ML20071Q361
| ML20071Q361 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 11/22/1982 |
| From: | Warembourg D PUBLIC SERVICE CO. OF COLORADO |
| To: | Clark R Office of Nuclear Reactor Regulation |
| References | |
| P-82528, NUDOCS 8212290319 | |
| Download: ML20071Q361 (11) | |
Text
f I
public servlee company *e Ceaende 16805 Road 39 1/2, Platteville, Colorado 80651-9298 November 22, 1982 Fort St. Vrain Unit No. 1 P-82528 Mr. Robert A. Clark Operating Reactor Branch No. 3 Division of Licensing U. S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20034
Subject:
Amendment No. 28 to the Fort St.
Vrain Technical Specifications
Reference:
- 1) NRC Transmittal, October 5, 1982 (G-82325)
- 2) P-82343 ORNL Letter, August 4, 1982
Dear Mr. Clark:
The purpose of this letter is to provide you with the Public Service Company of Colorado's (PSC) understanding of the Nuclear Regulatory Commission submittal dated October 5, 1982 (Reference 1), which issued Amendment No. 28 to the Fort St.
Vrain Technical Specifications, and from that, propose resolutions to the requests made by your staff in the supporting safety evaluation.
As PSC understands the Commission's submittal, the license condition restricting operation of the Fort St.
Vrain Nuclear Generating Station to 70*s power has been removed, and the Fort St. Vrain Technical Specifications have been revised to incorporate those restraints necessary to:
(1) assure safe operation using comparison regions and (2) directly limit the maximum individual region outlet temperature to within FSAR stated values.
Further, it is PSC's understanding that with respect to these license changes, the requests made by your staff in the supporting safety evaluation are not pending but serve rather to provide guidance in the clarification of procedures.
Based on this understanding, PSC intends to operate Fort St. Vrain per the revised Technical Specifications and proposes the following resolutions to the requests of your staff:
Item a):
As indicated in my letter dated August 23, 1982, to Mr. Kuzmycz (Reference 2), PSC intends to conduct a rise-tn power program following each refueling in which l
sufficient data is gathered to evaluate core parameters 1
8212290319 821122 hDRADOCK05000 h.O M
Mr. Robert A. Clark N3yember 22, 1982 as well as any effects of the temperature redistribution phenomenon.
Item b):
As deemed acceptable by Mr. Kuzmycz in a telephone conversation, PSC intends to incorpurate, as a subject in operator training, the method for calculating the core inlet helium temperature, the errors to expect, and how the errors influence the calculated outlet temperatures of the seven nortnwest boundary regions.
Item c)1): Pubiic Service Company continues to maintain that the performance of the percent region peaking factor (RPF) discrepancy surveillance after each temperature redistribution event is unnecessary.
I have enclosed a copy of a
General Atomic Company (GAC) internal memorandum which contains responses to the questions and issues raised in the ORNL letter dated August 4, 1982 (Reference 3). This particular issue is discussed on pages 6 to 9 of the memorandum.
Public Service Company requests that the Nuclear Regulatory Commission re-evaluate their position on this issue based on the GAC response.
Item c)2): Public Service Company has already incorporated, as a subject in operator training, the method for calculating the outlet temperatures of the seven northwest boundary regions.
In addition, the equation will be included in the SR 5.1.7 surveillance procedure.
We believe that the resolutions proposed above adequately address the requests made by your staff.
If you have any questions, please contact me.
Very truly yours, k ~ Wuth Don Warembourg F
Manager, Nuclear Production DW/cis Enclosure
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Enclosure INIDNAL CDRR.'5PCdDE:lC to GP-1621 GA 1076 1
KEd.
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FRCH D. Alberstein/K. E. Asnussen REFER 20 FSV:L:10:DA:KA:82
~
TO W. A. Graul EATE Septauber 1, 1982 S)BJECT Cak Ridge Review cf New FSV Technical Specifications Oak Ridge National Iaboratory (ORNL) review ctatments on PSC's July 6, 1982 request for release d FSV from the 70% ' power restriction have been sent to NRC and to PSC. A copy of the CRNL letter, dated August 4,1982, is attached.
PSC, via letter PC-0911, has requested that GAC review and cement upon the CRNL letter.
ne purpose cf this memo is to 6ac:.nent GAC's respnse to the CR 1 c:xaments.
Issues of contention are addressed in the same sequence that they are raised in the CRIL letter.
Dese responses have been reviewed by members of the Core Engineering and FSV Project staffs.
PSC will probably provide additional ing to these responses prior to sutraitting them to NRC.
Its ' At -Rerconse -to -ORNL Ot2estions -of -Mareb-15 3 -1982 QBHF:
... we would reczemend that as FSV ventures into the as yet uncharted areas of higher care o P, data-taking procedures te inplemented similar to those used during the R-500K test series."
GAC-Rescense:
GAC agrees that collecting redistribution data at core pressure drops higher than 5 psid, in a canner similar to that used in
, Ie-500K (but scrnewhat simplified), would be prudent.
Se most appropriate way to do this would be as part of the rise-to-power test progran which is c3nducted during startup af ter refceling.
GAC does not believe that a requirenent to do sucn testing should be made part cf the Technical Specifications. A ccr=it: ent by PSC to conduct sucts tests until an equilibriun core pressure drop is attained should be sufficient.
DBSL:
"... GA noted that tests run in 1976-7 in which large differences between adjacent region outlet temperatures were generated (approx.
3200F) showed no significant type II errors. In spite cf these 3
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. _ - _ _ _ _ _., _,. ~ - -
I
N FSV:L:10:DA:KA:82 2
Septaber 2,1982
'best ef fort' tests, we reaain unconvinced that type II errors are not possible outside the m corner.
(In order for a test d that type to be conclusive, one would need to have both a large A T and type II flow, and it is not possible to induce or measure the type II flow.) Also, an CRE analysis of a pre-RCD redistribution event concluded that significant type II flow errors did affect regions 12-13 outlet tmperatures.1 We agree, however, that the surveillance methods described in the revised Tech Specs give adequate assurance cf operation cf the core within the ISAR-approved temperature limits."
GAG-Rescense:
GAC agrees that the test runs in 1976-7 were not by themselve.*
conclusive evidence that type II flow. errors are not possible outside the m corner.
ney were simply one additional piece cf evidence corroborating our mnclusion that significant type II flow errors are confined to the seven m boundary regions. Re pre-RCD
" redistribution" event can be explained, as was done by GAC, in tems of cross flow effects.
Alternatively, in concurrence with ORNL, it could be explained by a cxxabination cf cross flow and type II flow ef fects.
If it is postulated that during this pre-ECD
" redistribution" event a large gap opened between the adjacent regions, then there could have been a source of relatively cool gap flow which could have resulted in a significant type II flow induced error.
However, the installation of RCD's now precludes large gaps frczn cpening between regions and therefore eliminares this source cf relatively cool gap flow which is necessary for there to be significant type II flow induced errors.
We important point to be noted, however, is that in spite cf.
- differences in technical judgement all parties (ORE /GAC/PSC) agree that the surveillance methods required by the new Technical specifications give adequate assurance cf operation of the core within FSAR-approved tsperature limits.
Item -Cr-Protesed -Technical -Soecif ications DBSh*
" Tech Spec 2.21 - Individual Refueling Region Outlet Tetperature:
The calculation cf the region outlet tenperatures for 20 and 32-37 1
S. J. Ball et al., High-Tenperature Gas-Cooled Reactor Saf ety Studies for the Division of Reactor Saf ety Research Quarrerly Progress Report, April 1 - June 30, 1979, CRE/NJRm/TP.-356 (Nove.ber 1979).
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FSV:L:10:DA:KA:82 3
SePtenber 2,1982 also involves the core inlet taperature.
Why isn't the equation (given on p. 3-1 of Item D [GA-C16781]) included in the Tech Spec for ccxspleteness?"
GAC-Resconse:
As has been noted in several discussions between members of the GAC and PSC staffs, GAC is in agrement with CRE that the equation for region outlet tmperature should be included in Technical Specification 2.21.
The equation was excluded f ram GA-C16781 at PSO's reques..
M:
"LCD 4.1.7 - Core Inlet Orifice valves:
'Ite parenthetical expression in its C) would be clearer if it were changed to
'(i.e., RPFmeasured shall not be less than 90% of RPFcalculated)'+"
GAE-Resot nse:
GAO agrees that ORE's proposed wording cinrifies the intent cf this technical specification.
GAC reemen-is that PSC accept this change in wording.
Item-D: *machnical-Soecifications for Oceratien of.FSV-with -Recion Outlet Tentarature -Menserement Discrecancies s " -GA-C16781 - (June; -1982)
M:
"In Section 3.0 - Ref ueling Region Outlet Temperature, the discussion implies that core inlet helius t m perature is neasured, while in fact it is calculated frcxn a rmber cf farm eters including 4 circulater inlet tcgratures and flows.
We would reccznmend including a description of how it is calculated, what
' errors are expected, and how these errors irfluence the cceputed outlet tmperatues of regions 20 and 32-27."
GAc-Restensa:
As noted by ORE, the core inlet hellun tmperature is not directly measured.
However, the circulator inlet heliun tegrature is measured and, as will be discussed below, the core inlet helius tmperature is only about 1 to 1-1/2% (about 10 - 120F at 100%
power) higher than the circulator inlet helius taperature. 'nius, simply taking the core inlet tenperature as being the measured circulator inlet helius tmperature would result in only approximately a 10 - 120F error at 100% power.
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FSV:L:10: W KA:82 4
September 1, 1982 s
he core inlet helium temperature i's determined by taking the measured circulator inlet helium temperature, adding a calculated temperature increase (approximately its) due to the compressive heating by the circulator, plus a calculated temperature increase (approximately 1%) due to the heat added to the helium as it passes over the outside surfaces of the steam generators, and subtracting a calculated temperature loss (approximately 1%) due to the heat lost to the PCRV liner cooling systen. As mentioned above the net impact of these three effects is a calculated temperature increase of about 1 to 1-1/2fs between circulator inlet and core inlet.
Based upon the above, the following assess =ent of the uncertainty in core inlet helium temperature can be made.
Conservatively assume that each of the three calculated adjustments made to the measured circulator inlet helium' tcmperature in arriving at the calculated core inlet temprature are uncertain by 50%.
- Further, conservatively assume that these uncertainties are not randem but are simply additive.
mis assu=ption then yields a 1-1/2%
uncertainty in the calculated core inlet helium temperature due to uncertainties in the three calculated adjustments. For additioral conservatism, add to this uncertainty another 1% uncertainty to account for the 1% accuracy of the circulator inlet helium
~
taperature thermocouples.
The result is a total uncertainty in core inlet helium te=perature of only 2-1/2%, or at full-power operating conditions, an uncertainty of only approximately 200F Next, the impact of a 200F uncertainty in core inlet helium taperature en the maximum fuel temperature of a W boundary region being operated per a ecx::parison region is essessed.
Consider the expression for' calculating the region outlet
..teroperature for Regions 20 and 32 through 37, as presented on page 3-1 of GA-C16781:
( EPF
) { f1CM) 3 B
W T = Tin + AT
--gy glag g
g l
" Tin (1 - 5) + Tcr <-
(3) o
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September 1, 1982, where Ter measured outlet temperature of cm parison region,
=
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{ flow l-\\
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(EFg)
(flCWi )
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i and for si=plicity it is assumed that 6 ecuals unity (i.e., the ntrr.ber of coltr=s in the Ni regica and the cmparison region is the sa=e). The other notations are as defined in GA-C16781.
Taking the partial derivative of the expressica for To with respect to core inlet helium temperature, Tin, and noting fran (2) that AT e
1 g
(4)
Sm C:
where o Ti temperature rise for region being operated based
=
upon ca parison region, i.e. To - T nt i
then it can be shown that T
AT
- AT g,
g g
3T.
- AT (5) 2n er Frm the above expression it is seen that the sensitivity of' the calculated region outlet temperature to core inlet helium temperature is dependent upon the relative magnitudes of d er and oT. 'Ihe sensitivity i
increases as the difference between 6Ter and ATi increases and as oTer decreases. A conservative evaluation of this sensitivity follows.
b Conservatively assume that the comparison region's mismatch is - 100 F while the ni boundary region being operated based upon this comparison region is assumed to have a mismatch equal to the LCD 4.117 limit of plus 2000F.
- Further, conservatively assume that these mismatches occur for core conditions which result in a temperature rise fran core inlet to core outlet of about only 6500F.
'Ihis ccmoination of assu=ptions
==vimizes the sensitivity since it results in a large difference between si and oTer coupled with a small Scr.
Under these assumptions, the sensitivity is approximately -0.5.
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September 1, 1982 Thus, a given errer in core inlet tmperature would result in an error in the calculated region outlet te preature (To) of only about 50% of the magnitude of the error in core inlet t m perature. For the case of higher power level, and hence higher core inlet t mperature, where the allowable mismatches are less and ATc is larger, the sensitivity is less.
Similarly, for typical relative mis:tstches, the sensitivity is also significantly less (e.g., 0.2).
Canbining the conservative estimate of 200F uncertainty in core inlet helium temperature at full power co:xiitica and the conservative estimate of the higher sensitivity (i.e., 50%) to core inlet helium te:perature error, which is appropriate for inlet tegeratures of approximately 6600F or less, one obtains only a 100F error in calculated region outlet temperature.
As discussed in GA-C16781, Section 6.0, analyses have been perrormed to asse.ss the imract of uncertainties in cert:parison regien outlet temperature on m boundary region maximum fuel tenperature. These analyses show that a 100F uncertainty in region outlet temperature results in a 10-200F uncertainty in =4mm fuel temperature, depending upon tne RPF/ tilt coccination of the particular W boundary region. h*nen combined with the other, larger, uncertainties using the root-su:n-scuares technique, the impact of the uncertainty in core inlet te::perature on the ccanbined uncertainty is less than loF.
Tnus, the uncertainty in core inlet helium temprature has a negligible effect upon W boundary regio 1 fuel temperature.
DENL:
"In Section 5.2 on Percent RPF Discrepancy Surveillance, we have several questions relating to PSC's conclusion that the discrepancy does not need to be recalculated after an observed redistribution
. event.
First, their analysis shows that for any previously-observed redistribution (following RCD installation),
the maximum
- temperature shift in a comparison region would be such that a 1500F fuel temperature increase (maximum) could result in a W
- border region.
However, the effects of multiple redistribution events (e.g. there were 3 in 2 days in Nov.1981) and the possibilities of larger effects due to larger coreoP's (> 5 psi) could result in larger fuel temperature increases. Furthermore, a series of redistribution events wculd probably result in taperature changes in the same direction, and may not be amenable
. to the assumption of randcznness used in Section 6.0.
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Second, tne analysis relating to the' fuel kernel migration eftects also considers a fuel temperature rise due to a single evenc, while kernel migration is a c=:lative phenecenon.
Because of the uncertainties in refueling region behavior following redistribution
'i events, we would recu end that RPF discrepancies be recalculated after each observed redistribution."
GAC RP M,:
I With regard to ORE's first paint, related to multiple redistribution events, the following observations should be made.
GAC has reviewed the data en region outlet tcperature behavior for i
all redistributions experienced to date. Tnis review includes the multiple redistributions of April 23-24, 1981 (2 events), May 13, 1981 (2 events), and November 5-6, 1981 (3 events).
She review indicated the following.
I
' 1.
Even when the terperature changes associated with multiple events during a single power rise are cmbined, no candidate rmparison region experienced a decrease in region outlet temperature, relative to the expected temperature change, which was larger than the case cited on pages 5-5 and 5-6 of l
GA-C16781. In fact, the =v4== cabined decrease relative to
[
the expe<2ed temperature for any candidate comparison region was about one-half that cited on pages 5-5 and 5-6 of
(
GA<16781.
2.
In none of the multiple redistributim events in which candidate c mparison regions experienced a decrease in region outlet temperature did temperature decreases occur for each i
redistribution.
In fact, in only cne multiple redistribution event (Nov.1981) did a candidate cmparison region experience
~
a decrease in region outlet tecperature for more than one of the redistributiens. In this case, the region experienced two decreases totaling less than 150F.
i Based upon these findings, it is concluded that the cumulative
' effects of multiple redistributions on candidate caparison region outlet temperatures is less than the maximum effect experienced to l
date, and that the direction of the temperature change in candidate l
cm parison regions fr m multiple redistributions is randm.
j With regard to the second ORE point, the following clarifications f
are provided.
In the evaluation of fuel kernel migration eftects i
in GA-C16781, both single and successive redistributica events were considered.
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O' FSV:L:10:DA:EA:82 8
Septaber 1, 1982 s,
g In Secticn 5.2 of GA-C16781, pages 5-5 and 5-6, a single redistribution of the largest magnitude ex;:erienced to date is a= = =ad to affect a hypothetical W boundary regien which experiences the maximira expected time average fuel temperature.
Sna11 incremental kernel migration is shown to occur.
Such a region was chosen in recognition of the fact that, as pointed out by ORNL, kernel migration is a cu:rulative phencnenon.
Regions which experience the =vi== time average fuel tepreature are expected to experience the most fuel kernel migration during normal operation, regardless of any additional kernel migratien which might be caused by redistributions or other perturbations on the canparison region method of operation'.
In Section 6.1 of GA-C16781, the effects of successive redistribution events are considered. Tne canbined im;:act of five uncertainties, including redistributiens, upon the mav 4 = =
projected time average fuel teprature in any of the seven W boundary regions is assessed.
Again, a region with the maxi =.:n time average fuel taperature was chosen in recognition of the fact that kernel migration is a cu:nulative phencuenon and such regions experience the most fuel kernel migration during normal operation.
It was shown in Section 6.l that if a comparison regicn experienced a redistribution equal in magnitude to the largest experienced to date, the fuel ta grature in the m boundary region would increase by up to 140cF.
It was then =*ewd that such a redistribution occurred during each surveillance interval, and that they went undetected and uncorrected as much as 25% of the time.
It was shown that even with these conservative assuydons the resulting
=av4== time average fuel temperature in the W boundary region is
'less than that at which the fuel kernels could migrate 20 microns in 6 cycles.
In order to obtain fuel te peratures which result in 20 microns of kernel migration in 6 cycles, it is necessary to assume that a
. redistribution of the largest magnitude experienced to date occurs during each surveillance interval and that these large redistributions go undetected and uncorrected as much as 98% of the time. Such assumptiens are extreely conservative and unrealistic, and they clearly illustrate that it is not necessary to recalculate RPF discrepancies after each redistribution.
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'FSV:L:10:DA:KA:82 9
Septaber 1, 1982 P
Furthermore, it should be noted that testing during cycles 2 and 3 has de=cnstrated that redistribuciens are repeatable.
Tne
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redistributiens experienced in cycle 3 were remarkably similar in nature to those experienced in cycle 2.
'Ite initiating conditions for the redistributions were power increases wherein the accm.panying flow increase resulted in the steady state core pressure drcp (oP) increasing frcza approximately 3.2 psid to approximately 3.4 psid with the AP peaking at approximately 3.8 psid during the load change transient.
(bviously, data recorded at or below 40% power (i.e., relatively low core SP conditicns) would not be as meaningful for evaluating-the impact of redistributions en RPF discrepancies as data recorded at higher power levels.
The most significant time to record data for evaluating the impact of redistributions on RPF discrepancies is under steady state conditions after reactor power has been raised to the level, for example 100%, at which it is expected to be operated during the caming surveillance interval.
Such data will reflect the immet of any and all redistributions (single or multiple) that may have occurred during the power ascension.
It is for this reason that Technical specification SR 5.1.7(b) specifies that RPF discrepancy data be taken after exceeding 40% power but allows 10 calendar days for the plant to achieve stable operation (e.g., at 100% power), for data to be recorded, and for an RPF discrepancy evaluation to be made.
. CCNCLUSIONS:
- m.
"... we agree with PSC's conclusion that operation of FSV at 100% power poses no undue risk to public health and safety."
GAC RESPONSE:
We are pleased that ORNL, based upon its independent review of the July 6,1982 subnittal, has also reached this conclusion.
i oc:
M. R'. Mackney A. J. Kennedy
(
J. Iopez R. Rosenberg J. C. Saeger I
W. A. Simon IDF (Project 1900) 1-b e
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