ML20204F619

Forwards Analysis of SEP Topic XV-8, Control Rod Misoperation, Per Integrated Assessment 821115-19 Meetings. Acceptable Fuel Damage Limit for Single Control Rod Withdrawal Accident Not Exceeded.Topic Resolved

ML20204F619
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	04/25/1983
From:	Toner K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	Crutchfield D Office of Nuclear Reactor Regulation
References
TASK-15-08, TASK-15-8, TASK-RR NUDOCS 8305020202
Download: ML20204F619 (7)

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@ consumem C0mpBRy G.swral offices: 1945 West Parnall Road, Jackson, MI 49201 + (517) 788 0550

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Dennis M Crutchfield, Chief Operating Reactors Branch No 5 Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - SEP TOPIC XV-8 " CONTROL ROD HISOPERATION" - CONTROL R0D WITHDRAWAL ANALYSIS During the November 15-19, 1983 Integrated Assessment meetings conducted at the plant site, Consumers Power Company committed to perform an analysis to assess the magnitude of fuel damage resulting from withdrawal of a single control rod. As documented in a February 28, 1983 letter, Consumers Power Company indicated that an analysis of control rod misoperation utilizing reactor physics methods recently approved by the NRC would be. performed and submitted to the NRC. The attachment to this letter provides the afore-mentioned analysis which reveals that no fuel assemblies from Cycles 16, 17 or 18 had critical power ratios less than the 95/95 confidence limit of 1.1834.

Thus, the acceptable fuel damage limit for a single control rod withdrawal accident is not exceeded. Based on the results of the attached analysis, Consumers Power Company considers this topic resolved.

d W Kerry A Toner Senior Licensing Engineer CC Administrator, Region III, USNRC -

O NRC Resident Inspector-Big Rock Point j Attachment i

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ATTACHMENT Consumers Power Company.

Big Rock Point Plant Docket 50-155 SEP Topic XV-8 " Control Rod Misoperation" -

Control Rod Withdrawal Analysis April 25, 1983 l

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CONTROL R0D WITHDRAWAL ANALYSIS (Extracted From Reference 1)

This analysis describes the. effects of a. single control rod withdrawal on the

. Big Rock Point reactor. The rod withdrawal is caused by one of two possible events: (1) the locking collet fails and the rod will drift out of the core under the force of gravity; or (2) an equipment malfunction will cause a rod to be-withdrawn continuously when the operator attempts _to withdraw the rod one notch. ,

The analysis shows that no assemblies from Cycles 16, 17 or 18 had critical power ratios less than the 95/95 confidence level limit of 1.1834 (see.

Table 1). Thus, the Acceptable Fuel Damage Limit for this accident is not exceeded.

The analytical tool used for this calculation was GROK, a three-dimensional nodal reactor simulator with full thermal and hydraulic feedback. A descrip-

. tion of GROK is contained in the NRC approved (Reference 2) Big Rock Point Physics Methodology Report (Reference 3). To evaluate fuel damage limits the critical power ratio.was calculated using the XN-2 correlation (Reference 4).

An uncertainty factor was calculated for the critical power ratio (Reference

3) and was used in this analysis.

! The assumptions used in this analysis and the justification for their use 4 were:

1. The reactor is in equilibrium with respect to Doppler and in-channel void feedback mechanisms. This was allowed because the rod motion is slow when l- compared to the time constant for fuel heat transfer.

2. 'It has been sh'own that constant subcooling is.a less conserva.tive assump-tion than equilibrium subcooling; therefore, equilibrium subcooling is used'in this analysis (Reference 5). Equilibrium subcooling implies that

.the reactor power stays in equilibrium with the stea.r drum and the rest of the plant so that inlet subcooling stays in step with reactor p9wer.

- Subcooling does not change instantaneously with reactor power; rather, subcooling rapidly catches up with the power level and comes to equilibrium.

1 L 3. The feedback due to xenon is not allowed to vary. This is because the rod withdrawal is much faster than the time it would take for the xenon to come to equilibrium. If xenon was allowed to change, the final power level would.be depressed. Therefore, holding xenon constant (allowing a

! greater increase in power) is conservative.

4. The reactor protection system is not allowed to scram the reactor at 120 + 5% of rated power. Since the power increase is localized around the i rod, the ex-core detectors (which sense reactor power for the reactor protection system) may.not detect the power increase.

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r Control Rod Withdrawal Analysis 2 Big Rock Point Plant April 25, 1983

5. The initial condition that gives the lowest margin to fuel damage limits is at beginning of cycle at 102% of rated power (this includes a 2% heat balance error). This is because, at beginning of cycle, the core will have the highest rod density.

6. The final power level achieved (with the rod fully withdrawn) is'not known exactly. Therefore, a power level uncertainty is utilized. The uncer-tainty in the power level is found by using the following correlation:

NCE = - 0.00000409 (PLCE - 244.8) + CE where a) NCE = New Critical Eigenvalue b) PLCE = Power Level (MWt), with rod withdrawn, at the critical eigenvalue c) CE = Critical Eigenvalue, with rod withdrawn d) 244.8 MWt is 102% of rated power e) -0.00000409 Eigenvalue/MWt is from Reference 10 and it is the slope of a line (y=mx + b) where: x = core power and y = eigenvalue normalized to unity for Cycles 13-17, at BRP

7. The critical power ratios calculated by GROK have the 0.1531 uncertainty (Reference 3) subtracted from them.

The analysis started with a search for the control rod pattern that would be critical at 244.8 MWt at the beginning of Cycles 16, 17 and 18. The rod patterns searched were taken from those presented in the Final Physics Packages for Cycles 16, 17 and 18 (References 6, 7 and 8). Once the critical rod pattern was found, for each cycle, one of the pair of rods that is furthest inserted was pulled out of the core and the new power level was calculated. The critical eigenvalue was changed to the new critical l eigenvalue for an evaluation of fuel damage limits by the XN-2 critical power l ratio calculation.

l The analysis showed that no assemblies from Cycles 16, 17 or 18 had critical power ratios less than the 95/95 confidence level limit of 1.1834 (Reference 1). 95/95 means that, with 95% confidence, 95% of the fuel rods will not j experience DNB. With the rod fully withdrawn, reactor power increased to 309 1 MWt, 295.9 MWt, and 289.8 MWt for Cycles 18, 17 and 16 respectively.

Table 1 shows the five most limiting assemblies and their critical power i ratios, for each cycle.

l The critical power ratio for the 95/95 statement was derived from the data presented in the Exxon document XN-75-34, Rev 2 (Reference 4). The critical power ratio was calculated using the statistics of between set variations.

This statistical technique has been used in the NRC approved Exxon document XN-NF-512, Rev 1 (Reference 9).

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Control Rod Withdrawal Analysis 3 Big Rock Point Plant April 25, 1983 There are at least two factors not accounted for in this analysis that would minimize the effects of a rod withdrawal. One is that the operator will have at least one alarm to respond to (the bypass valve opening). Since the control rod position indication will remain operable, the operator should be able to diagnose the problem. The other factor is that the plant would probably not be able to support a power level much greater than 295 MWt.

- Either the feed pumps will trip on low suction pressure or the condenser will trip on low vacuum.

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Control Rod Withdrawal Analysis 4

-Big Rock Point Plant April'25, 1983 TABLE'1 Results of Analysis CYCLE ASSEMBLY CPR (including uncertainty) 16 G-219 1.4058 G-227 1.3819 G-205 1.3535 G-309 1.3507 G-222 1.3379 17 G-301 1.3461 G-311 1.3317 G-315 1.3156 G-412 1.2901 G-227 1.2436 18 'H-113 1.2952 G-410 1.2945 H-107 1.2678 H-214 1.2433 H-109 1.2190 t

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>s Control Rod Withdrewal Analysis 5 Big Rock Point Plant April 25, 1983 References

1. Design Review by Mike Reed; Analyze the Effects of a Single Control Rod Withdrawal on the Big Rock Point Reactor, UFI-740/22*13*20, PERFIL B.Z.GROK.830125.

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2. Supplemental Evaluation of Big Rock Point Physics Methodology; Docket No 50-155 (LS05-83-02-019), February 9, 1983.

3. Big Rock Point Physics Methodology Report; Rev 3, October 11, 1982, by Consumers Power Company.

4. XN-75-34, "The XN-2 Critical Power Correlation"; Rev 1, August 1, 1975, K Galbraith and J Jaech.

5. Letter to Dennis M Crtuchfield from David P Hoffman,

Subject:

BRP-Systematic Evaluation Program Topic IV-2, Reactivity Controls Systems Design and Protection Against Single Failures; Docket 50-155 -

License DPR-6, Date: May 4, 1981.

6. Big Rock Point Cycle 16 Physics Package, UFI-740/-22*13*32 PERFIL B.C.16.790202.

7. Big Rock Point Cycle 17 Physics Package, UFI-740/22*13*32 PERFIL B.C.17.801205.

8. Big Rock Point Cycle 18 Physics Package, UFI-740/22*13*32 PERFIL B.C.18.820312.

9. XN-NF-512, "The XN-3 Critical Power Correlation", Rev 1.

10. Design Review by Glen Seeburger; Drive A 95% Confidence Level Error Based for the Calculated. Void Coefficient, UFI-740/22*13*10.

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