05000341/LER-2016-012, Regarding Unanalyzed Condition for Control Rod Drop Accident at Low Reactor Power
| ML16356A289 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 12/20/2016 |
| From: | Polson K DTE Electric Company, DTE Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC-16-0073 LER 16-012-00 | |
| Download: ML16356A289 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(i) |
| LER closed by | |
| IR 05000341/2017001 (1 May 2017) | |
| 3412016012R00 - NRC Website | |
text
Keith I Posn Site Vice ident 6400 N. Diie Hih Newport, M1 48166 Tel: 734,586.4849 Fx 734.586.4172 Email: kihplo~teeg~o DTE Energy 10 CFR 50.73 December 20, 2016 NRC-16-0073 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Reference:
Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43
Subject:
Licensee Event Report (LER) No. 2016-012 Pursuant to 10 CFR 50.73(a)(2)(ii)(B), DTE Electric Company (DTE) is submitting LER No. 2016-012, Unanalyzed Condition for Control Rod Drop Accident at Low Reactor Power.
No new commitments are being made in this LER.
Should you have any questions or require additional information, please contact Mr. Scott A. Maglio, Manager -Nuclear Licensing, at (734) 586-5076.
Sincerely, Keith
. Polson Site Vice President
Enclosure:
Licensee Event Report No. 2016-012, Unanalyzed Condition for Control Rod Drop Accident at Low Reactor Power cc: NRC Project Manager NRC Resident Office Reactor Projects Chief, Branch 5, Region III Regional Administrator, Region III Michigan Public Service Commission Regulated Energy Division (kindschl@michigan.gov)
Enclosure to NRC-16-0073 Fermi 2 NRC Docket No. 50-341 Operating License No. NPF-43 Licensee Event Report (LER) No. 2016-012, Unanalyzed Condition for Control Rod Drop Accident at Low Reactor Power
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 101312018 (06-2016)
Estimated burden per response to comply with this mandatory collection request: 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
Reported lessons learned are incorporated into the licensing process and fed back to industry.
E EVENT REPORT (LER)
Send comments regarding burden estimate to the FOlA, Privacy and Information Collections LICENS E EBranch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by e-mail (See Page 2 for required number of digits/characters for each block) to Infocollects.Resource@nrc.gov, and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, DC 20503. If a (See NUREG-1022, R.3 for instruction and guidance for completing this form means used to impose an information collection does not display a currently valid OMB control http://www nrcaov/readinq-rrn/doc-collections/nuregs/staff/sr1022/r3/)
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- 3. PAGE Fermi 2 05000 341 1 OF 5
- 4. TITLE Unanalyzed Condition for Control Rod Drop Accident at Low Reactor Power
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED
- 1~ll FACiLITrY NAME DOCKET NUMBER MONTH DAY YEAR YEAR SEQUENTIAL REV MONTH DAY YEAR N/A 05000 I INUMBER NO.
/
50 FACILITY NAME DOCKET NUMBER 11 02 2016 2016 -
012 00 12 20 2016 N/A 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 20.2201(b)
Q 20.2203(a)(3)(i)
L 50.73(a)(2)(ii)(A)
Q 50.73(a)(2)(viii)(A) 1 Li 20.2201(d) 20.2203(a)(3)(ii)
/ 50.73(a)(2)(ii)(B) 50.73(a)(2)(viii)(B)
L 20.2203(a)(1)
Q 20.2203(a)(4)
Q 50.73(a)(2)(iii)
Q 50.73(a)(2)(ix)(A) 20.2203(a)(2)(i)
L 50.36(c)(1)(i)(A)
[
50.73(a)(2)(iv)(A)
[
50.73(a)(2)(x)
- 10. POWER LEVEL L 20.2203(a)(2)(ii)
Q 50.36(c)(1)(ii)(A)
Q 50.73(a)(2)(v)(A) 73.71(a)(4)
L 20.2203(a)(2)(iii) 50.36(c)(2)
L 50.73(a)(2)(v)(B)
L 73.71(a)(5) 20.2203(a)(2)(iv)
Ei 50.46(a)(3)(ii) 50.73(a)(2)(v)(C) 73.77(a)(1) 97 20.2203(a)(2)(v) 50.73(a)(2)(i)(A)
L 50.73(a)(2)(v)(D)
L 73.77(a)(2)(i)
L 20.2203(a)(2)(vi) j 50.73(a)(2)(i)(B)
] 50.73(a)(2)(vii) 73.77(a)(2)(ii)
L 50.73(a)(2)(i)(C)
Q OTHER Specify in Abstract below or in Gland Sealing Steam Exhausters (GSEs). This release path was not explicitly evaluated under the original plant design and licensing basis; however, its consideration is consistent with current NRC Regulatory Guides (RGs) 1.183 (regarding Alternate Source Term (AST) methods) and 1.195 (regarding non-AST methods) as well as industry precedent. Under normal full power operating conditions, the fraction of total steam production delivered to the GSEs is small (less than 0.2%). However, at low reactor power, when the consequences of a postulated CRDA are credible, the fraction of gland sealing steam relative to total steam production is relatively large (2-20%). Furthermore, this path does not benefit from iodine holdup and NRC precedent does not allow credit for iodine washout for this path. Consequently, the modeled release is greater than regulatory dose limits. Performed as directed by SRP 15.4.9, the current plant CRDA analysis evaluates only the consequences associated with a 1% volume/day release from the main condenser and includes credit for iodine retention in the vessel and main condenser. These are removal mechanisms that significantly reduce the consequences of the release. The only forced release path that is considered is that associated with the mechanical vacuum pumps (MVPs). However, the MVPs are provided with an automatic trip on Offgas 2-Minute Delay Pipe HI-HI Radiation that is credited to mitigate consequences from this forced release path. The release path from the GSEs was not evaluated and an automatic trip of the GSEs was not provided. Although UFSAR Section 15.4.9 does not evaluate or report MCR operator doses, UFSAR Section 6.4 states that the habitability design of the main control room is adequate to protect the control room personnel for all accidents.
There is no degraded component. Rather, the condition is being treated as the result of an inadequate design and licensing basis. The results of the updated analysis establish that in the absence of an installed means of rapidly and automatically terminating the release associated with the GSEs, regulatory dose limits are exceeded unless credit is taken for assumptions and features not currently part of the licensing basis, such as more realistic input assumptions regarding core source term and existing installed methods of mitigation, such as Main Steam Isolation Valve (MSIV) ((ISV)) isolation and operation of the CREF system.
The condition is specific to the analysis of the CRDA for which the postulated fuel damage occurs at time zero and resultant activity instantaneously transports to the coolant in the reactor vessel. Reactor fuel damage associated with other accidents such as the DBA LOCA occurs on a time scale that is longer than the time required for isolation of the steam lines and a release via the GSEs is not postulated to occur. On this basis, the condition is only applicable to the CRDA. In addition, the postulated fuel damage resulting from the CRDA is only credible during low-power (less than 10%)
MODE 1 and 2 operation during a plant startup. With increasing power, the severity of the rod drop reactivity insertion is reduced such that the expected peak enthalpy is much less than the design limit at 10% power even if the worst conceivable or maximum worth control rod is dropped from the core. Consistent with this expected power dependence of the consequences of the CRDA, the Rod Worth Minimizer is required to be operable for a reactor power less than or equal to 10% (per Technical Specification Table 3.3.2.1-1). The lower the assumed power level, the larger the source term fraction associated with gland sealing steam. The results of the updated CRDA analysis demonstrate that the consequences of CRDA lessen as the assumed power level increases due to the fact that as steam production increases, an increasing fraction of the source term is directed to the condenser where it is held up and decayed, and released at a reduced rate.
At the time of the discovery, Fermi 2 was operating at 97% power. Therefore, the plant condition was bounded by the design and licensing basis such that the condition did not exist at the time of discovery. No immediate actions were required. However, Fermi 2 had operated at low reactor power levels during startup several times in the past three years.
A past operability review for the three years prior to November 2, 2016 was performed to identify when Fermi 2 had operated at the relevant power levels (i.e. less than 10%). The instances are listed below:
- 1) At approximately 0130 EDT on March 28, 2014, Fermi 2 entered MODE 2 following a refueling outage. Power remained below 10% during startup until after approximately 0730 EDT on March 31, 2014. The duration in the relevant power range was approximately three days.
- 2) At approximately 1730 EDT on April 21, 2014, Fermi 2 entered MODE 2 following an outage. Power remained below 10% during startup until after approximately 0730 EDT on April 23, 2014. The duration in the relevant power range was approximately two days.
- 3) At approximately 0730 EDT on April 3, 2015, Fermi 2 entered MODE 2 following an outage. Power remained below 10% during startup until after approximately 0730 EDT on April 4, 2015. The duration in the relevant power range was approximately one day.
- 4) At approximately 1300 EST on November 25, 2015, Fermi 2 entered MODE 2 following a refueling outage. Power remained below 10% during startup until after approximately 0730 EST on November 28, 2015. The duration in the relevant power range was approximately three days.
- 5) At approximately 1630 EDT on May 12, 2016, Fermi 2 entered MODE 2 following an outage. Power remained below 10% during startup until after approximately 1930 EDT on May 13, 2016. The duration in the relevant power range was approximately one day.
The five periods identified above where Fermi 2 was operating in MODE 1 or 2 below 10% power represent potential unanalyzed conditions where the occurrence of a CRDA during those periods could have resulted in offsite and main control room doses exceeding regulatory limits. Therefore, this Licensee Event Report (LER) is being made under 10 CFR 50.73(a)(2)(ii)(B) as an "event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety." An 8-hour non-emergency Event Notification (EN 52342) was previously made under the corresponding requirement in 10 CFR 50.72(b)(3)(ii)(B).
Subsequent to discovery but prior to the date of this LER, Fermi 2 entered MODE 2 on November 11, 2016 following an outage that began on November 7, 2016. This startup was performed in accordance with plant procedures which had been revised based on an engineering analysis, described in detail below, and was, therefore, not an unanalyzed condition.
SIGNIFICANT SAFETY CONSEQUENCES AND IMPLICATIONS
An engineering analysis was performed to evaluate the safety consequences and implications. The analysis took credit for assumptions and features that, while justified, are not currently part of the licensing basis. Regarding offsite doses, the analysis determined that crediting automatic closure of the MSIVs during the CRDA is sufficient to ensure offsite doses remain within the limits of 10 CFR 100.11 and the more restrictive SRP 15.4.9 requirement that consequences for the CRDA remain less 25% of the Part 100 limits. Crediting MSIV closure is reasonable because it is an available ESF controlled by Technical Specifications (TS 3.3.6.1). Additional margin to these limits exists when typical post-shutdown decay of the core source term and post-scram reduction in steam flow are considered. Regarding MCR doses, the analysis took credit for automatic closure of the MSIVs as described previously and also for operation of the CREF since it is also an available ESF controlled by Technical Specifications (TS 3.7.3). Credit for these features alone was insufficient to ensure operator consequences within SRP 6.4. However, credit for these features in conjunction with either of the following conservatisms was sufficient to ensure that current NRC acceptance limits were met:
- - Credit for the last as-found control room unfiltered in-leakage, which is nearly one-half the current 173 cfm design value
- - Source term decay due to normal post-shutdown delay greater than 5.1 days in startup (normally days to weeks for any shutdown from full power)
Note that the 30-day calculated MCR post-CRDA dose is subject to a current regulatory acceptance limit of 5 rem whole body/30 rem thyroid based on Revision 2 of SRP 6.4. Revision 3 of SRP 6.4 currently allows adoption of a higher thyroid dose limit of 50 rem for current operating reactors conforming to and implementing the guidance of RG 1.195 in conjunction with RG 1.196. Since the analysis was performed in accordance with those RGs, the higher 50 rem thyroid dose limit was adopted for the analysis.
This LER identified five instances in the past three years that represented unanalyzed conditions. In each of those instances, the plant was in the relevant power range (i.e. less than 10%) during startup following an outage. The duration that the plant was shut down prior to startup was greater than 5.1 days for each of the five instances. Therefore, the actual plant condition for these five instances met the second bullet identified above such that MCR doses for a postulated CRDA would have met NRC acceptance limits. Note that credit for either one of the bulleted items is sufficient to provide adequate mitigation; however, both may be in effect, which provides additional margin. In addition, washout of iodine in the gland seal exhaust condenser, which would also significantly reduce the release of iodine, was not credited in the evaluation.
In conclusion, the engineering analysis determined that even if a postulated CRDA had occurred during the three years prior to the discovery of this unanalyzed condition, all relevant dose acceptance criteria for both offsite and MCR doses would have been met based on the plant condition and available ESF equipment at the time. Therefore, there was no adverse impact to public health and safety or to plant employees. There were no radiological releases.
CAUSE OF THE EVENT
The cause was that the original design and licensing basis analysis of the Fermi 2 DBA CRDA described in UFSAR Section 15.4.9 was performed in accordance with the prescribed methods and assumptions of NRC SRP 15.4.9 that did not evaluate the direct release of nuclear steam from the GSEs as significant. As a result, the plant was not designed to mitigate the potential consequences associated with this release path.
CORRECTIVE ACTIONS
As discussed previously, no immediate action was required to restore compliance since the condition did not exist at the time of discovery. As a long-term corrective action, Fermi 2 plans to implement modifications to eliminate this potential CRDA release path. To prevent future occurrence of an unanalyzed condition in the interim until the modification is complete, the Post Outage Startup Checklist was revised on November 10, 2016, to impose restrictions on the plant startup based on the engineering analysis described above. This checklist was utilized during the plant startup that was begun on November 11, 2016.
PREVIOUS OCCURRENCES
No similar previous occurrences were identified.Page 5
of 5