05000483/LER-2018-002, Inadequate EOP Guidance for Asymmetric Natural Circulation Cooldown
| ML18184A389 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 07/03/2018 |
| From: | Wink R Ameren Missouri |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| ULNRC-06446 LER 2018-002-00 | |
| Download: ML18184A389 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 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)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability |
| 4832018002R00 - NRC Website | |
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Ameren MISSOURI Callava Plain July 3.2018 U LN RC-06446 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 50.73(a)(2)(i)(B),
50.73(a)(2)(v)(A),
50.73(a)(2)(v)(B),
50.73 (a )(2 )( v ) ( D)
Ladies and Gentlemen:
DOCKET NUMBER 50-483 CALLAWAY PLANT tJNIT I UNION ELECTRIC Co.
RENEWED FACILITY OPERATING LICENSE NPF-30 LICENSEE EVENT REPORT 201 $-002-t)0 Inadequate EOP Guidance for Asymmetric Natural Circulation Cooldown The enclosed licensee event report is submitted in accordance with 50.73(a)(2)(i)(B).
50.73(a)(2)(v)(A), 50.73(a)(2)(v)(B), and 50.73(a)(2)(v)(D) to report a deficiency in the Emergency Operating Procedure guidance for an asymmetric natural circulation cooldown that had the potential to have prevented the turbine driven auxiliary fedwater pump and atmospheric steam dumps from performing their specified safety functions.
This letter does not contain new commitments.
Wink.
Manager, Regulatory Affairs Enclosure Sincerely, P.O. Box 620 Fulton. MO 65251 AmerenMissouri. corn
ULNRC-(]6446 July 3,201$
Page 2 of 3 cc:
Mr. Kxiss M. Kennedy Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission
$201 NRC Road Steedman, MO 65077 Mr. L. John Kios Project Manager, Call away Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop O9E3 Washington. DC 20555-0001
ULNRC-06446 July 3,2018 Page 3 of 3 Index and send hardcopy to QA File A160.0761 Hardcopy:
Certrec Corporation 6100 Western Place, Suite 1050 Fort Worth, TX 76107 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)
Electronic distribution for the following can be made via LER ULNRC Distribution:
F. Nil. Diya B. L. Cox T. F. Hemriann S. P. Banker R. C. Wink T. B. Elwood Corporate Oversight Corporate Communications NSRB Secretary Performance Improvement Coordinator Resident Inspectors (NRC)
STARS Regulatory Affairs Mr. Jay Silherg (Pillsbury Winthrop Shaw Pittman LLP)
Missouri Public Service Commission
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 03/31/2020 iC4-20t Estimated burden per response to comply with this maniatory collection request 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> Reported lessons earned are incorporated into the lcensing process and fed buck to industry LICENSEE EVENT REPORT (LER)
Send comments regarding burden estimate to the Intormation Services Branch (1-2 P43).
U.S (See Page 2 for required number otdigits/characters foreach blk)
Nuclear Regulatory Commission, Washington.
DC 20555-0001, or by e-mail to lntocsllncts guy and to the Desk Otficer Office of Information and Regulatory Affairs NECB.t0202 (3t50-004)
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the NRC may not conduct or sponsor, and a person is not required to respond to the information collection
- 3. PAGE Callaway Plant Unit 1 05000483 1 OF 4
- 4. TITLE Inadeaiate EQP Guidance for Asvmmeiric,NaturaLCirculation Cooldown
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL I
AClLiTY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER MONTH DAY YEAR 05000 05 07 2018 2018
- - 002
- - 000
- - FACILITY NAME DOCKET NUMBER
- 9. OPERA11NG MODE
- 11. THIS REPORT IS SUBMtflED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1 El 20.2201(b)
El 20.2203(a)(3)(i)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
El 20.2201(d)
El 2D.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(B)
El 5D.73fa)(2)(viii)(B)
El 20.2203(a)(1)
El 2D.2203(a)(4)
El 50.73(a)(2)(Bi)
El 50.73(a)(2)fix)(A)
El 20.22D3(a)(2)(i)
El 50.36(c)(1)(i)(A)
El 50.73(a)(2)(iv)(A)
El 50.73(a)(2)fx)
- 10. POWER LEVEL El 20.2203(a)(2)(ii)
El 50.36(c)(1 )(ii)fA) 5D.73(a)(2)(v)(A)
El 73.71(a)(4)
El 20.22D3(a)(2)(iii)
El 50.36fc)(2) 50.73(a)f2)(v)(B)
El 73.71(a)(5) 100%
El 20.2203fa)(2)(iv)
El 50.46(a)(3)(ii)
El 50.73(a)(2)(v)fC)
El 73.77(a)(1)
El 20.2203(a)(2)(v)
El 5D.73(a)(2)(i)(A) 50.73(a)(2)(v)(D)
El 73.77(a)(2)(i)
El 20.2203(a)(2)fvi) 50.73(a)(2)(i)(B)
El 50.73(a)(2)(vii)
El 73.771a)(2)(ii)
El 50. 73(a )(2)(i)(C)
El OTHER Specify n Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER LICENSEE CONTACT TELEPHONE NUMBER llriclude Area Codet TB. ElWood, Supervising Engineer, Regulatory Affairs and Licensing 314-225-1905
- 13. COMPLETE ONE UNE FOR EACH COMPONENT F1LURE DESCRJBED IN THIS REPORT MANU-REPORTABLE MANU-REPORTABLE
CAUSE
SYSTEM COMPONENT FACTURER TO EPIX
CAUSE
SYSTEM COMPONENT FACTURER TO EPIX
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR YES (l(yes, complete 15. EXPECTED SUBMISSION DATE)
NO SUBMISSION kBSTRACT (Limit to 7400 spaces, i.e., approximately /5 single-spaced lypewfitten lines)
On May 7, 2018, during an engineering review of mission time requirements for Technical Specification related equipment, a deficiency was discovered regarding the Emergency Operating Procedure (EOP) guidance for natural circulation cooldown with a stagnant loop. This condition could be the result of a postulated Main Steam Line Break with a loss of offsite power.
During a natural circulation cooldown with a faulted steam generator, flow in the stagnant reactor coolant system (RCS) loop associated with the isolated faulted steam generator (SG) could stagnate and result in elevated temperatures in that loop. This could become an issue when RCS depressurization to residual heat removal system (RHR) entry conditions is attempted. The liquid in the stagnant loop would flash to steam and prevent RCS depressurization. In this condition, the ime required to complete the cooldown would be sufficiently long such that the nitrogen accumulators associated with Callaways atmospheric steam dumps and turbine driven auxiliary feedwater pump flow control valves would be exhausted.
s a result, the atmospheric steam dumps and turbine driven auxiliary feedwater pump would not be capable of performing heir specified safety functions of cooling the plant to entry conditions for RHR operation.
EOPs have been revised and the EOP change process will be revised to address this issue.
NRC FORM 3668 t04-20171 Page 1 of 4
1.
DESCRIPTION OF STRUCTURE(S), SYSTEM(S) AND COMPONENT(S):
The event reported in this LER involves procedural deficiencies in Callaways Emergency Operating Procedure (EOP) network that would have prevented the atmospheric steam dump and turbine driven auxiliary feedwater pump flow control valves from performing their specified safety function of cooling the plant to RHR entry conditions following a postulated main steam line break with a coincident loss of offsite power. The atmospheric steam dumps and turbine driven auxiliary feedwater pump flow control valves are supported by nitrogen accumulators that provide the safety-grade means of operating the valves. EOP guidance for an asymmetric natural circulation cooldown would have resulted in a cooldown duration that would have potentially exhausted the nitrogen accumulators, thus rendering the ASDs and TDAFP from performing a specified safety function.
Callaway Plant has four atmospheric steam dump valves (ASD5) tEllS: RV). There is one ASD in each main steam (EIIS:
SB) line outside of containment and upstream of the associated main steam safety valves and main steam isolation valve.
Thus, there is one ASD for each steam generator at Callaway. The ASDs, ABPV0001/2/3/4, are used to remove heat from the reactor coolant system (RCS) when the plant is being started up or shut down with the main condenser not available.
The turbine driven auxiliary feedwater pump (TDAFP) (EIIS system BA, component P) provides flow to all four steam generators. TDAFP flow to each steam generator is controlled by an air-operated flow control valve.
2.
INITIAL PLANT CONDITIONS
Callaway was at 100% Power/Mode 1 at the time of discovery of this event.
3.
EVENT DESCRIPTION
Prior to the initial startup of Callaway, the original Post-TMI symptom-based emergency response procedures used at Callaway did not consider the possible impact of the stagnation of an RCS loop when a steam generator is unable to remove heat (asymmetric cooldown) while on natural circulation. The lack of flow would cause the pressure and emperature in the faulted loop to remain elevated during the cooldown process, resulting in it acting as a pressurizer when he operators attempt to depressutize the RCS.
this issue was initially identified by the South Texas Project in 2003, and reported to the NRC by LER 03-006-00.
Subsequent to the South Texas Project LER, a generic industry resolution was developed by the Pressurized Water Reactor Owners Group (PWROG), This guidance was provided in 2007 by WCAP-1 6632-P. This document was intended o allow plants to avoid flow stagnation in the RCS loops by adhering to guidance on cooldown rate verses active loop differential temperature. This guidance was reviewed by Callaway and incorporated into the Callaway EOP network. The cooldown rate limits were incorporated as Figure 1 of procedure ES-0.2, Rev. 9, Natural Circulation Cooldown. The changes underwent an EOP Review, including a simulator validation of the revised EOP.
the incorporation of the PWROG guidance provided by WCAP-1 6632-P into Callaways EOP network was performed in accordance with existing administrative controls for EOP updates. The simulator validation did not cover the entire accident from start to finish, but terminated after the altered portions of the procedure were successfully diagnosed and entered. Neither the EOP Review nor the Simulator Validation detected the fact that the cooldown curve (Figure 1) would require cooldown rates that were sufficiently slow to result in cooldown durations that exceeded the values used in the licensing bases radiological consequence analyses and the mission time capabilities of the nitrogen accumulators associated with the atmospheric steam dumps and the turbine driven auxiliary feedwater flow control valves.
In 2018, during a Callaway engineering review of mission time requirements for Technical Specification related Structures Systems and Components (SSC), operating experience at another commercial nuclear facility was identified that led to a
- - eview of WCAP-16632-P.
It was determined that during low core decay heat situations, the revised ES-0.2 would require cooldown rate that was sufficiently slow such that the cooldown duration would exceed the capacity of the safety-grade dtrogen supply and be beyond the assumptions used in the licensing bases radiological consequence analysis for main steam line break.
Upon discovery of the deficiencies in the EOP guidance for an asymmetric natural circulation cooldown, the concern was entered into Callaways corrective action program and reported to the NRC as Event Notification EN 53388.
Dompensatory actions were issued to ensure the operability of affected SSCs and to maintain Callaway within its licensing ases.
4.
ASSESSMENT OF SAFETY CONSEQUENCES
The procedural deficiencies described in this LER represent an adverse condition of minor nuclear safety significance. During a natural circulation cooldown with a faulted steam generator, flow in the stagnant RCS loop associated with the isolated faulted SG could stagnate and result in elevated temperatures in that loop. This could become an issue when RCS depressurization to RHR entry conditions is attempted. The liquid in the stagnant loop would flash to steam and prevent RCS depressurization, In this condition, the time required to complete the cooldown would be sufficiently long that the nitrogen accumulators associated with Callaways atmospheric steam dumps and turbine driven auxiliary feedwater pump flow control valves would be exhausted. The atmospheric steam dumps and turbine driven auxiliary feedwater pump would not be capable of performing their specified safety functions of cooling the plant to entry conditions for RHR operation.
the extended cooldown duration for events involving an asymmetric natural circulation cooldown would result in an adverse impact on the sequence of events provided in the Callaway Final Safety Analysis Report (FSAR) licensing basis MSLB event. The licensing basis MSLB event described in Section 15.1 of the Callaway FSAR does not result in damage o the nuclear fuel cladding. Radiological consequences are driven by pre-existing fuel defects and iodine spiking associated with plant maneuvers. Specifically, the initial conditions used in the licensing basis MSLB analysis assume that pre-existing fuel defects have driven RCS concentrations of radio-iodines to the maximum levels permitted by Technical Specifications prior to initiation of the accident sequence.
It should be noted that Callaway has not had any fuel defects during the last three years. Another item of significance to note is that the most limiting Departure from Nucleate Boiling Rates (DNBR) values calculated for the licensing basis MSLB occur early in the accident sequence. Therefore, the long erm implications of the prolonged cooldown duration would not adversely impact the limiting MSLB DNBR values.
It only SSCs and actions available in a licensing basis safety analysis scenario ate credited, it is postulated that the nitrogen accumulators would be exhausted prior to reaching RHR entry conditions. Should this occur, the ASDs would close, and the plant would heat back up to the saturation conditions associated with the lift setpoint for the main steam safety valves (MSSV5). Plant operation in this condition would not result in damage to the fuel clad. The plant would remain in this condition until reasonable recovery actions not typically credited in licensing basis safety analysis could be aken. Following a MSLB event, radiological conditions would not restrict movement about the plant site. Reasonable recovery actions that would be taken to resolve the condition would include replenishing the Condensate Storage Tank (CST) and/or Hardened Condensate Storage Tank (HCST) inventories. The plant air compressors are non-Technical Specification, non-safety related components located in the Turbine Building. Therefore, credit is not taken for the air compressors in the licensing basis safety analysis. However, an air compressor is provided with electrical power from a safety-grade supply and cooling from the essential service water system. The Emergency Operating Procedure network includes guidance that directs the operators to start an instrument air compressor following entry into the Emergency Operating Procedures. Therefore, it is probable that a plant compressor would be available following a Design Bases ccident. Availability of the instrument air system would serve to significantly extend the availability of the atmospheric steam dumps. Use of the FLEX air compressors would also represent a reasonable recovery strategy that would significantly extend the service time of the atmospheric steam dumps.
Based on these considerations, it is concluded that the procedural deficiencies described in this LER do not represent a condition that has mote than minor safety significance.
5.
REPORTING REQUIREMENTS
This LER is submitted pursuant to 50.73fa)(2)(i)(B), 50.73(a)(2)(v)(A), 50.73(a)(2)(v)(B), and 50.73(a)(2)(v)(D).
The prolonged cookiown duration potentially caused by the inadequate OP guidance could have resulted in the As and TDAFP being incapable of performing a specified safety function following a postulated asymmetric natural circulation cooldown. Specifically, the ASDs and TDAFP would not have been capable of cooling the plant to RHR entry conditions following a main steam line break with a coincident loss of offsite power. SSCs incapable of performing their specified safety functions are considered to be inoperable. The period of inoperability for the ASDs and TDAFP exceeded the allowances of Technical Specifications 3.7.4, Atmospheric Steam Dump Valves (ASDs), and 3.7.5, Auxiliary Feedwater (AFW) System. Thus, this event is being reported as an operation or condition prohibited by Technical Specifications in accordance with 50.73(a)(2)(i)(B).
As explained above, the prolonged cooldown duration potentially caused by the inadequate EOP guidance would have resulted in the ASDs and TDAFP being incapable of performing a specified safety function following a postulated asymmetric natural circulation cooldown. Specifically, the ASDs and TDAFP would not have been capable of cooling the plant to RHR entry conditions following a main steam line break with a coincident loss of offsite power. Thus, this event is being reported as a condition that could have prevented fulfillment of a safety function in accordance with 50.73(a)(2)(v)(A), 50.73(a)(2)(v)(B), and 50.73(a)(2)(v)(D).
6.
CAUSE OF THE EVENT
The genetic accident mitigation strategy provided by WCAP-1 6632-P was not appropriate for Callaway. The EOP review and validation process in use at the time when WCAP-1 6632-P was issued was not sufficient to identify the flaws in the WCAP-J 6632-P accident mitigation strategy prior to that strategy being incorporated into Callaways EOP network.
7.
CORRECTIVE ACTIONS
A revised accident mitigation strategy has been developed to preclude excessive cooldown durations during an asymmetric natural circulation cooldown. This strategy has been incorporated into Callaways EOP network. Additionally, Callaways administrative procedures for the EOP program will be revised to provide an enhanced level of review to increase the likelihood of identifying deficiencies in generic industry guidance prior to the incorporation of the generic guidance into Callaways EOP network.
8.
PREVIOUS SIMILAR EVENTS
A review of LERs from the past three years found no other events in which the failure to detect deficiencies in generic industry guidance regarding EOP content led to a reportable condition.