and initial startup test program 4
in accordance with Regulatory Guide 1.68 to measure and confirm the themal-l}
hydraulic design aspects. The staff has reviewed the applicant's pre-a l
operational and initial,startup test program and has concluded that it is
[I acceptable. However, pfior to issuance of an operating license, the staff l
will require the applicant to perfom the following:
t I
[
(1) address the concerns regarding flow measurement capability with crud f
buildup as described in Section 4.4.4.2 of this SER; f
}
(2) address the concern on thermal-hydraulic design comparison as described in Section 4.4.5 of this SER; 9
{
(3) address the concern regarding the N-1 loop operation as discussed in Section 4.4.6 of this SER; (4) address the concerns regarding the loose parts detection program as described in Section 4.4.7 of this SER; (5) supply the information for Item II.F.2 of NUREG-0737 as requested in Section 4.4.8 of this SER; and J
(6) address the concern regarding DNBR for a steam line break accident as described in Section 4.4.9 of this SER.
These issues will be addressed in a supplement to this SER.
14 i
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- = - - -
t Fr 15.4.1 Uncontrolled Rod Cluster Control Assemb1_y (Rod) Bar.k Withdrawal From Zero Power Conditions r
Discussion l
The consequences of an uncontrolled rod cluster control assembly bank withdrawal l
at zero power have been analyzed.
Such a transient can be caused by a failure of the reactor control rod control systems.
The analysis assumed a con-L servatively small (in a,bsolute magnitude) negative Doppler coefficient and a conservative moderator coefficient. Further, hot zero power initial conditions with the reactor just critical are chosen because they are knowi to maximize the calculated consequences. The reactivity insertion rate is assumed to be equivalent to the simultaneous withdrawal of the two highest worth banks at maximum speed (45 inches per minute).
Reactor trip is assumed to occur on the low setting of the power range neutron flux channel at 35 percent of full power (a ten percent uncertainty has been added to the setpoint value). The maximum heat flux is much less than the full power value and average fuel temperature incresase to a value lower than the nominal full power value. The minimum DNBR at all times remains above the limiting value of 1.30.
Evaluation Findings We have reviewed this event according to the Standard Review Plan (NUREG-0800)
Section 15.4.1.
The possibilities for single failures of the reactor control system which could result in uncontrolled withdrawal of control rods under low power startup conditions have been reviewed.
The scope of the review has included investi-gations of initial conditions and control rod reactivity worths, the course j
of the resulting transients or steady-state conditions, and the instrument response to the transient or power maldistribution.
The methods used to determine the peak fuel rod response, and the input into the analysis, such as power distributions and reactivity feedback effects due to moderator and s
fuel temperature changes, have been examined.
1 l
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I-l We conclude that the requirements of General Design Criteria 10, 20, and 25 s
)
have been met. This conclusion is based on the following:
The applicant has met the requirement of GDC 10 that the specified acceptable
{
fuel desgn limits are not exceeded, GDC 20 that the reactivity control systems i
are automatically initiated so that specified acceptable fuel design limits f
are not exceeded, and GDC 25 that single malfunctions in the reactivity control I
system will not cause the specified acceptable fuel design limits to be exceeded. These requircments have been met by compcring the resulting extreme I
operating conditions and response for the fuel (i.e., fuel duty) with the I
acceptance criteria for fuel damage (e.g., critical heat flux, fuel temperatures, 5
and clad strain limits should not be exceaded), to assure that fuel rod failure will be precluded for this event.
The basis for acceptance in the staff review
' s that the applicant's analyses of the maximum transients for single error i
i f
control rod withdrawal from a subcritical or low-power condition have been l
confirmed, that the analytical methods and input data are reasonably con-i servative and that specified acceptable fuel design limits will not be exceeded.
15.4.2 Uncontrolled Rod Cluster Control Assembly (Rod) Bank Withdrawal at power t
{
Discussion The consequences of uncontrolled withdrawal of a rod bank in the power operating range have been analyzed.
The effect of such an event is an increase in coolant temperature (due to the core-turbine power mismatch) which must be terminated prior to exceeding fuel design limits.
The analysis is performed as a function of reactivity insertion rates, reactivity feedback coefficients, and core power level.
Protection is provided by the high neutron flux trip, the overtemperaturea T and overpower A T trips, and pressurizer pressure and pressurizer water level trips.
In no case does the departure from nucleate boiling ratio fall below 1.30.
Adecpate fuel cooling is therefore main-tained. The maximum heat flux reached including uncertainties does not exceed 118 percent of full power, thus precluding fuel centerline melting.
2
.+
.=
x __ __
- I J
Evaluation Findings We have reviewed this event according to Section 15.4.2 of the Standard Review P1an (NUREG-0800).
The possibilities for single failures of the reactor control system which could result in uncontrolled withdrawal of control rods beyond normal limits under
{
power operation conditions have been reviewed. The scope of the review has i
included investigations, of possible initial conditions and the range of f
reactivity insertions,'the course of the resulting transients and the instru-
{
mentation response to the transient. The methods used in determine the peak
. l fuel rod response, and the input into the analysis, such as power distributions, i
rod reactivities, and reactivity feedback effects of moderator and fuel temper-ature changes, have. been examined.
1 j
We conclude that the requirements of General Design Criteria 10, 20, and 25 have been met. This conclusion is based on the following:
l The applicant has met the requirements of GDC 10 that the specified acceptable j
fuel design limits are not exceeded, GDC 20 that the reactivity control systems are automatically initiated so that specified acceptable fuel design limits are not exceeded, and GDC 25 that single malfunctions in the reactivity control i
system will not cause the specified acceptable fuel design limits to be exceeded. These requirements have been met by comparing the resulting extreme operatinp conditions and response for the fuel (i.e., fuel duty) with the I
acceptance criteria for fuel dmage (e.g., critical heat flux, fuel temperatures and clad strain limits should not be exceeded), to assure that fuel rod failure will be precluded for this event. The basis for acceptance in the staff review is that the applicant's analysis of maximum transients for single error control rod malfunctions have been confirmed, the analytical methods j
and input data are reasonably conservative and that specified acceptable fuel design limits will not be exceeded.
1 i
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9
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f 15.4.3 Rod Cluster Control Assembly Malfunctions
?
{'
Discussion i
Rod cluster control assembly misalignment incidents incl'uding a dropped full
[
length assembly, a dropped full legnth bank, a misaligned full length assembly and the withdrawal of a single assembly while operating at power have been
,1 f
analyzed by the applicant. Misaligned rods are detectable by:
(1) asymmetric j
power distributions sen, sed by excore nuclear instrumentation or core exit l.
thermocouples, (2) rod deviation alam, and (3) rod position indicators. A deviation or a rod from its bank by about 15 inches or twice the resolution
. l[
of the rod position indicator will not cause the power distribution to
'f exceed design limits. Additional surveillance will be required to assure rod alignment if one or more rod position channels are out of service.
In the event of a dropped assembly or group of assemblies the reactor will typically scram on a naturon flux negative rate trip, and analysis indicates that thermal limits will not be exceeded for the event.
If the rod locations i
are such that the reactor does not scram, however, the automatic controller may return the reactor to full power and the control could result in a power overshoot. An analysis methodology for this event has been developed by Westinghouse and reported in WCAP-10297-P, " Dropped Rod Methodology for Negative Flux Rate Trip Plants," January 1982.
This methodology has been reviewed and approved by the NRC staff. The review is in a memorandum for F. Miraglia from L. Rubenstein, " Review of the Westinghouse Report ' Dropped Rod Methodology for Negative Flux Rate Trip Plant'", December 1983.
Generally, detailed analyses for most reactors, for most cycles, show that if this event occurs thermal limits will not be exceeded. However, the analysis is reactor and cycle specific, and the analyses for Vogtle Units 1 and 2 for Cycle I have not been completed as yet. The staff has also accepted an interim position for operating reactors which consists of a restriction on operations above ninety percent power such that either the reactor is in manual control or rods are required to be out greater than 215 steps. This restrict, ion will be applied to Vogtle Units 1 and 2 in the event that calculations for Cycle 1 operation are not completed in time for init'ial operations. With this' restriction thermal limits wil? act be 4
w
~,
exceeded. Approval of the analysis specific to Vogtle Units 1 and 2 for I
Cycle I will result in removing the restriction. Similar analysis will also j
be needed for each subsequent reload cycle.
[
f For cases where a group is inserted to its insertion limit with a single rod f
in the group stuck in the fully withdrawn position analysis indicates that departure fran nucleate. boiling will not occur.
We have reviewed the calculated estimates of the expected reactivity and power distribution changes that accompany postulated misalignments of representative assemblies.
g We have concluded that the values used in this analysis conservatively bound the expected values including calculational uncertainties.
The inadvertent withdrawal of a single assembly requires multiple failures in the control systen,' multiple opgrator errors or deliberate operator actions y
combined with a single failure of the control system. As a result the single I
assembly withdrawl is classified as an infrequent occurrence.
The resulting f
transient is similar to that due to a bank withdrawal but the increased h
peaking factor may cause departure fran racleate boiling to occur in the L
[
region surrounding the withdrawn assc,nbly.
Less than five percent of the
[
rods in the core experience departure from nucleate boiling for such a L
)
i Evaluation Findings We have reviewed this event according to Section 15.4.3 of the Standard Review Plan (NUREG-0800).
The possibilities for single failures of the reactor control system whis could result in a movement or malposition of control rods beyond normal limits have been reviewed. The scope of the review has included investigations of possible rod malposition configurations, the course of the resulting transients or steady-state conditions, and the instrumentation response to the transient or power maldistribution.
The methods used to det. ermine the peak fuel rod response, and the input t,o that analysis, such as power distribution changes, rod reactivities, and reactivity feedback effects due to moderator and fuel l
tanperature changes, have been ' examined.
5 g.
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We conclude that the requirements of General Design Criteria 10, 20, and 25 5
have been met. This conclusion is based on the following:
I The applicant has met the requirements of GDC 10 that the specified accept-able fuel design limits are not exceeded GDC 20 that the reactivity control systems are automatically initiated so that specified acceptable fuel design
{
limits are not exceeded.,and GDC 25 that single malfunctions in the
)
reactivity control system will not cause the specified acceptable fuel design l
limits to be exceeded. ' These requirements have been met by comparing the f
resulting extreme operating conditions and response for the fuel (i.e., fuel duty) with the acceptance criteria for fuel damage (e.g., critical heat flux, fuel temperatures and clad strain limits should not be exceeded), to assure that fuel rod failure will be precluded for this event. The basis for acceptance in the staff review is that maximum configurations and transients for single error control rod malfunctions have been analyzed, that the analysis methods and input data are reasonably conservative and that specified acceptable fuel design limits will not be exceeded.
15.4.7 Inadvertent Loading of a Fuel Assembly into Improper Position I
Discussion Strict administrative controls in the form of previously approved established procedures and startup testing are followed during fuel loadings to prevent operation with a fuel assembly in an improper location or a misloaded burnable poison assembly.
Nevertheless, an analysis of the consequences of a loading error has been performed.
Comparisons of power distributions calculated for the nominal fuel loading pattern and those calculated for five loadings with misplaced fuel assemblies or burnable poison assemblies are presented by the applicant. The selected non-nonnal loadings represent the spectrum of potential inadvertent fuel misplacemente Calculations included, in particular, the power in assemblies which contain provisions,for monitoring with incore detectors.
6 q5
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f.
As part of the required startup testing, the incore detector system is used to detect misloaded fuel prior to operating at power. The analysis described above shows that all but one of the above misloading events would be detected by this test.
In the excepted case, an interchange of Region 1 and 2 assemblies l
'near the center of the core, the increase in the power peaking is approximately equal to the uncertainty in the measurement of this quantity (5 percent). This f
uncertainty is allowed for in analyses so that this misloading event does not i
result in unacceptable, consequences.
Evaluation Findings j
We have reviewed this event according to Section 15.4.7 of the Standard Review i
P1an (NUREG-0800).
l We have evaluated the consequences of a spectrum of postulated fuel loading i
errors. We conclude that the analyses provided by the applicant have shown i
a for each case considered that either the error is detectable by the avail-3 able instrumentation (and hence remediable) or the error is undetectable but the offsite consequences of any fuel rod failures are a small fraction of 10 CFR Part 100 guidelines.
The applicant affims that the available incore instrumentation will be used before the start of a fuel cycle to search 'or fuel loading errors.
We conclude that the requirements of General Design Criterion 13 and 10 CFR Part 100 have been met. This conclusion is based on the following:
4
'l The applicant has met the requirements of GDC13 with respect to providing l
adequate provisions to minimize the potential of a misloaded fuel assembly going undetected and meets Part 100 with respect to mitigating the g
consequences of reactor operations with a misloaded fuel assembly. These requ'irements have been met by providing acceptable procedures and design features that will minimize the likelihood of loading fuel in a location other than its designated placa.
1 i
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15.4.8 Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster
[
Control Assembly Ejection Discussion The mechanical failure of a control rod mechanism pressure housing would result in the ejection of a rod cluster control assembly.
For assemblies initially inserted, the consequences would be a rapid reactivity insertion i
together with an adverse core power distribution, possibly leading to localized fuel rod damage. Although mechanical provisions have been made l
to make this accident extremely unlikely, the applicant has analyzed the consequences of such an event.
Methods used in the analysis are reported in WCAP-7588, Revision 2 "An Evaluation of the Rod Ejection Accident in Westinghouse Reactors Using Spatial Xinetics Methods," which has been reviewed and accepted by the staff. This report demonstrated that the model used in the accident analysis is conservative relative to a three dimensional kinetics
?,-
calculation.
The applicant's criteria for gross damage of fuel are a maximum clad temper-ature of 2700 degrees Fahre'nheit and an energy deposition of 200 calories per gram in the hottest pellet. These criteria are more conservative
- than those proposed in Regulatory Guide 1.77. Therefore, they are acceptable.
i Regulatory Guide 1.77 he.: en acceptance criterion of 280 calories per g
gram energy depcation and no criterion for clad temperature other than that implicit in requirements for fuel and pressure vessel damage.
t, a
e 8
4 lEy.
.. -. :. ::-- = =.:=== :=.= :............
,I t-l; Four cases were analyzed: beginning-of-cycle at 102 percent and zero power i, -
and end-of-cycle at 102 percent and zero power. The highest clad temperatures, I
2426 degrees Fahrenheit, and the highest fuel enthalpy,172 calories per gram,
[
were reached in the beginning-of-cycle zero power and beginning-of-cycle full
[
power cases respectively. The analysis also shows 'that less than 10 percent of the fuel experiences departure from nucleate boiling and less than 10 percent cf the hot pellet melts. Analyses have been performed to show that
!j the pressure surge prod,uced by the rod ejection is mild and will not approach the Reactor Coolant Sys' tem emergency limits.
Further analyses have shown that a cascade effect, i.e., the ejection of a further rod due to the ejection of the first one, is not credible.
i Evaluation Findings We have reviewed this event according to Section 15.4.8 of the Standard Review j
P1an (NUREG-0800).
t l
We conclude that the analysis of the rod ejection accident is acceptable and meets the requirements of General Design Criterion 28. This conclusion is based on the following:
The applicant met the requirements of GDC 28 with respect to preventing postulated reactivity accidents that could result in damage to the reactor coolant pressure boundary greater than limited local yielding, or cause sufficient damage that would significantly impair the capability to cool the core. The requirements have been met by demonstrating that the regulatory positions of Regulatory Guide 1.77, " Assumptions Used for Evaluating a Control Rod Ejection Accident for PWR's" are complied with.
The staff has evaluated the applicant's analysis of the assumed control rod ejection accident and finds the assumptions, calculation techniques, and consequences acceptable.
Since th'e calculations resulted in peak fuel enthalples less than 280 cal /gm, prompt fuel rupture with consequent rapid heat transfer to the coolant fron finely dispersed molten U0 was 2
assumed not to occur. The pressure surge was, therefore, calculated on the basis of conventianal heak transfer fran the fuel and resulted in a pressure increase below " Service Limit C" (as defined in Section III, " Nuclear Power 9
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Plant Components," of the ASME Boiler and Pressure Vessel Code) for the maximum control rod worths assumed. The staff believes that the calculations
[-
contain sufficient conservatism, both in the initial assumptions and in the 1
analytical models, to ensure that primary systen integrity will be maintained.
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December 13, 1984 -
Docket Nos: 50-424 and 50-425 MEMORANDUM FOR: Elinor G. Adensam, Chief
' Licensing Branch No. 4
/
Division of Licensing
'FROM:
Melanie-A: Miller, Pro]ect Manager ~
Licensing Branch No. 4
~
Division of Licensing
'y
SUBJECT:
NOTICEOFFORTHCOMINGMEETINGONV0GTLEGEb7ECHNICALREVIEW DATE&TkNE:
January 23, 1985 8:00 a.m. - 4:00 p.m.
LOCATION:
RRC offices, P-110 7920 Norfolk Avenue Bethesda, MD PURPOSE:.
To discuss open items on Vogtle in the geotechnical area as ll identified in DSER dated November 6,1984.
PARTICIPANTS:
NRC Southern Company Services M. Miller M J. Bailey g.
ne l{gnK.Kopecky Beck.,p Georgia Power Company S. Cereghino D. Hudson DN W. Ferris SdA. % %
v GMt. M. Perovich 6,l it..J.
7 g {g kig l
h Z. Yazadani i
(dt. h%b 6d- -hadg g
p TomCrag.hudh.nMj U
Melani.e A.' Miller, Project Manager
[
Licensing Branch No. 4 Division of Licensing i
s cc:
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ipril 19,1984 DISTRIBUTION:
Docket File 50-424 CSB Rdg. File CLi JShapaker WButler MEMORANDUM FOR:
E. Adensam, Chief Licensing Branch No. 4. DL FROM:
W. Butler, Chief Containment Systems Branch, DSI
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION: V0GTLE ELECTRIC GENERATION PLANT (DOCKET NO.: 50-424)
Plant Name: Vogtle Electric Generation Plant Docket No.: 50-424 Responsible Branch: LB No. 4. DL Project Manager:
M. Miller Review Branch: CSB Review Status: Incomplete
.The enclosed Request for Additional Information (RAI) for the Vogtle plant has been prepared bytthe Containment Systems Branch after having reviewed the appropriate sections of the FSAR. The RAI numbering sequence is a continuation of that used in the OL application acceptance review.
The governing SRP sections are identified in parantheses below the item number for each RAI.
CriS nal signed by i
yialterR. Butler W. Butler, Chief Containment Systems Branch, DSI
Enclosure:
As stated cc:
M. Miller R. Mattson R. W. Houston
Contact:
C. Li, CSB
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!O F@o u DtO 810 8 0s t.R C W C 2 d C OFFICIAL RECORD COPY W u 5 Cm "*3"co-
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_ _-- _x u
-m ENCLOSURE i
4 REGUEST FOR ADDITIONAL;INFORMATION CONTAINMENT SYSTEMS BRANCH
,s-480.9 Revise Table 6.2.1-1 (Containment Peak Pressure and l
(SRP 6.2.1). Temperature) to sinclude the containment design pressure' requirements (including both internal and 3
external design pressure) and'!the containment design temperature.
480'.10
, Revise Table 6.2.1-3 'to incLdde two conditions of
- (SRP 6.2.1) operation for the fan coolers and' containment t t i-
~
sprays, namely full capacity and the capacity used in the containment analysis.
l_
480.11
- The init'ial containment p'ressures for the peak l
(S RP 6.2.1) ', cont ai nmen't pressure analysis, minimum containment pressure analysis 3and subcompartment analysis are L
N5 15.0,14 7, and 13.2 psia, respectively.
assumed to be 3
e, Discuss and jus,tify thy differences.
480.12 T a b'l e 6. 2.1 - 8 Lists the calcutated maximum pressure (SRP 6.2.1) dif f erentials for subcompa rtment analyses.
Discuss the design ba sis, f or the subcompartment walls, and the adequacy of the structural design margins.
480.13 Identify the source of the mass and energy (SRP 6.2.1) release data Listed in Tables 6.2.1-26 through 6.2.1-74 for the subcompartment analysis, and the
'i u approvat status of the methodology (e.g., topical report).
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480.14 Provide an analysis of the forces and moments (S RP 6.2.1-) acting on the reactor vessel due to the differential pressure across the vessel caused by a reactor coolant system break within the reactor cavity.
The guidelines of SRP 6.2.1.2 and Section 3.2 of NUREG-0609 should be followed.
480.15 Provide additional information and/or analysis to (SRP 6.2.1) resolve the concerns of IE But Letin No. 80-04 regarding main steam Line breaks with continued feedwater addition.
Discuss whether the MSLB analysis included the impact of other energy sources, such as a continuation of feedwater or condensate flow.
Discuss the ability to identify and isolate the damaged steam generator and the' capability of the pumps to remain operable after extended operation at runout flow.
480.16 Provide a figure showing the transient energy (SRP 6.2.1) distribution (energy balance) in the containment, including the ener'gy inventories of the containment atmosphere, sump water and structures, and the energy removal from the containment syst'em for the worst case LOCA.
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w 3-480.17 Extend the MSLB. results shown in Figures 6.2.1-27 and (SRP 6. 2,.1 ) 6.2.1-28 beyond 1800 s econds, to about 10,000 seconds, to assure that the peak containment p re s s u re has been reached and to provide a longer term containment temperature profile 480.18 Provide additional information on the net positive (SRP 6.2.2) suction head (NPSH) analysis of the spray pumps during the recirculation phase, in sufficient deta'1L, to permit the staf f to assess the adequacy of the a na ly s i s.
Explain ho'W the results shown in Figure
- 6.2.2-4 were obtained.
Provide the numerical values of each term in the NPSH equation shown in Section 6.2.2.2.2.3.2 and' the basis for these values.
480.19 Table 6.2.2-2, Containment Fan Cooling Heat Removal (SRP 6.2.2) Capacity, indi cates the data is used for the MSLB accident.
Discuss the applicability of the data to the LOCA analysis.
If it is'not applicable, provide a similar f an cooler heat removal capacity table for LOCA c o n s i de r a t i,o n.
480.'20 In the NPSH calculation, assuming a containment (SRP 6.2.2) sump fluid temperature of 212 F is inconsistent.with Regulatory Guide 1.1.
The maximum expected temperature of the pumped fluids should be assumed.
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. 480.21 The containment spray system is designed to be (SRP 6.2.2h manually switched f rom the injection to the
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recirculation mode.
It is our understanding that the operator initiates switchover of the containment spray system af ter completing ECCS switchover; the switchover operation is initiated upon receipt of the RWST Low-Low level alare.
Provide additional
,inf ormation rega rding the operator actions requi red in the swit chover of the water source from the injection to the recircu'Lation mode for containment spray system operation.
Justify that adequat e time wi LL be available for carrying out these actions.
480.22-FSAR Section 6.2.4.3 states that the 24-inch preaccess (SRP 6.2.4) purge Lines are only opened in the cold shutdown condition.
NUREG-0737 at It em II.E.4.2 recommends that purge valves be seated closed during operational modes 1, 2, 3, and 4.
Furthermore, these valves should be verified closed at least every 31 days.
Confirm that the 24-inch purge Lines witL be sealed closed and subject to the prescribed surveillance.
Discuss and justify h ow this vi LL be accomplished.
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480.23 The containment isolation provisions for each fluid (SRP 6.2.4) Line penetrating containment must conform to the requirements of General Design Criteria 54, 55, 56 or 57, as appropriate.
Those containment penetrations whose isolation provisions do not satisfy the explicit requirements of the General Design criteria but which
/
are acceptable on some other defined basis should be discussed Line by Line with the deviation identif!ed and the specif.ic "other defined, basis'" Justified.
Provide this information for staff' review.
480.24 Confirm that at L fluid Lines penetrating containment Listed in Tab' e 6.2.4-1, with' the isolation valves (SRP 6.2.4) are L
identified (include test, vent and drain connections).
Provide justification for each containment isolation valve that vill not be Type C (i.e., local f' Leak rate) tested.
480.25 The purge and ver?. system debris screens should (SRP 6.2.4) satisfy the fotLowing criteria:
a.
The debris screen should be seismic Category I design ahd installed about one pipe diameter away from the inner side of the inboard isolation valve.
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Th e piping between the debris screen and-the isolation valve should also be sei smi c Category I design.
c.
The debris screen should be designed to withstand the LOCA generated differential pressure.
Discuss and justify how the VEGP purge and vent system debris screens meet the above '
cr'iteria.
480.26 As shown in Figu re 6.2'.4-1, penetration numbers (SRP 6.2.4) 59 and 60 have isclation valves inside the containment but do not have any valves outside the co nt ai nme nt.
Justify the isolation provisions for containment penetration numbers 59 a,nd 60 relative to GDC 55 isolation valve requirements.
480.27 As shown in Figure 6.2.4-1, penetration number (SRP 6.2.4) 87 does not have any isolation valve inside the containment.
Table 6.2.4-1 indicates that penetration numb,er 87 meets the requi rements of GDC 56.
Justify the isolation provisions for penetration number 87 relative to GDC 56 containment isolation requirements.
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480.28 It is recommended in Regulatory Guide 1.7 that the
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(SRP 6.2.5) containment combustible gas control' systems be deskened, fabricated, erected,andtestedtothe Group B quality standards of R.G.
1 '.' 2 6.
Table 3.2.2-1 indicates that the hydrogen recombiner
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and hydrogen monitoring systems are not so classified.
Discuss your plans for complying with this staff position.
480.29 SRP Section 6.2.5 re commends that the fission (SRP 6.2.5). product decay energy used in the calculation of' hydrogen f rom radiolysis of the emergency core cooling water and sump water is acceptable if it is equal to or'more conservative than the decay energy model given in Branch Technical Position ASB 9-2 i n SRP s ection' 9.2.5.
Discuss and compare the decay energy model used in the FSAR Section 6.2.5.3.1.2 with the one in SRP 9'. 2. 5.
480.30 The post-LOCA cavity purge system is designed to (SRP.6.2.5) prevent hydrogen pocketing in the reactor cavity f o llowi ng a LOCA.
L iscuss and justify the need for this systes.
Discuss the performance
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. l 480.31 Discuss and justify how operation of Hydrogen
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monitoring system is initiated.
If it is done i
automatically, describe the initiation signals.
If it is done manually, describe the procedure required.
480.32 SRP 6.2.6 provides detailed guidance on how (SRP 6.2.6) instrument Lines penetrating containment should be treated during the conduct of the containment integrated Leak rate test (CILRT).
The following instrument Lines are of concern:
Penetration Numbers 13C, 67C, 69C, 70C, 71C, and 85C.
Discuss how the p'otential Leakage contribution of these Lines wi t L be included in the CILRT.
j 480.33 FSAR Section 6.2.6.3 states that Type C testing of (SRP 6.2.6) the safety injection Lines, containment sp ray lines, and Long term re ci rculation Lines wiLL not be done on the basis that these Lines are water-seated.
Additi ona l j ust i.f i cation is needed for the elimination of Type C tests (note that Table 6.2.4-1 indicates l
Type C testing f or the spray L ine s) :
s.
For ea ch Line, discuss and justify that a sufficient water inventory wilL be availaote for at least 30 days f ollowing a LOCA.
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b.
For each line, discuss your' plans for hydrostatif calLy testing the valves to show that water Leakage f rom the isolation valves is compatible with'the 30-day inventory requirement.
The Leakage Limits for these y
valves should be included in the plant Tec hni ca l S pe cifi cations. '
c.
FSA'R Section 6.2.6.3 st at es that the isolation valves in the charging Line of the chemical and 2
volume control system are Type C tested using 4~
water.
Type C testing using water as the test f l u i d i s p e r m i s s ab l e Nov.ey erAewaW-
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b p 6 if it can be shown that pa rts
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a, and b. are satisfied.
s 480.34 FS AR Section 6.2.6.4 ref ers to Section 6.2.6.2 (SRP 6.2.6) regarding the periodic testing i nt e rva ls of the c ont ai nm en t hat ches.
It is not c le a r in S e ct io n 6.2.6.2 t h'd t the testing intervals meet the requirements specified in Appendix J to 10CFR50.
C la ri fy the statements in the FSAR to explicitly comply with Appendix J requirements, or identify and justify the differences.
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, 480.35 The cont a inment spray system i s a safety related (SRP 6.2.2) system and should be Q-Listed.
As shown in Table 3.2.2-1, sheet 13, item numbers 12, 13, 14, 15, 17, and 18, portions of the cont ai nment spray system are neither safety related nor Q-Listed.
Explain and justify, or correct the table.
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