ML20042F983

Forwards Response to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47, 'Safety Implication of Control Sys in LWR Nuclear Power Plants.' Concurs & Endorses BWR Owners Group Analysis & Conclusion Re Plant Design

ML20042F983
Person / Time
Site:	Limerick
Issue date:	05/04/1990
From:	Helwig D PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-A-47, REF-GTECI-SY, TASK-A-47, TASK-OR GL-89-19, NUDOCS 9005100260
Download: ML20042F983 (7)

Text

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10 CPR 50.54(f)

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~' GL 89-19 PHILADELPHI A ELECTRIC COMPANY NUCLEAR GROUP HEADQUARTERS 955 65 CHESTERBROOK BLVD, WAYNE. PA 19087 5691 (315)6404660 D AVID M. HELWKb vees ensespout

.. .a . . . .. .. . . ... a e a n v i c e . .

May 4, 1990 Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 U.S. Nuclear Hegulatory Commission Attn Document Control Desk Washington, DC 20555

SUBJECT:

Limerick Generating Station, Units 1 and 2 Response to NRC Generic Letter 89-19, " Request for Action Related to Resolution of Unrosolved Safety Issue A-t,7 Safety laplication. of Control Gyaters in LWR Nuclear Power Plants"  !

REFERENCES:

(1) Letter frce J. G. Partlow (NRC) to 3 S. D. Floyd (BWROG), dated March 20, 1990 (2) Letter from D. R. Helwig (PECo) to U.S. Nuclear Regulatory Commission, dated March 20, 1990 (3) Letter from S. D. Floyd (BWROG) to J. G. Partlow (NRC), dated April 2, 1990 Gentlemen:

NRC Generic Letter 89-19 dated September 20, 1989, required licensees to provide, within 180 days of the date of the Getteric Letter, a statement as to whether licensees will implement the recommendations provided in Enclosure 2 of the Generic Letter. The Generic Letter also required licensees to provide a schedule for implementation of the recommendations and the basis for the schedule .,

'p or-to provide appropriate justification if the recommendations were not going to be implemented.

9005100260 900504 PDR 0 ADOCK 03000352 h P pyc z\

i Document Control De;k May 4, 1990 CL,89-19 P;ga 2 5

In Reference 1, the NRC approved a 45-day extension for submittal of the Generic Letter response. This extension was approved in order to allow individual utilities to review the generic response to be provided by the Bolling Water Reactors Ownerc' Group (BWROG) and verify its applicability to their plant (s). In Reference 2, Philadelphia Electric Company (PECo) stated our plans to exercise this extension for Limerick Generating Station. The BWROG response was provided to the NRC by Reference 3.

Accordingly, attached is our response to Generic Letter 89-19 for the Limerick Generating Station. The NRC recommendations are restated followed by our response.

If you have any questions, or require additional information, please contact us.

Very truly yours, s

I.ttacnment cc: T. T. Martin, Adiainistrator, Itegion I, USNRC '

T. J. Kenny, USNRC Senior Resident Inspector, LGS f

1 l

COMMONWEALTH OF PENNSYLVANIA : I ss.  !

COUNTY OF CHESTER  :

ll D. R. Helwig, being first duly sworn, deposes and says:

That he is Vice President of Philadelphia Electric Company, -

that he has read the response to Generic Letter No. 89-19, and knows i the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge, information and belief.

, 4 .c

( ,. -

Vice Presip ,,t .

V Subscribed and sworn to ,

before ree thj a '/^ <Aday of /11b 1990.

WA c) JJ kr/L Notary Public NOTARLAL SEAL CATHERINE A MENOEZ. Notary Pubtle Trecyttnn Two., Chester County gogssgppares Sept A 1983

.. .~-.

Limerick Gen 3rcting StCtien, Unito 1 and 2 '

, Corponto to Gen;ric Lettor 89-19

" Request for Action Related to Resolution of Unresolved Safety lasue A-47 Safety Implication i of Control Systems in LWR Nuclear Power Plants" r

NRC Recommendation (1)(a) i i

It is recommended that all GE boiling-water-reactor (BWR) i plant designs provide automatic reactor vessel overfill protection ,

to mitigate main feedwater (MPW) overfeed events. The design for the overfill-protection system should be sufficiently separate from the MPW control system to ensure that the MPW pump will trip on a reactor high-water-level signal when required, even if a loss of  ;

power, a loss of ventilation, or a fire in the control portion of r

the MPW control system should occur. Common-mode failures that could-disable overfill protection and the feedwater control system, 1

but would still result in a feedwater pump trip, are considered  ;

acceptable failure modes.

It is recommended that plant designs with no automatic  ;

reactor vessel overfill protection be upgraded by providing a .

commercial grade (or better) MPW isolation system actuated from at '

least a 1-out-of-1 reactor vessel high-water-level system, or justify the design on some defined basis.

PECO Response Philadelphia Electric Company has reviewed the Dolling

• Water Reactor Owners' Group (BWROG) generic responso provided to the NRC by letter from S. D. Floyd (DWROJ) to J. G. Partlov (NRC), dated April 2, 3990, and hae determined that it is applicable to Limerick Generating Station (LGS}. The installed reactot prescure vensel (RPV) automatic overflll protection configuration of LGS Units 1 and ,

2 corresponds to the "Orcup C Plants" described in Tanle 2 of the PWROG generic response. The LGS design is not in complete 1 accordance witn the recommendations identified in the Generic Letter since a fire in the auxiliary equipment room feedwater control system panel could disable reactor feedwater pump turbine trip logic which could ultimately result in a feedwater overfeed. We concur 1 with and endorse the BWROG analysis and conclusion that this design provides adequate RPV overfill protection and that any safety benefits from modifying this protection to be in complete accordance with the Generic Letter'would not be significant.

l NRC Recommendation (1)(a) Continued  !

In addition, it is recommended that all plants reassess their operating procedures and operator training and modify them if necessary to ensure that the operators can mitigate reactor vessel overfill events that may occur via the condensate booster pumps

'during reduced pressure operation of the system.

PECo Response i

Limerick Generating Station does not have condensate I booster pumps; however, the multi-stage condensate pumps installed l

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t Attcchment Page 2 at. LGS produce sufficient discharge pressure and capacity such that a booster pump _la not required. Our assessment assumes that the LGS ,

condensate pumps perform similarly to condensate booster pumps.

Reduced pressure operation, as it is discussed in the

' Generic Letter, is when kPV pressure is less than the shutoff head of the condensate pump, and, therefore, the condensate pumps can pump directly into the RPV. At LGS, the condensate pumps have a shutoff head.of approximately 600 pai. Pressures of less than 600

. psi in the reactor vessel can only occur during the following three scenarios: 1) during a major plant transient, 2) startup or 3) ,

shutdown. For each of these scenarios, PECo evaluated whether current operating procedures and operator training were adequate to ensure that operators can prevent and mitigate reactor vessel overfill events due to condensate pumps during reduced pressure operation.

The procedures that provide direction to mitigate RPV overfill events are Operational Transient (OT) Procedure OT-110,

" Reactor High Level", and Transient Response Implementation Plan (TRIP) Procedure T-101, " Reactor Pressure Vessel Control." These procedures are symptom based and include immediate and follow-up operator actions in response to high RPV water level conditions.

These procedures direct the operator t.o control RPV level and -

pressure within prescribed bands so ar to ulLigate or prevent overfilling the RPV.

Procedure OT-110 10 entered whenever there in an uneypected or unexplained rise in nPV water level, which could be 6:e to condensate-injection, This ptocedure is applicable whenever.E RPV overfill transient would be a conceru during normal cperation, plant transients, clartups And shutdownn. The operators are alerted to this condition by annunciation of tho Reactor H1-LO* alarm cr by visual observation of the RPV level instrumentation. The immediate operator action in to reduce feedwater flow until normal level is '

restored. Follow-up actions include cloning of the M41n Stuart  :

Inclation Valves (MSIVs) and the primary containment isolation valven for the High Pressure Coolant Injection (EPCI) and Reactor Core Isolation Cooling (RCIC) systems before the RPV water level l reaches the level of the steam lines.

TRIP Procedure T-101 provides direction to control and ultigate RPV overfill transients following nearly all plant shutdowns and major plant transients. This procedure specifically includes a step to control condensate pump injection during RPV pressure control actions to prevent inadvertent RPV overfill should-RPV pressure drop below the shutoff head of the condensate pumps.

In addition, there are several procedures that provide direction to prevent RPV overfill events by specifying the method of i

RPV level control and level setpoint. These procedures include l General Plant (GP) Procedure GP-2, " Plant Startup", GP-3, " Normal ,

l Plant Shutdown", and GP-5, " Steady State Operations", and TRIP l Procedures T-101, T-100, " Scram", and T-99, " Post-Scram

[ Restoration."

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Attcchment )

P ga 3 )

RPV overfill is addressed in three (3) portions of i operations training: Licensed Operator Training, Licensed Operator j Requalification Training and Simulator Training. l In the Licensed Operator Training Program the following have been incorporated to address this issuer ,

l. Licensed operator training lesson plans for the HPCI ,

System, RCIC System and Feedwater/Feedwater Level Control '

(PW/PWLC) Systems address the design and bases for high RPV water level trips. Included in the PW/PWLC system lesson ,

plan is a discussion of Institute of Nuclear Power 9 Operations' Significant Event Report (INPO SER) 5-88,

" Flooding Main Steam Due To Improper Reactor Water Level Control." '

2. Procedure OT-110 is covered with all licensed operator trainees.

t

3. TRIP Procedure steps preventing RPV overfill conditions are i covered with all licensed operator trainees.

For the Licensed Operator Requalification (LOR) Program, the following relate to RPV overfill:

1. Du-ing LOR Cycle 88-06, NRC Information Notice (INN) 88-77, -

" Inadvertent Reactor Vessel Overfill" and INPO SER S-88,

" Flooding Malu Steam Due To Improper Reactor Water Level l Control", including LGS procedural actions, which are taken  !

to prevent RPV overfill, were addressed.

2. Procecure OT-llo is covered with all licensed operators '

currently on au annoal basis.

3. LGS TRIP Procedures, which contain specific actions to prevent RPV overfill, is covered with all licensed i operators currently on an annual basis.

• LGS licensed operators have been trained in the simulator on scenarios that have the potential for overfilling or. uncontrolled >

flooding of the RPV. Events which have the highest probability for uncontrolled flooding are those in which the RPV depressurizes and water is injected by the condensate pumps when RPV pressure decreases below condensate pump shutoff head. Because this situation can only be prevented by operator intervention, a high degree of emphasis has been placed on detection and mitigation of these types of events in simulator training. Numerous simulator training scenarios have been used to address the issue of RPV overfill.

The basic philosophy and guidance to which the operators are trained and evaluated is to maintain RPV level between Level 3 and Level 8 (+12.5 inches and 454 inches) as delineated in the TRIP Procedures. Since the bottom of the main steam lines are at 118 inches, there is sufficient time to enable-the operators to implement corrective actions between notification of RPV high level and the point at which RPV level reaches the main steam lines.

Attcchment Page 4

,1,.

Based on this review, we conclude that current operator training and procedures are sufficient to ensure that operators can mitigate a RPV overfill event.

Further, a separate investigation was performed concerning a RPV rapid depressurization transient which could result in a overfill condition. As a result of this investigation, additional procedural enhancements vill be made to ensure that operators can prevent a RPV overfill event, specifically as a result of a rapid RPV depressurization. Details of the investigation and procedure enhancements will be included in a Special Report to the NRC expected to be issued by the end of August, 1990.

NRC Recommendation (1)(b) 4 It is recommended that plant procedures and Technical Specifications for all BWR plants with main feedwater overfill protection include provisions to verify periodically the operability of overfill protection and ensure that automatic overfill protection to mitigate main feedwater overfeed events is operable during power operation.

The instrumentation should be demonstrated to be operable by the performance of a channel check, channel functional testing, and channel calibration, including setpoint verification.

The Technical Specifications should include appropriate limiting conditions for operation (LCOs). These Technical Specifications  !

abould be commensurate with the requirements of existing plant Technical Specifications for channels that initiate protective actions. Previously approved Technical Specificctions for autveillance intervalo and limiting coaditions for operation (LCOs) for overfill protection are considered acceptable.

DECO Response The LGS Units 1 and 2 Technical Specifications (TS) have I been reviewed, and we have concluded that the TS include provisicus to ensure that overf411 protection is operable during power cperation. Technical Specifications Section 3/4.3.9 "r>3dwater/ Main Turbine Trip System Actuation Instrumentation," requires that the ,

overfill protection system be demonstrated operable by performance of a channel check, channel functional test, and channel calibration. In addition to these TS requirements, a Logic System Punctional test and Simulated Automatic Actuation are performed on an 18-month frequency. Consequently, no changes to the TS or plant Surveillance Test procedures are necessary.

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