ML20039G977
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OBattelle Pacific Northwest Laboratories P.O. Box 999 Richland, Washington U.S.A. 99352 Telephone (509) 375-2455 Telex 15-2874 October 20, 1980 Mr. S. Gupta Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comission Washington, D.C.
20555
Dear Mr. Gupta:
l I have reviewed the B&W documents on the LYNX-1 and LYNX-2 codes
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(BAW-10129 and BAW-10130) about as much as possible without some further l
information. The explanations are really quite sketchy and I can make no firm conclusions from them.
To complete the review I need to get the following items:
The B&W TEMP code document BAW-10021 (April 1970)
The B&W CHATA code document BAW-10110 (January 1976)
Response from B&W to the questions listed on the attached sheets.
Thank you for your assistance.
Sincerel,'
f i
C. W. Stewart Senior Development Engineer l
CWS: pas
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8201190368 810403 PDR FOIA MADDEN 80-515 PDR
OUESTIO!is ON LYNX-1 (BAW-10129) 1.
In the energy equation (2-2) on page 2-2 how is the entnalpy of crossflow, hjj, computed?
2.
In the lateral momentum equation (2-4) on page 2-4 how is the axial velocity, Vjj, computed? Also, what values are typically used for s/t and K for bundles in equation (2-4)?
3.
Paragraph 2.5 implies that the fluid density is a function of the local pressure rather than a unifonn system pressure as in COBRA.
Is this true? If so, are the saturation properties also computed as functions of local pressure? Page 3-5 mentions system pressure so the issue is unclear.
4.
It appears that, except for the two-phase multipliers, there is no provision for two-phase slip in the momentum equation (3-10) on page 3-3.
Why isn't the definition of specific volume fcr momentum applied as is done in LYNX-27 5.
Paragraph 3.3 states that the inlet pressure profile is improved based on the difference between the calculated and specified exit pressure profiles. How is this done? What is the new value of local pressure obtained? What is the difference in the solution between cases where inlet flows are specified and those where inlet pressures are given?
6.
Is the crossflow solution on page 3-5 solved iteratively or directly?
7.
The energy equation (3-8) is implicit (contains h on both sides) i2 and should be solved simultaneously for all channels at each level to conserve energy properly. How is this done?
8.
Item 5 on page 4-1 states that LYNX-1 is not designed for subchannel analysis. What is considered the limiting factor?
9.
What are the essure drop and assembly lift modifications to the fo,n loss coefficients mentioned on page 5-6? They are not discussed or references elsewhere.
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- 10. For the COBRA-3C comparison discussed on page 7-3, did B&W run COBRA-3C separately or were the results taken from Sutey and Rowe's report? If B&W ran the code, what input options and parameters were used?
- 11. The flow regimes for heat transfer discussed on page A-2 mention bulk boiling and superheated regions but no transition boiling or film boiling. Why aren't these parts of the boiling curve included?
- 12. Baker's flow regime map (page A-3) was-developed for horizontal i
flow. What is the qualification for its use in vertical flow?
- 13. On page A-3, what friction multiplier is used for grids? Also, how is the effect of two-phase flow accounted for in acceleration pressure drop?
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4 QUESTIONS ON LYNX-2 (BAW-10130)~
1.
How are values determined for the factor f in equation (2-2)?
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2.
Is the definition M =MlM in the footnote on page 2-4 used for both momentum flux and drag terms?
3.
Why are derivatives of V' with respect to pressure and enthalpy included on page 2-57 Is the density computed as a function of local pressure or system pressure? Are saturation properties computed with local or system pressure?
4.
Why is the degenerate crossflow equation (2-23) used in LYNX and not LYNX-l?
5.
Baker's chart (page 2-10) was developed for horizontal flow. What is the justification for its use in vertical subchannels?
6.
Romies' form loss multiplier (page 2-16) was developed for sudden expansions in a pipe. What is the justification for its application to-grids?
7.
Why not use a boundary value solution like LYNX-1 and COBRA-3C7 8.
What is the actual form of the set of equations (3-5)? Please show the actual difference equations.
9.
It is not clear what is being done in section 3.4 on transverse exchange. Please explain the technique in terms of how interbundle exchange is incorporated into the subchannel equations.
- 10. What provision is made for reverse flow in the momentum and energy equations?
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- 11. What is the friction reduction factor (SPACER) on the flow chart i
i on page 6-37 It isn't mentioned in the text.
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3/19/77
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g Docket Nos.: 50-329/330 f
T. 1 % y a.C.
P Memorandum FOR: -ik 9: 4issal-lo, A/f for LWRs, DPM P,_L._Jedeycop./4-Soc RSv 056 t. 6. (ecsec P l b C -> C. A C., W L FROM:
MIDLAfiD 1 AND 2 SER INPUT FROM -f,*b:!" SIS-BRANCH, DEN b s
SUBJECT:
C o t2 t R t # men te Plant Name: Midland I and 2 Licensing Stage: OL Milestone No.: 24-2 Responsible Branch and Project Manager: LWR-4, D. Hood DSS Branch Involved: Analysis Branch Description of Review: SER Input Requested Completion Date: 4*rh4d3 Status of Review:
Incomplete M L eh-c e 8*
The'Analys'.s Branch input top', Midland 1 and 2 SER is enclosed. The 4
input includes Section 4.4 (Thermal and Hydraulic Dasign)dectica-t
.6.2 (Containment Design Bast'LJCA Calculations) ah Sec.tj6E15.2.
is (AMlytical Techniques)t-Section 4.4 has been reviewed to determine the adequacy of the methods used in the thermal-hydraulic design of the Midland plant. The SER attached identifies a number of topical reports for which the W
I staff revief is incomplete. Specifically, these are -Reactor. Vessel r
-Model-Flow-Tests -(BAW-10037 Rev;2)g CHATA (BAW-10110), LYNX-1 (BAW-10129),HYTRAN(BAW-10109), POWERTRAIN(BAW-[10070),and TRAP-2 (BAW-10128). Our review of the topical reports will be addressed in a supplement to the Safety Evaluation Report.
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methodstcp(aket1emap ly(e to 'i nd. He have l to the
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l conci ed that the mass amd energy c ease ao e conscrvat"ve
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for t ie containmpt'analv ', of Midla%
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1 rom the % applicant r\\ J nticipa /te2d6tjoba informa tion We egarding
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nal' is of a b p ning of life st anriine break and the feed-water line break (Section 15.2).
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C:Al.YSIS Li.X.CH MIDLAND 1 and 2 SER II;PUT Decket Nos.
50-329/330 4.4 Thermal and ifydraulic Design The principal criterion for the thermal-h dr aulic design of a reactor i
is avoidance of thereally induced fu61 damage during noir.a1 steady-state operation and during anticipated operational occurrences. :.:$nndIand.
2 uses the following design limits to satisfy this erfiterion:
- 1. The fuel pin cladding, fuel pellets, and fuel-to-clad gap characteristics ensure that the maximum fuel temperature does not exceed the fuel melting limit at the 112 percent design overpower at any time during core life.
The fuel mc1 ting tcmperature is 5080 F at BOL and reduces Ifncarly to 4800 F at EOL ( 43,000 MND/ NTU ).
- 2. The minimum allowable departure for _ nucleate boiling ratio during steady-state operation and anticipated transients is 1.30-with the BAN-2 correlation.
- 3. Flow stability is required during all steady-state and operationd transient conditions.
The thermal and hydraulic design parameters for the reactors are listed I
in Tabic 4./.
A comparison of the[se parameters with thos'e of Three Mile /. Idland 2 is given in the table.. The design of the Three Mile F
el us.
Idland reactor has previouby been found acceptable by the-stafd(
The comparison indicatch that the h2aland bermal-hydraulic design par -
meters are within previously accepted design values.
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The core hydraulic analysis for Midland 1 and 2 is based on design data and the results of vessel model flow tests l
for a 177 fuel assembly design. The applicant has stated that, based on flow model tests described in BAW-10037,0 Revision 2, no fuel assembly in the interior region of the core will receive less than 95% of the average coolant flow rate under four loop operation. The staff concurred with B&W on the approach used to verify the core flow dis-tribution in connection with our review of the Oconee Nuclear Station Unit 1 (Reference 1). The modification to ti.c core to remove the orifice rod assemblies and add retainers to the burnable poison rod assemblies has the effect of increasing the flow rate; however, the slight change in flow distribution is considered to be minimal and the staff conclusion is unchanged.
Prevention of departure from nucleate boiling (DNB) for steady j
state operation and anticipated transients will assure that the hot spot of the fuel cladding is at a temperature only slightly greater than that of the coolant and that the fuel cladding will maintain its integrity. The peak power density that will occur for a re,eactor trip at 112% maximum overpower trip is 16.6 kilowatts per foot. At this linear heat generation rate, B&W calculated a centerline temperature of appproximately 4250 F at beginning of life, thus indicating.no fuel melting.
The margin to DNB 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 kat flux required to produce DNB at the calculated local conditions to the actual heat flux.
The DNB correlation used for the design of the Midlane cores is the BAW-2' correlation. The BAW-2 correlation was derived from data on six-foot heaters which simulated the rod diameter and spacing of 15 x 15 fuel assemblies. Tests were conducted on uniformly heated bundles and non-uniformly heated tubes and annuli. The BAW-2 correlation with a minimum DNBR of 1.30 has been-aph
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-is therefore accptstle for ujp in the Midland design.
Rod-to-rod bowing within fuel assu;.blics is accounted for in the Midland core design by a D:lB penalty applied to the hot channel.
The applicant has provided an acceptable currelation for fractional clearance reduction due to rod bow for B&W I*. ark B 15x15 fuel rods.
Prior to issuance of the operating license for Midland 1 and 2, the staff will ensure that appropriate provisions for accordating the reduction Luc bu..inur a41 in thermal r.argin due to rod bow in the Tecyanical Specifications.
5 Parallel channel ffow stability analysis' for Midland 1 and 2 was performad with the HYTRN1 computer code.) The HYTP.AN code predicted stable flow during steady state operation and operational transients.
The HYTRAN topical report (B&W-10109) is presently under review in connection with the staff generic study of the hydraulic stability characteristics of pressurized water reactors, including the evaBuation of analytical methods used for Midland 1 and 2.
Any limitations to the thermal-hydraulic design resulting from the staff study will be compensated for by appropriate operating restrictions.
In the interim, the staff concludes that past operating experience, flow stability experiments and the inherent thermal-hydraulic characteristics of light water reactors provide a basis (Q for accepting the Midland stability evaluation for nofnal operation an nticipated transient events.
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The applicant has provided a description of thefoose 3rts s
ptlbnitoringSystemtobeprovidedforliidlandIand2.
The design will include two sensors at each selected natural collection region.
The sensors will be stud nounted on the reactor vessel and steam generators.
The system will be capable of detecting a loose part weighina from 0.25 to 30 pounds and impacting within 3 feet of a sensor.
The applicant has stated that the system will be designed to remain operational for all seismic events up to and including the operating basis earthquake, and has stated that the components within containment are to remain operable under normal environmental conditions of f I
the plant.
The loose parts detection program description provided in the 111dland FSAR follows the cuidelines of Regulatory Guide 1.133 (draft) and is acceptable to the staff.
g 4t pSv.4.kc e staff requires verification of system transient codes used i e th nalyses of transients reported in Section 15 of all FSAR The ej, type o system transients applicable to the checkout of tr sient f
codes are illneated in Section 5 (Paragraphs h-h to n-of Regula-tory Guide
, Revision 1, January, 1977 concerni, power ascen-sion tests foll ing the pre-operational and low,p wer tests conducted price to mmercial operation.
The, transient codes used by B&W for Midland 1 2 are TRAP and POPERTRAIN.
ig AP h-A l'Our review of the Gw, !.e 1 f
to the point that there is reafonableOWERTRAIN code [ has progressed m
will not be appreciably alt (red by any surance that analyses results 1
. hod revision that may be required by the staff.,)!6 will require con irmatory tests in support of the codes for ver}fication.
The test prog m must be submitted
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for NRC review.
or5e of these tests will be coi ucted in the pre '
program, and the remainder will(Dgperfomad with I
operational te a specified J el of power.
We will require cormitm' t that the i
needed da a'and test results, obtained with proper ins umantation, will b ubmitted to the NRC and will also be used by th =pplicant to cp firm the pretest predictions by the codes and other p i
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)ich have been used for the pretest prediction.
The results f grams it 6ur completed review will be applicable to this plant?
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The thermal-hydraulic design analysis fer Midland 1 and 2 assumes a uniform core outlet pressure distribution.
In response to a question on radial pressure gradient, the applicant referenced a comparative analysis using the closed channel computer code (CHATA) and the cross-flow computer code (LYfiX 1). The results of the analytit comparison indicated an increased ma,dgin with the use of the LYf(X 1 code. Our review of CHsff and LyflX 1 is incom-plete at this time and will be addressed in a supplement to the Safety Evaluation Report. Our review of these codes has pro-gressed to the point that there is reasonable assurance that the conclusions based on these analyses will not be appreciably altered by completion of the analytical review.
Crud deposition in the core and an associated change in core pressure drop and flow have been observed in some pressurized reactors.
The applicant has stated that (1) operating experience fem other Babcock and Wilcox reactors indicated very low levels 4
of crud buildup in the core; (2) sufficient design margins are included in the methods employed for predicting fuel performance to account for crud buildup; and (3) reduced flow would be observed by coolant flow monitors and periodic heat balance pro-gram analyses. The staff has reviewed this information and we conclude that it adequately addresses our concerns relative to uniform or preferential crud deposition in the core.
The proposed Technical Specifications for Midland 1 and 2 are being reviewed to assure that appropriate consideration has been given to detection and action relative to significant crud deposition.
He conclude that, with the exceptions noted, the thermal-hydraulic design of Midland 1 and 2 conforms to the Commission's regulations and to applicable Regulatory Guides and Staff technical positions and is acceptable.
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TABLP 4.1 THERMAL-HYDRAULIC DESIGN COMPARISON _
0F MIDLA!!D 1 AfiD 2 AfiD THREE MILE ISLAfiD 2_
MIDLAt1D THREE MILE 1 AND 2 ISLAND 2_
2452 2772 Design Core Heat Output, Megawatts thermal 2200 2200
!!ominal System Pressure, psia Vessel Coolant Inlet Temperature, F 554.9 557.
F 603.1 607.7 Vessel Coolant Outlet Tcmperature, O Total Core Heat Transfer Area, ft 49,734 49,734 2
2 163,684 185,090 Average Heat Flux, Btu /hr-ft 2
437,036 494,190 Maximum Heat Flux, Btu /hr-ft 5.40 6.10 Average Thermal Output, Kilowatts per foot 16.6 19.0 Maximum Thermal Output, Kilowatts per foot l'aximum Cladding Surface Temperature, F
654.
654.
t'aximum Fuel Temperature at Hot Spot, F 4000 4400 Total Reactor Coolant Flow,10 pounds /hr 131.3 137.8 6
15.13 16.52 Core Average Coolant Velocity, ft/sec Departure from Nucleate Boiling Ratio 2.07 1.65 at Design Overpower Departure from Nucleate Boiling Ratio'.
2.50 2.06 at Design Power o
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d References _
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1.
Supplement No. 2 to the Sparety Evaluation by the Directorate of Licensini-U. S. f.tomic-Energy Connission in the Matter ofuke Power Company, Oconee Nuclear Station Unit 1, Docket No. 50-269. December 19, 1972.
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2.
Letter, J. F. Stolz (NRC) to K. E. Suhrke (B&W),' April 15, 1976.
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6.2.1
_ Containment Design Basis LOCA Calculations 1
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j The data for mass and energy release to the containment foll owing a postulated primary system rupture were calculated using the CRAFT-2 code for the blowdown period, the REFLOOD code fo cc:pq. Pfp i n-rc u&
e refloodingperiodandtheB&W(che;(ptcodeforthelongter m
period of decay heat boiling'.
The analysis considered boiling withintheco[eandsteamgeneratorsuntilallcoeI generator sensible heat was removed.
These methods are designed to conservatively m'aximize containment pressure.
The methods are i
documented in Appendix 6A of B-SAR-205 ftandard Design PS n
were approved by the flRC in the Safety Evaluation Report fo dated May 1978.
i Since f4idland is not a B-SAR-205 design, the inputs to the mass and energy release methods were modified na acceptable manner for Midland 1 and 2.
Short-Term 810wdown for Subcompartment Analysis The blowdown rates Rdm postulated primary system rupture n
containment subcompartments were calculated using the CRAF i
BrodHted code.
This code uses thelBurnell correlation to calcula when the break fluid is subcooled and the Moody slip flow m calculate flow when the break fluid is saturated.
o Stagnation conditions at the break are approximated by removing the mcT flux option from the CRAFT-2 code.
m This method is documented in Section 6.2.1 of B-SAR-205 Standard Design PSAR and was a pproved by the f1RC in our Safety Evaluation Report dated May 1978 eo e g
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Main Steam Line Greak Within Containment Mass and energy releases for a spectrum of steam line breaks were calculated using the TRAP-2 code that is described in Section 15.1.14-1 of B-SAR-205.The TRAP-2 code models the primary and secondary systems of a PWR including the core an power excursion which may occur in the core following a main steam line break.
The code calculate $ heat flow from the primary system into the broken steam generator.
The primary. system heat flow produces additional steam,which is added to the containme n.
The TRAP-2 code was found to be acceptable for mass and release calculations in the f;RC Safety Evaluation for B-SAR-20 i
dated May 1978.
Since Midland is not a B-SAR-205 type plant, the TRAP-2 analyses were modified with input information specific y
for Midland 1 & 2.
c ha rac. !r+h rics The feedwater pump A
were modeled in the analyses so that following the pesSkted postulated gam line rupture feed at j
w er flow was calculated to increase into affected steam gfeerator a s the discharge pressure decreased until the feedwater system w ae.
Since all fluid within the steam generator is assumed to be add to the contaiment as steam the increased feedwater j
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increase containment pressure and temperature.
q-The 50,200 pounds of steam contained in the process syste ng of Midland - 2 was also included in the TRAP-2 model.
The valve isolating this system was assumed to fo,il so that this steam w also assumed to be added to the containment.
He have concluded that the applicant's calculations for main steam line break m ass i
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and energy release are conservative and therefore acceptable.
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l 3/12/79 15.2 Analytical Techniques The analysis methods for postulated transients and accidents are normally reviewed in a generic sense.
In this regard, we have received submittalf from B&W for the loss-of-coolant accident.
In addition, we expect B&W to submit topical reports on steam line break and feedwater line break in the near future.
The description of the computer programs used in the
, analysis of these accioents have also been submitted.
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The loss-of-coolant accident review has been completed and analysis methods were found acceptable.
Our safety evaluation is documented in a letter dated February 18, 1977.
84 As soon as B&W submits an analysis method / for steam line j
break and feedwater line break, we will review these I
reports.
The status of code reviews as well as the status of steam line breakf"'feedwater line break are described below:
(
1.
The following topical reports have been approved:
a)
RADAR (BAW-10069A)
Approved 10/74 b)
PUMP (BAW-10073A)
Approved 12/75 c)
CADD (BAW-10076PA)
Approved 9/74 l
d)
FLAME 3 (BAW-10124A) Approved 5/28/76 l
2.
POWER TRAIN and TRAP-2 topical reports are currently under review by the staff.
These analyses methods are described in BAW-10070 and 10128.
Our review of these topicals has progressed to the point that there is reasonable 4
assurance that the concusions based on these o
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analyses will not be appreciably altered by
' completion of the analytical review.
If the l
final approval of these topical reports indicates that any revisicns to the analyses are required, Midland /'& J will be required to implement the results of such changes.
3.
Main Steamline and Feedwater Line Breaks - At present our review of steamline break and feedwater line break is continuing and incomplete.
There are outstanding questions,I ApIlicantmustanswer before we can approve the analyses methods used.
In addition, the,a licant has promised to analyze JSt:,w m e 1.pp a.Ev 4
steamline break case using three dimensi^onal thermal-hydraulic and neutron kinetic codes and submit the results in December 1979.
Following the applicant's submittal of the answers to all our outstanding questions, itaff will review the submitted material.
At that time upon completion of our review, we will issue a supplement to the present SEP..
Based on previous acceptable analyses for B&W plants, on comparison with other industry modr i-on independent j
staff audit calculations, and on previo;u : tart-up testing experience, we conclude that with the exceptions noted above, the analytical methods used for Midland p & J are acceptable for the operating license stage.
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MEMORANDUM FOR: Robert L. Tedesco,' Assistant Director for Reactor Safety, DSS FROM:
Zoltan R. Rosztoczy, Chief, Analysis Branch, DSS
SUBJECT:
COMMENTS ON " GENERIC ASSESSMENT OF FEEDWATER TRANSIENTS IN REACTORS DESIGNED BY BABC0CK & WILC0X".
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I have reviewed the draft of the report I received on 4-30-79. My general comments are:
1.
The report in its present form is too lengthy, it does not concentrate H
on the import' ant issues, has dozens of incorrect, misleading statements, there are many statements which are not self explanatory and are not justif.ied in the report.
2.
The report should be reorganized to. focus on the important issues as follows:
I.
Introduction'and Background II. Basic Design Deficiencies of B&W Reactors a.) The second fission product barrier, the primary coolant system pressure boundary is not single failure proof.
b.) Uninterrupted natural circulation is not possible, following depressurization events, like steam line break, small break, small break' LOCA,' steam generator tube rupture, so on.
c.)
Instrumentation is not available to assure adequate safety in case of A00's and accidents.
d.) The reliability of the feedwater system and the auxiliary feedwater system is very low.
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'e.) Temination of safety infection intiates automatic pumping of the containment sump into the auxtliary building.
f.) Because of the above deficiencies B&W plants do not satisfy the requirments of NRC regulations.
III. Shortconting of Plant Safety Evaluation a.) The feedwater transient was not analyzed in an appropriate
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manner.- Only overpressurization was considered, core cooling and radioactive material release was not evaluated.
b.) Single failure were not properly considered in the safety evaluation.
C c.) Operator error was not considered in the safety evaluation d.) The safety evaluation of feedwater transients is limited to the first few minutes of the event. A detail analysis should be perfonned until a stabilized condition is reached.
1 e.) Method of analysis is not sufficiently documented and is unapproved at the present time.
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f.) The available safety evaluation is not sufficient to pass' a judgement on the safety of the plant.
g.) The available safety evaluation is not sufficient for the development of acceptable emergency procedures.
IV. Operator Training and Emergency Procedures a.)' The mergency procedures do not provide sufficient guidance for the operators.
b.) Operators are not trained properly to handle multiple events, events complicated by additional fatures and operators errors.
c.) Most likely, the instructors training reactor operators are not, properly trained either.
V.
Lessons Learned from the TMI-2 Incident a..) Plans with serious design defficiencies shoudl not be pennitted to operate.
b.) No operating license should be granted until the applicant demonstrated compliance with all existing regulations.
c.) Tech. Specs. should be strictly enforced.
I d.) Past operating experience has not been factored into the licensing process.
a-VI. Recommend,ed Actions a.) The primary relief valve should be " valved out" on all plants. This is the only acceptable solution for continued operation of B&W plants.
W b.) A detailed study on the natural circulation character l
1 of B&W plants should be submitted within 60 days by each licensee. NRC5hould issue its findings on natural circulation 30 days after the submittal.
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_4 c.) NRC should issue within 60 days a -policy statement on instrumentation requirements with respect to GDC 13.
Licensees should be required to comply with these requirements within 12 months.
d.) NRC should review the feedwater and auxiliary feedwater systems of all operating B&W plants within 60 days.
Licensees should be required to correct all shortcomings
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within 12 months.
e.) Eliminate automatic pumping of the sump as a condition fdr continued operation.
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f.) NRC should present a sample safety evaluation and emergency procedures for feedwater transients to interested licensees.
The presentation should take place within 30 days.
g.) Consideration of passive failures is required in plant safety evaluations. However, are NRC position on passive failures has not yet been generated. NRC should issue a position paper on passive failures, within 30 days. Special attention should be payed to safety valves.
h.) NRC should issue a position paper on " operator errors" within 30 days.
i.) When the calculational methods are not already approved, the licensees shoud be conditional on receiving an NRC approval of the calculational methods within a specified time period.
J.) NRC should revise the train,ing manuals on feedwater transients within 90 days.
k.) A cmplete analysis of the feedwater transient and emergency procedures for FWT's should be provided by each licensee within 60 days. NRCshou]dissue
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its findings 30 days after submittal of the feedwater transient analysis.
1.) Each licensee should submit a revised Ch.15 to the SAR and a set of emergency procedures within 6 months.
f(RC shBuld cmplete the safety evaluation of all B&W plants within 12 months.
m.) NRC should perfom audit calculations for a selected B&W plant with the best available methods. The feedwater transient audit calculations should be completed within 90 days. Audit calculations for other transients and accidents should be perfomed within 12 months.
n.) NRC should set up a pemanent audit group that monitors and checks the way how the various NRC branches are enforcing NRC regulations and policy.
~
o.) A task group, similar to this task group should evaluate the safety of CE, GE and Westinghouse plants and make appro-priate recommendations.
It is likely that may of the findings of this report apply to plants of other designers also.
l
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p.) NRC should tussue a position paper on the utilization of operating experience in the licensing process.
3.
The, Task Group should meet with B&W to discuss questions relating to this report and the recanmendations of the report. A meeting with Mr. Michelson would also be useful.
4.
A large portion of the information presently in the report should be presented as Appendices to the report. Of course these parts first must be corrected.
5.
The Task Group should meet as soon as possible and discuss the report in detail... This is t.he only.way how a task group report can be prepared. Approximately half of the present draft, in my opinion, is either incorrect or inappropriate. I certainly could not support this draft.
i 9
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Responses to NRC Questions on llYTRAN Topical Review Babcock & Wilcox.
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1.'
Equation 1, which 10 en antrgy equatien, 10 writtsn in terms of a flow-
~
weighted enthalpy ratner than the standard definition of enthalpy.. This should be made clear in the definition of terms.
Response
The definition of the enthalpy term of equation 1 will be revised to l
"h = flow-weighted or mixing cup enthalpy., Btu /lbm."
2.
Equations 2 and 3 for the apparent density are incorrect.
Please provide the corrected equations and verify that the correct equations are in the code.
Response
Equations 2 and 3 will be revised. The correct forms are C [X + (1 - X)(p"/p )] + $ - XC [1 - (p /pg)]
f E
p" = [p X + p8(1 - X)]
(2) g (C,[X + (1 - X)(p /p )] + $)2 f
88 (P
-A c f g}
, 38.26 (p )
G 2
8 p
f The code programming was verified, and the correct equations were used in the code.
The error might have originated during preparation of the report.
3.
Equation 7 requires the heat flux at the inner surface of the c kdding as input. Demonstrate that the functional forms chosen to r'present the heat e
flux conservatively model the actual heat flux during oscillatory tran-sients.
Response
The question will be answered by discussing the limiting assumptions made in the formulation of the cladding thermal analysis:
The most limiting assumption was the omission of the thermal capacitance of the fuel peliets inside the cladding. To examine the effect of this assump-j tion on the problem solution, the formulation of the same problem without this assumption is shown below. The nomenclature used here for the cladding portion Babcock & Wilcox s
f of the problem is identical to that shown in the original topical report sub-mitted.
The lumped form of the energy equation for the cladding is identical to equa-tion 7 in the topical report; i.e.,
BT*
p,C,Pil L(1 - 1/2r ) 3
= PH 6 - PH $
g 1
with boundary condition k
$, =
(T,- T,)
The differential form energy equation for the fuel pellets is (r
) + q'j' PC
=
gf p = density of fuel, f
C = thennal capacity, f
k = conductivity, g
q"'
= volumetric heat generation rate.
- The boundary conditions are E
=0 and
- r=o,t E
-k
=$.
- r=r
,t g
For an oscillatory heat flux $g as used in the present HYTRAN input, the equi-valent physical reality would be an oscillatory heat generation rate, q'y'.
For example, consider q'f" = a sin (bt) + C a = amplitude of heat generation rate, b = equivalent time constant,
[
c = constant or base heat generation cate.
Babcock s.Wilcox 3-I l
.a
The two energy equations abov'e can be solved by coupling to the boundary condi-tion of U, the fluid film coefficient, and T, the fluid bulk temperature at g
each time step.
The two differential equations with the common boundary condition at the in-terf ace of the fuel and the cladding are highly nonlinear in nature. The finite difference numerical metbod is generally used to solve the problem.
For illustration purposes, it may be assumed, within reasonable tolerance, that the resulting heat flux $ can be approximated by the following function-al form:
$ = a' sin (b't + c') + d' 1
where a' and b' are the new amplitude and equivalent time constant, respec-tively, c' is the phase lag of $ with respect to q'g", and d' is a new con-stant.
In other words, the heat flux $, considering the thermal capacitance 1
of the fuel pellet, will still be oscillatory in nature.
In reality, the feedback effect of the thermal capacity of the cladding, particularly the film heat transfer coefficient of the cooling fluid on the heat flux is rela-tively small. Note that the HYTRAN code is limited to heat transfer analysis f
at heat fluxes below the critical heat flux of the coolant. Therefore, the film resistance of the fluid is generally small at the cladding outer surface.
, The preceding discussion illustrates the assumption that using a known heat flux at the fuel cladding inside surface is generally a good approximation of the oscillatory volumetric heat generation in the pellet.
i The analysis also shows that the input of the oscillatory heat flux to the llYTRAN code could be at a different frequency (most likely lower frequency due to damping) and amplitude (could be' higher due to damping) of the realis-tic oscillatory heat generation inside the fuel.
B&W HYTRAN users were aware l
1 of this fact and have been using conservative values for the input.
4.
Equation 10 for the equivalent cladding thickness is incorrect but appears to be conservative. Verify the conservativeness of this equation.
Response
The exact form of equation 10 is 4_
Babcock & Wilcox
_ _ _. _ _. _ _. _. _ _ _ _ _ _ _ _ ~. _ _
2
- 2
~
r, r
r
. 2in El - 1 g
2(r', - r )
2 r
e r
o
. i g
2 3
Approximation of r r a r was made.
For the B&W PWR design, r, a 0.215 inch, g
and r a 0.191 inch. This approximation was about 12% off in underpredicting g
the equivalent length 1,.
This would cause an overestimate of the heat flux o
(see equation 8) by the same amount. The net result should lead to a con-g servative analysis of the thermal hydraulic response of the fluid flow.
S.
Equation 11 is incorrect.
Please provide the corrected equation and veri-1
]
fy that the correct equation is in the code.
Response
Equation 11 will be revised to 3T 2r r
I 4 ).
(11)
J
=
at p C (r2-r) i Y
0 i
mm o l
The correct form has been used in the code.
t 6.
Equations 12 and 13 appear to be superflous. Why are they included in the report?
Response
The equations are superflous and will be deleted in the forthcoming report re-vision.
7.
The references to section 2.4 on page 8 are incorrect. Please provide the correct references.
Response
The references should be section 2.1.6 in both cases. Corection will be made in the report revision.
8.
The value for the n term in equation 25 is incorrect as given.
The expon-ent of 10 in the equation should be negative.
Please update. Babcock 8.Wilcox
.s a wse m ow= -wo-wm~
moe-
~w~~
- ~ * * ~ ~ * * ' ~ ~ ~ ' * * ^ ^ *
Response
The correct form of equation 25 is n = 1.863 x 10- (14. 0 + 0. 0068P).
Update will be made in the revision.
9.
Equation 26 for the void fraction is strictly valid only for steady-state conditions. Justify its use for transient calculations.
Response
Although equation 26 was derived from a steady-state analysis and data, it is not necessary to imply that the model is completely unacceptable for use as 2
an approximate method in predicting the void level under transient conditions.
Because of the scarcity and high degree of uncertainty of the void models for the transient state, a decision was made to use Maurer's model in HYTRAN. To ensure the reliability of the code, experiments were conducted, and its ac-curacy was found to be satisfactory.
Some recent works on the subject of transient void prediction were brought to B&W's attention. The possibility of improving Maurer's model used in the code is under consideration.
10.
Reference 3 of your report gives several values of the (1 + c) term of equation 28.
What value is used in HYTRAN? What is the sensitivity of the HYTRAN results to the value of (1 + c)? Justify the value you use.
Response
The value used for (1 + c) in the code is 2.30.
From the sensitivity analysis of (1 + c) in the reference (see discussion in secti": 4.3 of the reference, page 14), quote:
It was found that (1 + c) was remarkedly independent of subcooling, heat flux, inlet velocity and void fraction within a probable error or deviation of i 15%. The effect of geometry and pressure was al-so not statistically significant so that putting (1 + c) = 2.3 cor-related all the ANL data analysed.
It is reasonable to assume that the sensitivity of the HYTRAN results to the value of (1 + c) is also small.
. Babcock a Wilcox
o o
11.
On page 11 you state that equation 29 results from rearranging equations found in reference 2.
That should be reference 3.
Response
This will be revised.
12.
Reference 2 is incorrect. The WAPD-TH-362 should be WAPD-TH-326.
Response
This will be corrected.
. Babcock s Wilcox
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W Docket flos.: 50-239/330 MEM0RAf!DUM FOR:
D. B. Vassallo, Assistant Director for LWRs,'PM D
i FROM:
R. L. Tedesco, Assistant Director for Reactor Safety, DSS
SUBJECT:
DRAFT SAFETY EVALUATI0ft FOR MIDLAtlD, UNITS 1 Ar1D 2 Plant flame: Midland Units 1 & 2 Docket flos.: 50-329/330 Licensing Stage: 0L Responsible Branch and Project Manager: LWR-4, D. Hood DSS Branch Involved: Analysis Branch Description of Review: Draft Safety Evaluation Review Status: Complete The Analysis Branch has reviewed the applicant's methods for determining mass and energy release to the containment for use in the containment design el culations. Our Draft Safety Evaluation is attached. The applicant's methods for mass and energy release from primary and system secondary ruptures are documented in B-SAR-205 PSAR.
These were approved by the NRC in the Safety Evaluation Report dated May 1978.-
j The applicant has modified the input to these methods to make them I
applicable for Midland. We have concluded that the mass and energy release data is conservative for containment analysis of Midland 1 & 2.
R. L. Tedesco, Assistant Director for Reactor Safety
~
Division of Systems Safety Encisoure:
As stated cc:
R. Mattson J. Guttmann D. Hood L. Phillips D. Picket M. McCoy l
J. Shapaker W. Jensen S. Varga Z. Rosztoczy g
P. florian l
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ilaId Vm i t 5 2?2 vocw
'n u m.,
ro-n 9/ 550 s
6.2.1 Containment Design Basis LOCA Calculations The data for mass and energy release to the containment following a postulated primary system rupture were calculated using the CRAFT-2 code for the blowdown period, the REFLOOD code for the gur rc~ c u t' refloodingperiodandtheB&Wfo[nte,mptcodeforthelongterm period of decay heat boiling.
The analysis considered boiling within the ccIe and steam generators until all coie Tteam M generator sensible heat was removed. These methods are designed to 4
I conservatively maximize containment pressure. The methods are j
documented in Appendix 6A of B-SAR-205 standard Design PSAR and were approved by the fiRC in the Safety Evaluation Report for B-SAR-205 dated May 1978. Since Midland is not a B-SAR-205 design, the inputs to the mass and energy release methods were modified in a acceptable manner for Midland 1 and 2.
Short-Term Blowdown for Subcompartment Analysis The blowdown rates (Sh postulated primary system ruptures within containment subcompartments were calculated using the CRAFT-2
% di1ted code. This code uses the1Burnell correlation to calculate flow when the break fluid is subcooled and the Moody slip flow model to calculate flow when the break fluid'is saturated. Stagnation conditions at the break are approximated by removing the momentum flux option from the CRAFT-2 code. This method is documented in Section 6.2.1 of B-SAR-205 5tandard Design PSAR and was approved by the tiRC in our Safety Evaluation Report dated May 1978.
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Main Steam Line Break Within Containment Mass and energy releases for a spectrum of steam line breaks were calculated using the TRAP-2 code that is described in Section 15.1.14-1 of B-SAR-205. The TRAP-2 code models the l
primary and secondary systems of a PWR including the core and the power excursion which may occur in the core following a main I
steam line break. The code calculate $ heat flow from the primary
{
system into the broken steam generator. The primary system heat I
flow produces additional steam which is added to the containment.
I The TRAP-2 code was found to be acceptable for mass and energy release calculations in the NRC Safety Evaluation for B-SAR-205 dated May 1978. Since Midland is not a B-SAR-205 type plant, the TRAP-2 analyses were modified with input information specifically for Midland 1 & 2.
c hu vac f r"*/srtc5 The feedwater pump 4.
were modeled in the analyses so that following the pea 1 icted postulated steam line rupturej eedwater f
flowwascalculatedtoincreaseintoaffectedsteamgIeeratorasthe discharge pressure decreased until the feedwater system was isolated.
Since all fluid within the steam generator is assumed to be added to the contaiment as steam,the incr_ eased feedwater flow acts to increase containment pressure and temperature.
The 50,200 pounds of steam contained in the process system piping of Midland - 2 was also included in the TRAP-2 model. The valWb isolating this system was assumed to fail so that this steam was also assumed to be added to the containment. We have concluced I
that the applicant's calculations for main steam line break mass l
l
3 f
?
eI, and energy release are conservative and therefore acceptable.
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