ML20003D733
| ML20003D733 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/20/1981 |
| From: | Baynard P FLORIDA POWER CORP. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| 3-031-21, 3-31-21, TAC-43263, NUDOCS 8103310392 | |
| Download: ML20003D733 (6) | |
Text
-.. _
s
.4 h'Nf w..
Flos'ida F
g\\L9fl.xg c o..o....o.
March 20, 1981 7
U G
.4; u.,. Qk'*32Vmer30/gg A File:
3.B.13.A 3.A.3 A
ng 4
- 3-031-21
'A
>A s
Mr. Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation Nuclear Regulatory Conmission Washington, D.C. 20555
Subject:
Crystal River Unit '3 Docket No. 50-302 Operating License No. DPR-72 Nuclear Instrumentation (NI) Induced Error Transients
Dear Mr. Eisenhut:
Mr. Robert Rid's letter, dated January 14, 1981, to all operating Babcock and Wilcox reactor licensees concerning N1 induced error trans-ients requested that Florida Power Corporation provide justification for continued full power operation of Crystal River Unit 3.
Florida Power Corporation had evaluated this concern and securad the services of Babcock and Wilcox to conduct a study.
j One concern identified in Mr. Reid's letter was that postulated over-cooling events could result in the actual reactor power level exceeding 112% full power before a reactor trip as assumed in the safety analysis in the FSAR.
The evaluation by Babcock & Wilcox has shown that the increase in reactor power is offset by the beneficial effect of the l
temperature decrease on core thermal margins.
Departure from Nucleate l
Boiling Ratio (DNBR) analyses performed for the most limiting condition (of indicated power. at high flux trip limit of 105.5%, reactor coolant l*
pressure at the low pressure trip limit of 1800 psig) demonstrate that the minimum DNBR will be greater than 1.30 for conditions under which a reactor trip would be initiated at reactor power levels up to 150% of rated power.
In addition, Cycle 3 calculated core power distributions i
at all allowable rod index and Axial Power Shaping Rod (APSR) positions for normal full power operation were examined and margin existed for both Departure from Nucleate Boiling (DNB) and Centerline Fuel Melt (CFM) assuming an actual core power of 123%.
It is therefore concluded that the induced flux measurement error does not compromise the safe operation of Crystal River Unit 3 during overcooling events initiated from anywhere within the allowable operating range.
s103'3103%
[
General Othee 320i innt -fourin se,eet souin. P O Bor 14042. St Petersburg Horda 33733 813-8645151 f
.-_,Z~~~7_
T Mr. Darrell G. Eisenhut, Director Division of Licensing I-March 20, 1981 Page 2 i>
The concern on the rod ejection transient is that the high flux trip may not be activated for an ejected rod with a reactivity worth less than 0.2%ak/k.
Under these conditions, current models would show unaccept-able results (peak fuel enthalpy > 280 calories / gram).
Although no 4
. reanalysis has been performed, an engineering evaluation of the conser-l!
vatisms in the original analysis, such as adiabatic heatup, has led to l
the conclusion that a reanalysis using realistic assumptions will show l!
l{
that the peak fuel enthalpy will not exceed 280 calories / gram.
There-j fore, this concern is not censidered to compromise the safe operation of Crystal River Unit 3, Cycle 3.
f The specific response to the concerns stated in Mr. Reid's letter are i
attached.'
If you require any further discussion concerning our response, please contact this office.
Sincerely, FLORIDA POWER CORPORATION 1
L gl,
'1kW e
Patsy Y. Baynard Manager Nuclear Support Services Department Attachment 3
Klein(WO6)DN51-2 i
t
.I 1
1
,i
=.
ATTACHMF.NT Concern 1.
Determine if the high flux trip setpoint for your plant is affected j
by the accident-induced neutron flux errors discussed above.
Pro-vide us with information establishing that your present accident and transient analyses are valid and that the present Technical i
Specification limits provide as a minimum the original protective mar 9in derived from the safety analyses; if not, provide the fol-lowing information:
a.
Confirm that only the two non-205 FA plant concerns discussed in the B&W letter of October 29 affect your plant; namely, small overcooling events including a small steam line break, and a rod ejection accident.
Answer 4
Babcock and Wilcox's safety analysis study for a 205 fuel assembly plant indicated that there were three possible accident types that could induce a neutron transient error greater than the 2% full power transient error assumed in the safety analysis contained in-the Final Safety Analysis Report.
The three accident types were:
(1) Small overcooling and small steam line break.
(2) Large steam line break in containment.
(3) Rod ejection accident.
I Crystal River Unit 3's Reactor Protection System has a High Reactor Building Pressure Trip that adequately protects against the large steam line break accident.
Therefore, the only two concerns for Crystal River Unit 3 are the small overcooling events including a small steam line break and a rod ejection accident.
Concern 1.b.
Provide the effects of the error on your plant, supported by i
appropriate analyses.
Answer Although no transient analysis has been performed specifically for Crystal River Unit 3, an engineering evaluation analysis has shown that the maximum transient induced error for moderate frequency overcooling transients will be approximately 13%.
Therefore, for l
Klein(WO6)DN51-2 l
l
moderate frequency overcooling transients only, the instrumentation error that should be considered is:
0.8%
Heat Balance (The N!'s are recalibrated whenever their indicated power is less than 0.8% of the heat balance).
2.0%
Steady State Neutron Measurement.
O to 13% Transient Neutron Measurement Dependent on Coolant Temperatures.
0.5%
Instrumentation Error.
16.3%
Total.
To justify full power operation, one must demonstrate that opera-tion up to 122% full power is acceptable during these overcooling transients. This power level is based on a high flux trip setpoint of 105.5% full power plus a total error of 16.3%.
It should be remembered that 122% full power can only be reached during certain overcooling transients that provided specific core. conditions.
The analysis of induced flux errors during overcooling transients had led to the quantification of the ratios of indicated power to actual reactor power as a function of downcomer fluid temperature and reactor average coolant temperature. The primary concern is to determine the conditions that would pennit the actual reactor pow (r to exceed 112% without a reactor trip occurring.
The error calcu-lations were used to determine the maximum actual reactor power as a function of temperature for the case where the indicated power would be 105.5% which is the high flux trip setpoint. A series of heat balance calculations have been performed, using the minimum licensed RCS flowrate (374,880 GPM), to determine the corresponding reactor core operating conditions.
In order to quantify reactor core thermal margin for the conditions corresponding to operation at an indicated power level of 105.5%,
DNBR calculations were performed.
All 2200 psia Reactor Coolant System pressure points allowed by the Reactor Protection System and corresponding to operation with indicated power equal to 105.5% lie well above the Technical Specification minimum DNBR of 1.30 for the B&W-2 correlation (l).
For low pressures corresponding to the RPS low pressure setpoint (1800 psig), the variable low pressure trip provides protection to a minimum DNBR of 1.3 for reactor power levels up to 121% full power.
In addition, the high flux trip pro-vides protection to the minimum DNBR of 1.3 above 120% full power.
(1) B&W -2 correlation is documented in " Correlation of Critical Heat Flux in Bundle Cooled by pressurized Water," BAW-10000A, Babcock and Wilcox, Lynchburg, Virginia, May 1976.
Klein(WO6)DN51-2 l
4 Three dimensional power distribution calculations were performed to assess the - reactor power distribution perturbation at 123% full power due to an overcooling transient, and to determine the margins to centerline fuel melt and DNB limits.
Identical calculations were generated from normal steady-state operation at 100% full 1
power and from operation at 123% full power with a 16*F inlet temperature - reduction.
All power distribution calculations were initiated from within or near the normal rod index, APSR and axial imbalance limits of operation, such that the core behavior over the entire allowable operating range was examined.
In addition, these calculations acconnodate a proposed power level upgrade for Crystal River Unit 3 from 2452 to 2544 MWt.
Centerline fuel melt and DNB margins were computed for the 123%
full power cases to determine if core safety limits would be pre-served during an overcooling transient.
Since all calculations were performed from near steady-state condition, appropriate peak-ing factors were included in the 123% full power margins calcula-tions to account for potential peaking increases due to transient xenon and quadrant tilt.
Mdximum allowable peaking curves for CR-3, Cycle 3, were used to evaluate the DNB margins. The applic-ability of these curves at 123% full power was verified by DNBR i
analyses performed for the limiting cases.
The increased power level resulting from certain overcooling trans-ients can be acconnodated with the present Cycie 3 operating limits while allowing the upgraded power level of 2544 MWt.
The analysis of the most limiting peaking distribution yielded DNB margins in excess of that required to offset the 2.8% rod bow DNB penalty for this cycle.
The high flux trip frovides DNBR protection to the minimum DNB limit.
The concern on the rod ejection transient is that the high flux trip may not be activated for an ejected rod with a reactivity worth less than 0.2% k/k.
Under these conditions, current models would show unacceptable results (peak fuel enthalpy 280 calories / gram).
Although no reanalysis has been performed, an i
engineering evaluation of the conservatisms in the original analysis, such as adiabatic heatup, has led to the conclusion that a reanalysis using realistic assumptions, will show that the peak fuel enthalpy will not exceed 280 calories / gram.
Therefore, this i
concern is not considered to compromise the safe operation of Crystal River Unit 3, Cycle 3.
l l
Klein(WO6)DN51-2 l
e 4
Concern 1.c.
Provide your program and schedule for mitigating the effects of the error.
Answer The analysis performed to answer Question 1.b shows that the reactor core is adequately protectec c.c present and no additional i
analyses or equipment is required to mitigate the effects of the neutron flux induced errors identified in Question 1.a.
Concern 2.
Provide justification for continued full power operati')n of your plant until your program to mitigate effects of the error is com-3 l
pleted.
Answer The analysis done for Question 1.b shows that Crystal River Unit 3 can acconinodate a power level upgrade from 2452 to 2544 MWt at its present design configuration and its present Reactor Protection System.
No program to mitigate the effects of the induced neutron transient error is required.
)
i i
i i
1 1
i s
Klein(WO6)DNSI-2 9-
-awy--
v w-7 ar-g-
=T7' y
,r--
sy--
er
-e e
e my9-r-g
---*+-9 g-