ML20082A843
| ML20082A843 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 11/07/1983 |
| From: | Warembourg D PUBLIC SERVICE CO. OF COLORADO |
| To: | Wagner P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| P-83363, RO-83-43, TAC-52634, NUDOCS 8311180269 | |
| Download: ML20082A843 (7) | |
Text
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.s PLIhlie SerVIPO (Ot11 patty T CE3DH/dD 16805 WCR 19 1/2, Platteville, Colorado 80651 SOOY November 7, 1983 Fort St. Vrain Unit #1 P-83363 3 R@10W1HN Mr. Philip Wagner i
Project Manager E IAiM U. S. Nuclear Regulatory Commission n[
611 Ryan Plaza Dr., Suite 1000 W-Arlington, TX 76011 Faew x@o W 7'~
a
SUBJECT:
Technical Specification LCO 4.1.9
REFERENCE:
Reportable Occurrence 83-43 Cear Mr. Wagner:
On October 11, 1983, Public Service Company (PSC) notified the NRC of various non-conservative errors that were made in the original analyses which constitute the basis of Technical Specification LCO 4.1.9.
A follow-up, preliminary Reportable Occurrence was subsequently submitted (see reference).
By way of a telephone conversation on November 4, 1983, PSC committed to provide further information on interim measures we are taking to insure that compliance with the intent of LCO 4.1.9 is being achieved.
As was stated in the reference above, the following non-conservative errors and assumptions were determined to have been made in the development of LCO 4.1.9 as it presently exists in the Technical Specification:
1.
The length o f-the upper and lower reflectors was not included in calculating the gravity term in the momentum
- 1 equation. Only active core length was used.
2.
An intra-region power peaking factor (tilt) of 1.44 was used in the analyses.
3.
The calculation of the friction factor (laminar or turbulent) was indexed to the value of the Reynolds number at the core inlet.
4.
The effects of various coolant channel bypass flows were neglected or improperly taken into account.
h 8311180269 831107 0500026 gDRADOCK
. 5.
The flow defect in the highest power regions was neglected. With orifice valves set for equal flow, the highest power region (RPF = 3.0) in fact receives about 10% less flow than the average region.
GA Technologies supplied interim curves which conservatively corrected there deficiencies. These curves were reviewad by the Plant Operations Review Committee and placed into use via a Station Manager's Operations Order. Note that at the time, the reactor was operating at greater than 15% power, and therefore, the requirements of LCO 4.1.9 did not apply.
Specifically, the interim curves received on October 11, 1983, corrected the aforementioned errors as follows:
1.
The length of upper and lower reflectors was included in the calculation. This correction increased the effect of the helium bouyancy terms.
As a result, more core flow was required to prevent flow stagnation.
2.
A worst-case intra-region power peaking fa'ctor (tilt) of 1.61 was used. This factor is consistent with that as noted in the basis for LCO 4.1.3.
3.
The friction factor was indexed to the core average outlet temperature rather than the core inlet temperature.
This represents a worst-case assumption.
4.
The improper treatment of bypass flows was corrected by using those that are assumed in the FSAR.
5.
The flow defect was taken into account assuming a worst-case region peaking factor (RPF) of 3.0.
Again this is consistent with assumptions made in LCO 4.1.3.
l The net effect of correcting the errors in the original analyses and of using ultra-conserative assumptions was that the interim curves were more restrictive than those that are presented in the Technical Specifications.
It was apparent that these curves were unduly conservative, but had to serve as a short-term, enveloping calculation while GA Technologies continued their re-analysis.
i
I d On November 2,
1983, PSC received a telecopy of a second set n' curves based on the continuing analyses (Attachments 1 and 2).
The formal transmittal letter was received on November 4, 1983. The analyses for these curves differed from those previously in use as follows:
1.
In the October 11, 1983, letter, the calculation of the friction factor was indexed to the value of the Reynolds number at core outlet temperatures.
.The updated calculation modeled the local variation in the Reynolds number as the coolant flows axially through the coolant channel.
2.
In the October 11,
- 1983, letter, the primary system coolant inventory was assumed to be at the normal full-load value.
The updated calculation assumed a 7.5%
system overpressure, consistent with that possible by Technical Specification LC0 4.4.1.
Additional curves assuming 80% helium density and atmospheric conditions were also provided.
3.
In the October 11, 1983, letter, and in Figure 4.1.9-2 of the existing Technical Specification, region outlet thermocouple measurement uncertainty was not taken into account.
In the updated analysis, a 50 F uncertainty was assumed.
Additionally, all uncertainties were statistically considered, rather than assuming an arbitrary 25% uncertainty.
In the case of the " equal flows" curve of LC0 4.1.9, the combined effect of these additional considerations made the updated curves less restrictive - than those that were provided in the October 11, 1983, transmittal.
In the case of the "non-equal flows" l
curve of LC0 4.1.9, the combined effect resulted in 3 figure which I
was more restrictive than that provided in the October 11, 1983, transmittal.
Information from these updated curves was reviewed by the plant staff and the Plant Operations Review Committee, and incorporated into the previously mentioned Operations Order.
To date, there is no evidence that operation of the reactor with the figures in the existing Technical Specifications has had any adverse impact upon core performance.
I I
1 l
l I
l I
. For example:
1.
Primary coolant activity levels have remained low and constant.
Recent equilibrium power operation data has shown that:
a.
The product of_ primary coolant noble gas beta plus gamma times E is 0.035 Ci-MeV/lb while the allowable product per LCO 4.2.8.a) is 2.40 C1-MeV/lb.
b.
The primary coolant circulating I-131 equivalent is 0.623 Ci while the allowable value per LCO 4.2.8.b) is 24 Ci.
c.
The plateout I-131 equivalent is 155 Ci total while the allowable value per LCO 4.2.8.c) is 5,000 Ci per loop (10,000 Ci total).
d.
The plateout SR-90 equivalent is 1.38 Ci total while the allowable value per LCO 4.2.8.d) is 140 Ci per loop (280 Ci total).
2.
Post irradiation destructive examinations of Segment 1 and Segment 2 fuel rods has confirmed satisf actory performance of the coated fuel particles.
In summary, continued operation based on the updated figures is justified for the following reasons:
1.
No evidence exists which would indicate that previous operations under the current LCO 4.1.9 have had adverse effects on core performance.
2.
Since the updated figures are even more conservative than those in the current LCO 4.1.9, compliance with them decreases the probability of adverse effects on core performance. Any significant increase in primary coolant activity due to fuel particle failure would be detected via on-line monitoring and routine surveillance testing.
3.
The updated figures were formulated using worst-case assumptions with respect to region peaking factors and region tilt.
These conditions will not be experienced during the remainder of this refueling cycle.
. 4.
In addition to the conservatisms cited above, no heat transfer from coolant channels experiencing flow stagnation was assumed to take place during the development of the updated figures.
In reality, a
substantial amount of heat transfer can occur from these areas.
With your concurrence, we commit to operate Fort St. Vrain using the updated figures on an interim basis until a formal Technical Specification change can be submitted and approved.
It is anticipated that the necessary documentation and independent reviews to support the formal change request will be completed by November 30, 1983. - We will make every effort to submit proposed changes by December 9, 1983.
Please note that these changes will not address " equal flow" condition operations with specific orifice valves not set in their equal flow positions. Those changes will be submitted separately in early 1984.
If you have any questions, please contact Mr. Chuck Fuller or Mr. Frank Novachek_ at (303) 785-2224.
Very truly yours,
- k..W$4 Don W Warembourg Manager, Nuclear Prod tion Fort St. Vrain Nuclear Generating Station DWW/dje Attachment
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Curves are from FGV Dperations Order 93-19.
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