ML19252B718
| ML19252B718 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Rancho Seco  |
| Issue date: | 10/20/1981 |
| From: | Thies A DUKE POWER CO. |
| To: | Bernero R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| Shared Package | |
| ML19252B719 | List: |
| References | |
| NUDOCS 8111170021 | |
| Download: ML19252B718 (18) | |
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.k DUKE POWER COMPANY CH ARLOTT E, N. ? 28242 A. C. THIE'S (7043 373-4P49 semon Wct Pnt.siotNT f'HODUCTION AND TRM&Mih5 ION October 20, 1981
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Mr. Robert M. Bernero, Director N,', g g. Wh}'.
Division of Risk Analynis Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission
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Subj ect: ORNL Evaluation of the Threat to PWR Vessel My D
Integrity Pased by Pressurized Thermal Shock Pressur
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Draft Interim Report
Dear Mr. Bernero:
Duke Power Company appreciates the opportunity to provide comments on the subj ect document. As you are aware, Duke has provided certain specific technien1 information regarding the Oconee Unit I reactor vessel in an effort to assist the NRC in the completion of this evaluation.
Duke engineers have reviewed the subj ect document and consider that the evaluation contains sig-nificant deficiencies in the area of thermal-hydraulics conditions and repre-sents unrealistic transient conditions. The applice* ion of these transients to the Oconee 1 specific materici properties results in misleeding and mean-ingless calculated vessel lifetime. Our more salient concerns are in the following paragraphs with additional details provided in the attached.
The evaluation of the reactor vessel thermal shock issue is extremely complex and reg.5es a thorough understanding of several highly technical disciplines.
Among the technical areas involved are instrumentation and controls, systems analysis, reactor vessel materials, non-destructive examination techniques, linear c]astic fracture mechanics, an1 probabilistic risk assessment.
In order to do a meaningful evaluation, these technical areas need to interact in a coordinated manner; t he results of one area cannot be input into subse-quent analyses without a thorough understanding of the basis of the ir,put.
This document does not indicate that any coordinated effort was attempted by the various organizations involved to assure that the results provided,were realistic.( In fact, the document tends to imply that the individual tasks were perfot'med independently of each other with the end result being a totally disjoint ed document that is not suitable for understanding and communicating the real perspective of the issue.
One of the principal mechanisms contributing to the occurrence of pressurized reactor vessel fracture is the creation of certain unique temperature-pressure t.ime histories at the reactor vessel, the calculation of which would require insights into plant design features, system failures and effects, plant per-formance constraints, and transient behavior.
The fracture mechanics analyses embodied in this document are, in most part, based on arbitrary, artificial 170021 011020 SUBJ R2912
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Mr. Robert M. Bernero, Director October 20, 1981 Page 4 is a very important issue that requires careful study and timely resolution, and the way to approach the issue is by means of a cogent and systematic analysis of relevant accidents and by consideration of plant specific features both in regard to system capabilities and vessel parameters. Duke has been fervently working on such an effort, and it is our hope that when this work is completed, the necessary perspective on this matter will be obtained.
In summary, the report in its present form is not suitable for understanding and communicating the real perspective of the issue.
In fact, it could unduly distract attention from the orderly efforts now being pursued on the resolution of the issue. Accordingly, we ask that the report be modified significantly taking into account our comments or be withdrawn from general release.
Very truly yours, A. C. Thies RLG/php Attachment cc:
Mr. R. C. Kryter Instrumentation and Contcols Division Oak Ridge National Laboratory P. O. Box X Oak Ridge, Tennessee 37830
DUKE POWER COMPANY Detailed Comments on OlWL Draft Interim Report Evaluation of the Threat to pWR Vessel Integrity Posed by Tressurized Thermal Shock Events CM t e r 1. 0 page 1-2, 3rd paragraph:
We agree with the statement about the need to perform " realistic systems analyses to determine appropriate input temperature and pressure tr annients for the vessel integrity studies, and [to evtluate accurately] the mechanical int.egrity of the pressure vessel" through plant specific studies. Iloweve r,
the analyses conducted thus far fall short of this goal, as recognized on page 1-3:
"...because thermal-hydraulic behavior needs to be fatther evaluat ed as recommended later in this report and because there are special control system provisions in Oconee 1 limiting transients, more analysis needs to be done before their results are applied to Oconee 1 or generalized to other plants."
Chap er 2.0 A clear and consistent defieition of the " runaway feedwater transient" is necessary.
The thermal-hydraulic analyses utilized in this report consider this transient to consist of an unmitigated main feedwater overfeed transient event with a concurrent failure of the turbine bypass valve system following a reactor trip transient.
Iloweve r, the probability discussion of Section 3.1 apparently visuali:*es this accident as a more general secondary system upset condition which includes steam generator overfeed transients, steam generator pressure cont rol malf unctions, and events involving f ailures in f eedwater flow control and SG pressure contrel functions.
C_hapter 3.0 page 3-1:
In order to obtain t he real perspective of the safety significance of this problem, one needs to consider the probability of occurrence of a break in the reactor vessel at the correct location and of sufficient size to com-promise adequat.e core cooling capability as a result of crack initiation and propagation. This probability is composed of several (possibly independent) pr otiabili ti es, including (1) the probability that a break large enough would occur given that the fracture mechanics calculations predict a through-wall crack propagation.
(2) the probability that a through-wall crack propagation would eccur given the specific pressure-temperature condition (this probability
Mr. Robert M. Be.mero, Director October 20, 1981 Page 2 thermal-hydraulic accident conditions and not germane to the real plant situation, especially for the Oconee reactors. The major deficiencies in the thermal-hydraulic analyses are identified in the attachment to this letter. Portions of the ORNL report have also very appropriately discussed the limitations and deficiencies in the thermal-hydraulic analyses. Yet fracture mechanics calculations were done for these extraneous and irrelevant accident conditions.
The subject document is inconsistent within itself, which can cause signifi-cant interpretation dilemmas. The report was originally intended to be an evaluation of the B&W NSSS design and itc susceptibility to pressurized thermal shock. However, the document contains statements which make it unclear as to whether or not the intended purpose was achieved as noted by the following.
In Chapter 1.0, it is stated that although Oconee 1 was selected for the initial
- study,
"... thermal-hydraulic behavior needs to be further evaluated as recommended later in this report and because there are special control systems provisions in Oconee-1 limiting transients, more analysis needs to be done before their results are applied to Oconee-1 or generalized to other plants."
This is further elaborated upon by the following from Chapter 5.0:
"All the current simulations possess limitations which give concern for the realism of the thermal-hydraulic predictions. These limitations are, in part, inherent in the codes and also result from modeling deficiencies and questionable input assumptions..."
And yet the following statements occur in Chapter 8.0 without qualification:
"A summary of results for the five overcooling accidents analyzed is presented in Tables 8.6 and 8.7.
Table 8.7 indicates the total number of EFPYs that a B&W-type reactor can operate before the overcooling transients considered would likely result in vessel failure."
and also,
"...the inclusion of cladding in the analysis will also result in smaller threshold fluences. Thus, in this respect the results in Table 8.6 and 8.7 are somewhat optimistic."
We consider these latter two statements as misleading and inappropriate con-sidering the significant limitations of the study.
Mr. Robert M. Bernero, Director October 20, 1981 Page 3 An additional concern is that the subject document does not sufficiently address significant programs currently in progress that address the areas of vessel material properties that are supported not cnly by Duke Power, but also by other utilities that own plants with the B&W NSSS design. This is particularly surprising because by letter dated May 12, 1981, J. Mattimoe, SMUD, on behalf of the B&W Owners Group, submitted a letter report to the Staff outlining such programs that had been completed and those still under-way.
By failing to recognize the other ongoing studies on this issue, the report implies that it is "the best available information." This is incorrect.
It should be noted that certain branches within tne NRC Staff are aware of these programs.
The evaluation of the reactor vessel fluence aspect, the interpretation of the Oconee reactor vessel material parameters, and the fracture mechanics calculations contained in this report have also several limitations.
It is apparent that the chapter on fracture mechanics calculations contains several pessimistic presumptions and opinions based on unsubstantiated data and limited information. A technical report of this nature should be based on an objective analysis.
Further, the document fails to address two important items which are associated with this issue. One is the enhanced inservice examination of the reactor vessel beltline region welds in order to achieve a higher confidence level in selection of initial flaw size. As the NRC Staff is aware, such an enhrnced examination was performed on the Oconee 1 reactor vessel during the current outage, using an ultrasonic technique with a stand-off distance that a?. lows detection of neer-surface flaws. Not only were all results within ASMI code allowable, but also they were smaller than those sizes critical to the thermal shock issue. All indications were considered to be pre-service induced rather than service induced. The second item is thermal annealling, which is briefly mentioned, and then only in a positive sense. While the technique used in controlled conditions may seem promising, extensive work and effort will be required to perfect a technique suitable for use on an irradiated PWR reactor vessel.
It is misleading to state that such a technique is currently practical, particularly when solely based on a personal communication and preliminary laboratory results.
As in the case of many other severe accidents, reactor vessel thermal shock cannot be envisioned to be forgiving to all bounding and overly conservative assumptions. In order to obtain meaningful conclusions of the severity of the problem, it is necessary to analyze systematically accident conditions by considering relevant initiating events, mechanistic system failures, and credible operator actions and by utilizing phenomenological models and methods that take into account realistic system boundary conditions and plant performance constraints. Duke has recognized that the reactor vessel thermal shock issue
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+ q-s is dependent [e the probability that flaws of certain unique size and orienta-tion capable ~ of through-wall propagation exist at the location of rainimum material str;ty.th), and
/3) the probability that the potemial transient events produce ' the prcasure-temperature conditions necessai'f for unarrested ere k propagation.1 page 3
.1, Table aro: 3a1 paragraph:
No basis is >rovided Yor the assigned probability of a runaway f eedvater transient (RFT). 'lde value provided is arbitrary and is not based on any review of operatinp,.cr.ocr pnce or quantitutive assessmen. af probability of RFT that causes severn overrooling conditions.
The RFT characterized by a f requency of occurre e of 1/Ky rept 'sents a general secondary system upset condition of an overcoolin;; nature and not the accident t;:eated in the sub-sequent sections of the reporr.,.
page 3-3, 1st paragraph.
The EPPY results provided in this paragraph are not valid due to the inherent errors and limitations of the thermal-hydraulic conditions utilized.
Further-more as discussed in detail later, the assumed fluence t *e per EFPY is Inc.ccurate.
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page 3-3, 2nd parar,raph:
j i
j The basis of this statement is not app $ rent.
Ihare5-4ht,At.hudicted
.emperature response for all transients including RFT apl % I.P (IRT). Within 600 secs ior RFT and 250 secs f o r t he MS LU, p r lha't, cool >ni. tem;erature is predicted to be below 200"F.
This figure would tchd to 'iodicate phhc the p redic t ed t empera t ure is well velow 212 F during ad.3t of the; transient rather than well above 212"F at the time of predicted failure.
Cljapter 4.0 page 4-1, 2nd paragraph:
The last sentence is incorrect. The neutror. power signal obtained from the RpS can modify main feedwater demand if its mismatch with the ICS reactor demand level exceeds a set tolerance only if the conditions in the steam generator _yermit, i.e.,
BTU limits, high and low S.G.
level limits override.
Loop A and B steam generator feedwater demands are reduced to zero in 15-20 seconds following reactor trip due to the combined actions of cross limits and BTU limits.
Tripping of RC pumps due to HPI actuation also requires that the operator verify the reactor has tripped. When the reactor trips, the ICS controls feedwater flow as described above.
page 4-1, 3rd paragraph:
The section is entitled " Reactor Protection System."
The integrated control system (ICS) discussed in the paragraph is not part of the RPS and should be separated out.
w A reactor trip will not only occur upon turbine trip, but also will occur on loss of main feedwater.
The P.PS low pressure trip is 1800 psi.
page 4-2, 3rd paragraph:
The Low Pressure Injection System is incorrectly described. Only two LPI pumps are start eu automatically. The third pump can be manually started and aligned to either A or B train.
Although the core flood tanks (accumulators) are mentioned in Sections 4.4.2, 4.4.3, t here is no description in the system descript ion paragraph.
page 4-2, 6th paragraph:
The LPI System is actuated when the primary system falls below 500 psi.
Substantial flow, however, by this system, could occur only when the system pressure falls below 200 psi, page 4-2, 7th paragraph:
While the discussion of the main feedwater control is fairly accurate, the discussion fails to include any mention of turbine bypass valve control and only brie.'ly discusses features of the ICS that tend to limit the potential for an overcooling event.
Further, therc is no mention of the Emergency Feedwater System and its controls and instrumentation, which, in fact, are totally independent of the ICS.
page 4-5, 2nd paragraph:
The centence is incorrect as written. The main f eedwater pumps are supplied water from the condenser hot well through three condensate booster pumps and three hotwell pumpc. The surge tank ana condensate storage tanks provide makeup to the hotwell, not directly to the feedwater pumps.
page 4-5, 4th paragraph:
3 Although the maximum invent.ory from all sources is 295 x 10 gallons, the actual unable inventory ir 142,000 gallons in the hotwell. Although conden-sate makeup to the hotwell can be achieved from the UST and CST, the maximum condensat.e available for an uncontrolled main feedwater flow event is 192,000 rallons (or for 9 minutes at full flow rate).
page 4-5, 5th paragraph:
The first setpoint is incorrect.
The total feedwater demand will run back at a maximum rate of 20% per minute to track generated megawatts following a reactor trip it' the conditions in the steam tenerator will permit that demand.
If the steam generators cannot accept a 2G% per minute runback, the BTU limits
will reduce the demand to whatever value is appropriate.
The cecond setpoint is partially described correctly; the following should be added. The f eedwater valves will transfer to emergency Icvel control which compares the actual Icvel in the steam generator to a 50% level setpoint.
This circuitry will either open or close the startup s_.1ve as appropriate with the pumps controlling on D/P.
In addition t o the listed trips, each main f eedwater pump will trip on low suction pressure or on overspeed.
page 4-6:
An attempt is made to represent functionally the main feedwater portion of the ICS.
This figure shoulc' ba redrava to represent more accurately the control system. As a minimum the ]cvel limiter aL uld be moved above the controller and another controller added to control the startup valve on loss of all RC puups, page 4-7:
For single control failures occurring below the manual control points a.lso, the high level trip of the main feedwater pu:nps will be available to mitigate the events.
No discussion of the aval.1abi]!ty of instrumentation und controls is presented.
A description of the present system was provided to ORNL (copy of July 23, 1981 Ittler of William O.
Parker, Jr. to NRL) and yet no mention is made of the multiple iristrumentation available to the operator.
The fdrst sentence on page 4-7 should be changed as follows:
This review divided the main feedwater_ portion of the ICS into three general areas, as shown in Figure 4-3.
page 4-7, 2nd paragraph:
control is required following 1CS failure.
In tae second sentence, ma a.>.
Sections 4. 5.4 t hrough 4.9 use the t erm c>:cessive feedwater on numerous occcsions with no at-vt to define the an',unt of excess.
Someone who does not know t'r syster not unders omd the.lif f erences and in f act could interpret excessivt
.o m inr to mean the hypothetical runaway feedwr.ter trusient.
This should h larified in future reports.
- tr x ss
. tc e sW1d cls 7,ed as fo]Iows:
It should be noted that with-c.t ri, e t.. z se u.*:. " b>at avel trip for the feedwater pumps, failure of
_ startup 1 w c i i
.a 1 e
" low" condition can result in an overfeed of one st eam gener:a.or.
page 4-8:
In Table 4-1, it shculd be noted that several indicated failures cause over-feed to only one steam generator.
The Oconee 1 event sequences referred to were submitted to the Staff in July 1981 as part of the Abnorm al Transient Operating Guideline Program.
These are currently under review by the Staff.
page 4-10, Section 4.7:
Overcooling transients are alerted to the operator by numerous alarms and are easily recognized by lecreasing temperatures and the cause identified by steam generator conditions.
The continuance of main feedwater at 100% flow rate requires multiple ICS failures and failures of other flow limiting functions or deliberate operater action to open feedwater valves to both steam generatcrs and to disable certain trip functions.
Even then, the condition can persist only for a short duration because it is self-limiting (due to high SG pressure conditions or due to rapioly d2minishing inventory).
page 4-11, 2nd and 5th paragraphs:
Based on a detailed review of the IRT and TRAC calculations, we believe that to characterize them an being "approximately bounding" is overly optimistic.
Chapter 5.0 Of the fcur Oconee events, only two events can be considered as representative initiatine, events of the general se ondary system upset condition category of events of interest in reactor vesse:' overcooling. These t.wo: events are the 1/4/74 switch ard isolation event of Oconee 2 and the November 10,1979 Joss of 3CS power event of Oconee 3.
In the Oconee 2 transient the overcooling was caused by excessive steam load combined with a high initial design pre-scribed steam generator level, which has subsequently been reduced. For the Oconee 3 event also, the major contributor was excessive steam load (auxiliary 6t eam drawdown und partially open turbine bypass valve) with sone minor con-tribution from overfeeding one steam generator.
In both cases the primary system cooldown was limited to 420 F, and even if the operator had failed to take action the transient would have progressed cnly to a modest overcooling event and not of the severity calculated to occur in the present analyses.
The third Oconee event (June 13, 1975 event in Oconee 3) involved a stuck-open FORV, and the actual thermal-hydraulic trannient behavior was milder than the calculated small break LOCA transient.
The fourth event involved a temporary undercooling in Oconee 1 on December 14. 1978. During this type c' a event, the primary system undergoes a rapid but finite cooling of the primary side when normal cooling is reestablished. The primary system cocidown is limited to 520 - 540 F and as such is not different from typical reactor trip events as far as overcooling events of interest for reactor vessel integrity are concerned.
It is worthwbfle to examine the coerating history of the Ocence reactors with regard to the occurrence of the "PJT" event, which is characterized by the allure of the main feedwater ficw control system tr un back feedwater flow af ter a reactor t rip, f allowed by the iailure of the 7; high 1cvel trip of the MFWP's and concurrent stuck-opan lailure of the TEV System, tnd not con-n
sidering any operator actions. The three Ocence units combined have now accumulated 23 reactor years of operation, during which time 186 reactcr trip events have occurred.
Our review of these reactor trip events indicates that in all cases the feedwater was run back, either promptly or with acceptable delay, after the reactor trip and did not represent a perpetual full flow condition. Furthermore, the SG high level trip of the main feedwater pumps have been challenged nine times as a result of moderate overfeed conditions due to slow feedsater runback or during loss of ICS power events.
In all cases successful trip of the system occurred as designed. With regard to the turbine bypass valve system, we have had no instances in which all the turbine byrass valves stuck open.
Although we have had a few instances involving excessive steam loads and/or partial failure of the TEV System, these events produced only modest overcooling of the primary system.
In all cases successful and timely operator action has been found to occur. Additionally, it should be pointed eut that design changes have been made and operating procedures have been written to prevent / reduce the probability of steam generator overfeeds (RFT). The present response to all three of the overfeeds listed in Appendix B would be a trip of the main feedwater pumps which would automatically initiate auxiliary feedwater. Auxiliary Feedwater would maintain steam generator IcVel at 25" (240 if all the RC pumps trip), thereby preventing both steam generator dryout and overfill. Operator confusion would not result on loss of ICS power since adequate backup instrumentation and controls and emergency procedures are availabic.
page 5-7, 1st paragraph:
Fluid mixirg between the Hpl and cold leg is of minimal importance during overcooling transients. The temperature on the downcomer RV is affected pri-marily by the temperature of the fluid at the weld location of interest and the fluid flow rate which govern the heat transfer coefficient.
page 5-7, 5th paragraph:
Flow distribution is important in determining the rate of heat removal from the RV wall and thus the temperature gradient in the wall.
The assumption of an arbitrary flow affects not only the thermal-hydraulic calculation but also the heat transfer from the wall.
page 5-7:
Additional deficiencies in the thermal-hydraulic predictions beyond those identified in Section 5.3.2 are evident and are as follow:
A.
It is inappropriate to use the IRT code for any external and released applications since the code is still under development. This is evidert by the fact that the code does not have a momentum equation and therefore all the flow rates in the analyses are non-mechanis tic.
In addition, the code has not been widely used in the industry and its capcbilities have not been demonstrated.
B.
The justification for using IRT to simulate a E6W configuration has not beca established.
Once a code has been verified (this has not beca completed
for IRT), the nodalization of the system being modeled must he qualified by comparison with data from the system being modeled. There is no indica-tion that this has been done using IRT on a E6W plant.
An example of this is the apparent failure to consider reactor vessel upper head circulation flow in the analyses and, also, the failure to consider the feedwater injection location and the pre-heating in the steam generator.
C.
There is no indication that the analysts had the necessary intimate familiar-ity with the Oconee plant to set up a realistic and appropriate set of boundary conditions for a simulation. Overcooling transients are strongly affected by boundary conditions. Without a realistic set of conditions, the transient response will not represent the true response, ar.d the results are essentially meaningless. A lot of simulation experience in terras of plant system f amiliarity and knowledge of code capability and limitations are essential.
Some examples of boundary condition crrors are:
a.
Incorrect feedwater flow.
b.
Incorrect turbine bypass setpoint and capacity.
c.
Incorrect HP1 actuation setpoint and flow versus pressure.
d.
Feedwater enthalpy versus integrated flow delivered.
c.
There are no secondary steam relief valves or atraospheric dumps.
f.
Omission of control system responses or additional assumed failures that are not identified, e.g.,
high SG 1evel trip of both main feedwater pumps.
D.
It is very misleading to label a particular analysis withaut explicitly identifying the failure assumptions made in the analysis. As an example, the IRT analysis labeled, " Turbine Trip", is actually a turbine trip with failure of the main feedwater to run back, with a failure of the high a
SG 1evel trip of the main feedwater pumps, with a failure of the turbine bypass valves on both steam generators, assuming a rapid decrease in feed-water temperature, and assuming a f ailure of t he operator to terminate the overcooling or perform any other mitigative action.
The assumptions which determine the transient response should not be lost in the generation of plots of results, and neither should the limitations of the code utilized.
Chanter 6.O page 6-1, Section 6.1, last paragraph:
Although the variations in Table 6.1 do occur in source parameters, they are not necessa rily uncert ainties in t he calculation of fluence. Many of these
'tems are accounted for in the calculational procedure.
For example, cycle and cycle-to-cycle core power distributions are averaged over the capsule irradiation period with the use of PDQ generated power distribution data at selected time intervals during fuel cycles.
The basis of these statements is experience in the analysis of 12 capsules from 8 B&W reactors.
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page 6-1, 6-2, Section 6.2.1, 1st paragraph:
Although this procedure was used to calculate fluence from Oconee 1 capsules GCIF and OCIE, an improved procedure is presently being used which incorporates the capsule geometry and P scattering cross sections directly into the r-0 reactor model, therebyelidinatingtheneedforcorrectivefactors.
pagt 6-3, Tabl e 6.2 :
An importast step was omit?.ed, that of normalizction of calculated flux to flux derived from measured dosimeter activities.
This table should read:
j.
Calculate capsule flux (E>l MEV) by multiplying the value from the r-0 model times the P /P3 and capsule perturbation factors and times an axial 3
shape factor based on the axial power shape in a peripheral fuel assembly.
k.
Obtain a normalization factor from the ratio of flux (E>l MEV) derivad from dosimeter reactions to calculated flux (E>] MEV) in the capsule.
1.
Perform an axial 2-D, P, r-z calculation.
y m.
Correct flux values from the r-G model with the P !P psule perturbation, 3
1, axial shape, and normalization factors.
n.
For weld locations, displacement factors from the r-z model and r-0 model are applied to the vessel flux (or fluence).
page 6-5, 6-6, Section 6.2.3, last paragraph:
The spread in nornalizing factors is misleading with respect to calculational uncertainties because only fission reaction data f rom the OCIE capsule were used to calculate fluence. Datn from OCIF were discounted because of suspected errors in activity neasurements. This was the first capsule analyzed at B&W and such large discrepancies have not been observed in any subsequent capsule analysis.
page 6-6, Section 6.3.
first paragraph:
The uncertainty evaluation in BAW-1485 was primarily based on conservative estimates with relatively little experience. Thus values of 1 30% for predicted beltline region fluence and 50% for certain weld locations were reported.
Since then, E&W has participated in the Blind Test, a calculational benchmark aponsored by the laight yater Reactor P_ressure Vessel Dosimetry,I_mprovement P_rogram (NRC funded) and the OC1F fluence calculation has been checked by anot.her phase of the LWRpVDIP.
(R. L. Simons at UEDL did the analysis.) The Blind Test indicated that the B&W transport calculationc1 procedure vould produce a fast flux (E>l MEV) that deviated 1 5% from a normalized capsule location to vessel surface and T/4 locations. The HEDL calculation of capsule fluence was 5% greater than the B&W calculated value.
In addition, analyses of 12 capsules from 8 B&W reactors have consistently shown E/C values with_a i 10% for fission
_9-
~
reactions.
Pused on these developments, recent estimates of fluence uncer-tainties are 1 10% at the capsule, i 15% in the vessel for time periods corres-ponding to capsule irradiation periods, and + 18% for predicted fluence in the future.
Comparable values for vessels in reactors without capsules are
+ 18% and + 21%.
It must be emphasized that these are conservntively estimated values in the absence of a detailed uncertainty analysis.
The 1 50% value reported in EAW-1485 for wcld locations was intended to indicate the added uncertainty (above 130%) of using axial and azimuthal displacement factors. Apparently, this was misunderstood in the ORNL analysis. A displace-meat factor of.89 (as is used for the critical weld location in the ORNL analysis) cannot be in error more than -12% when compared to the beltline region fluence. When statistically combined with the vessel uncertainty of i 30%,
this would result in a i 32% uncertainty.
pages 6-6, 6-8, Section 6.4:
The implication that there has been no verification of the B&W calculational procedure for the determination of fluence is incorrect.
In fact, E6W has successfully participated in the Light Water Reactor Pressure Vessel Dosimetry Improvement Program to benchmark both the calculational procedure and the dosimeter measurement technique.
page 6-7, Table 6.5:
DM a in this table apparently are based on extrapolation of the fast flux rm raged over cycles 1 and 2.
To obtain more realistic values, the extrapola-on (in time) should be based on a predictive procedure described in BAW-1485.
For Oconee 1, the use of this precedure is particularly important because of a conversion to an 18-month fuel cycle in cycle 6 with a corresponding reduction in ex-core fast flux of approximately 30%.
Inside of Outside of RV Wall T/4 3T/4 RV Wall 2.20E+18 1.22E+18 2.86E+17 1.053+17 2.67E+18 1.48E+18 3.47E+17 1.27E+17 3.26E+18 1.81E+18 4.23E+17 1.55E+17 3.66E+18 2.03E+18 4.75E+17 1.74E+17 Basis is assumption that relative effect of 18-month cycle in ANO-1 will be the same in Oconee 1.
Predictive data are avallable for Oconee 1 through cycle 7 but the calculations have not been made.
Chapter 7.0 page 7 1,
Section 7.1, Table 7.1:
Detailed descriptions of all data used and certification that such data are appropriate.for those analysis have not been provided. Also, error analysir far input dat? has not been provided.
page 7-3, first paragraph, next to last sentence:
The basis of the statement that uncertainty is not large in the parameters included in ASME Section III is not provided, first paragraph, last sentence:
This sentence conflicts with the previous statement.
Data should be to support this position. Explanation should be given as to how such data relate to the l'IR data which are used to evaluate vessel integrity. Also, data for the uncertainty in the determination of RTNDT sh uld be provided, second paragraph, first two sentences:
These two sentences appear to be in conflict. They should be clarified and supported with actual data to substantiate the opinion expressed in this para-graph.
second paragraph, last sentence:
The reference to support this statement should be provided.
third paragraph, first sentence:
The reference to support this statement should be provided.
third paragraph, last sentence:
This statement does not recognize the Oconee Unit 1 and the B&W Owners Group Research Program which is in progress and is generating this data.
page 7-4, second paragraph:
This paragraph appears to express an cyinion which should be based on sound data.
Since the statement is made that Regulatory Guide 1.99 is "not exces-cively conservative" for Oconee 1 weld metals, the data should be either pre-sented or referenced so that a better definition of " excessively conservative" can be betcer understood.
second paragraph, lart sentence:
Over 35 surveillance capsules have been removed from power reactors and the data support the conservatism of Regulatory Guide 1.99.
As for irradiation programs at test reactors, most of these are completed and the data are available, page 7-5, third pa, graph:
The statenent regard!'a reduction in upper-shelf energy of weld cladding is misleading. No mention is made of the fact that these fluences are well above that expected at EOL of any operating PWR.
At the fluence levels predicted, the cladding is expected to lessen the degree of crack propagation.
page 7-5, Reference 8:
This information should be included in the analysis and should not be stated as a reference since it represents a significant reduction in EOL fluence.
Chapter 8.0, page 8-2, first paragraph:
It is stated that if the temperature of a major portion of the coolant in the primary system is above 212 F, the opening due to crack propagation may be excessive and core cooling not maintained.
It is interesting to note that RFT and MSLB transient show bulk temperature decreasing to below 200 F.
Even with the considerable errors in the transients analyses provided, it could be postulated that the copious amounts of LP injection available would be more than sufficient to maintain the core cool, in much the same way as it is predicted to occur during a postulated LBLOCA.
page 8-2, second paragraph:
It should be noted, again, that only the vessel material properties are approxi-mately representative of Oconee 1.
The accidents analyzed are not at all representative of Oconee 1 or any other plant with B&W NSSS.
fourth paragraph:
For overcooling events, little if any vent valve flow will occur because there is minimal differential pressure between the core outlet and inlet. The thermal analysis is dependent on the downcomer temperature arf the flow conditions.
It is not apparent what flow conditions were assumed in the thermal-hydraulic calculations and thus what was assumed in thermal analysis of the RV wall.
page 8-3, second paragraph:
The ORNL analysis ignores axial gradient in fluence. Axial gradient of fluence is utilized in the B&W analyses.
The assumption of a pre-existent long sharp crack is unrealistically conserva-tive. This is particularly true for Oconee I which recently underwent a 100%
examination of beltline region welds with no cracks indicated.
page 8-4, third - fifth paragraphs:
No basis ic provided to support the assumption that the fluid-film heat-transfer coefficient is 1000 Etu/hr-ft2. F.
It is stated that this corresponds to full-flow conditions, but it is not stated what flow conditions were actually assumed in the transient calculations.
It is inconsistent to assume one mode of system operation during the transient calculation and a heat transfer coefficient based
>n a different mode of operation. This is particularly important in that severe overcooling transients may interrupt RCS flow and thus reduce heat transf er, and with RC pumps assumed running, a finite amount of heat is in fact added to the BCS.
~
page 8-6, Tabic 8.1:
The GRNL analysis should have utilized actual weld parameters inasmuch as these data were provided by Duke.
The RT is that of the base metal rather NDT than the weld and the chemistry is that of a hypothetical weld metal.
page 8-13:
It is inappropriate to perfonn the fracture mechanics analyses of the IRT steam line break of RFT cases with all their known deficiencies and atypical-ities. The analysis, discussion and results for these two cases should be deleted.
page 8-16, Section 8.4, last paragrcph:
2 The fluence rate of.046 x 1019 n/cm /EFPY is appropriate for the critical weld location (lower long weld at inner surface of vessel) for the first 4 EFPY (through cycle 5).
Thereafter, because of the conversion tggtheIg/EFPYat month fuel cycle, a better value for >4EFPY exposure is.033 x 10 n/cm the critical weld location.
Also, as noted previously, the 1 50% uncertainty does not apply to welds close to the beltline region.
Based on the initial analysis at the time BAW-1485 was written, this value would be + 32%; more recent estimates indicate about 20%. These are estimates because a detailed uncertainty analysis has not been performed.
page 8-17, Tabic 8.6:
The column with WPS and without WPS appears to have the line item designations reversed. WPS should provide additional time to fracture in all cases except LULOCA. Table 8.7 appears to have this relationship correctly identified.
page 8-19, Table 8.7:
The revised values of fluence /EFPY to account for the 18-month fuel cycle will lengthen the threshold time calculation f or times j>EFPY.
Threshold Time" (EFPY d)
Comments and Qualifications 26 V
26 WPS not assumed effective (44EFP if it were).
Through-wall crack predicted.
b 19 2
Based on fluence accumulation rate value of 0.046 x 10 n/cm /EFPY for the 2
19 n/cm /EF?Y thereafter. This value may have an initial 4EFPY and 0.033 x 10 uncertainty of as much as i 32%.
4F I
Chapter 9.0 page 9-1,
-2, Mitigative Measures:
In view of the inherent weakness contained in the report, it is considered that identifying the need for any changes is premature.
As a point of clarification, increasing the BWST temperature would have essentially no effects on reducing the severity of an overcooling transient.
Because a SBLOCA was not performed, it is not clear what basis there is for stating that an increased BWST temperature would reduce the degree of over-cooling caused by actuation of HPI.
In fact, with vent valve flow and plant specific analyses, the thermal shock concern for SBLOCA is minimized.
page 9-2, sixth paragraph:
The practicality of in-place annealling is overstated. While the principle may have been demonstrated under controlled laboratory conditions, extensive work is necessary to implement in-place annealling on an operating reactor vessel.
Extensive evaluations are necessary to demonstrate the acceptability of annealling at temperatures of 750 -850 F if a plant and its support syst ems were designed to lesser temperatures.
page 9-2, Section 9.9.2, General Changes:
An additional change that is presently occurring in BEM plants that will sig-nificantly lower the fluence on the reactor vessel wall, is the result of going to 18-month LUP reload cycles.
Once-burned fuel is loaded on the periphery of the core thus lowering the peripheral fuel assembly's power and the corres-ponding leakage flux or fluence to the vessel. The results of this are noted in the references of Chapter 7, Reference 8.
ChaEter 10.0 page 10-1. Sec tion 10.1:
See previous cocunents on fluence analysis.
Section, Csneluding Remarks:
All remarks are negative in context and are leading to uncertainty and lack of confidence in the final results.
Yet, statement is made that, "Nonetheless, for all their shortcomings, the analyses at hand are the best presently avail-able on a nonproprietary basis, and... [ merit] a great deal more study using refined techniques."
page 10-2, Section 10.2:
The implication here is that fluence calculations in general have uncertainties in the range of + 30% - + 50%. The uncertainty in the Oconee 1 fluence calcula-tions is much smaller as discussed in the comments on Chapter 8.0.