ML20042F389

Responds to Generic Ltr 89-19, Request for Action Re Resolution of USI A-47, 'Safety Implication of Control Sys in LWR Nuclear Power Plants.'

ML20042F389
Person / Time
Site:	FitzPatrick
Issue date:	04/30/1990
From:	Brons J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-A-47, REF-GTECI-SY, TASK-A-47, TASK-OR GL-89-19, JPN-90-036, JPN-90-36, NUDOCS 9005080267
Download: ML20042F389 (8)

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April 30,1990 i JPN 90036 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Mall Stop P1137 Washington, D. C. 20555  ;

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Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50 333 Generic Letter 8919 Safetyimplications of i Control Systems in LWR Nuclear Power Plants

References:

1. NRC Generic Letter 8919, " Request for Action Related to Resolution of Unresolved Safety lasue A 47, ' Safety implication of Control Systems in LWR Nuclear Power Plants, oursuant to 10 CFR 50.54(f)," dated l September 20,1989.

2. NYPA letter, J. C. Brons to USNRC, dated March 19,1990 (JPN 90-014/IPN.90024) regarding Generic Letter 8919.

3. BWROG letter, S. D. Floyd to J. G. Partlow, (NRC) dated April 2,1990 (BWROG 9048) transmits BWROG response to GL 89-19.

Dear Sir:

In Generic Letter 8919 (Reference 1), the NRC recommended that the Authority reassess and, if necessary, modify the existing reactor overfill protection system at the James A. FitzPatrick Nuclear Power Plant. In Reference 2, the Authcgity rescheduled its response to GL 8919 to permit

, our review of a generic response prepared by the BWROG (Bolling Water Reactor Owners' Group). The BWROG report was submitted to the NRC as an attachment to Reference 3.

The Authority has reviewed the BWROG report anci has determined that it is applicable to the FitzPatrick plant. The Authority endorses Reference 3.

The BWROG report summarized the design of installed overfill protection at several BWRs, estimated the cost of hardware modifications recommended by GL 89-19, and reviewed relevant operational experience for BWRs throughout the world. The BWROG report concludes that the '

recommended modifications are expensive and do not contribute significantly to overall plant C

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, safety. Operational experience does not suggest that improvements to the existing system are l warranted. No feedwater control system common mode failures were identified by the BWROG that might have resulted in reactor vessel overfill.

l Based on this report, the Authority believes that the FitzPatrick's existing overfill protection is ,

adequate. No hardware modifications are planned in connection with this issue.

l l' The Authority recognizes the importance of a reliable overfill protection system. To confirm the conclusions of the BWROG report, the Authority will expand the FitzPatrick IPE/PRA to quantify i the risks associated with vessel overfill and feedwater control system failures, if the results of the

! IPE/PRA Indicate that the existing overfill protection system is a significant contributor to overall risk, modifications will be considered.

Plant shutdown and startup procedures will be changed to assure system operability as recommended by GL 8919. Surveillance procedures already adequately assure overfilt protection  !

equipment operability. Simulator training for licensed operators includes feedwater control

, problem scenarios. FitzPatrick's Technical Specifications require periodic verification of overfill protection system operability. No additional overfill protection LCOs or STs are being considered. -

The Authority's complete response and further details are included in Attachment 1. Should you or your staff have any questions regarding the Authority's plans or position, please contact ,

Mr. J. B. Ellmers of my staff.

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i Very truly yours,

/ bn C. ELrons Executive Vice President Nuclear Generation STATE OF NEWYORK COUNTY OF WESTCHESTER Subscribed and sworn to before me .

this Som day of ApevL 1990.  ;

b 4A4L f Notary Public

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cc: Next page

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cc: . U. S. Nuclear Regulatory Commission Region 1 475 Allendale Road King of Prussia, PA 19406 Office of the Resident inspector U. S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 Mr. Dave LaBarge Project Directorate 11 Division of Reactor Projects.1/II -

U. S. Nuclear Regulatory Commissloi.

Mall Stop 14 B2 .

Rockville, MD 20555 b

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NEW YORK POWER AUTHORITY James A. FitzPatrick Nuclear Power Plant 1.0 Background and Introduction in Generic Letter 8919 (Reference 1), the NRC recommended that the Authority reassess and modify existing reactor overfill protection at the James A. FitzPatrick Nuclear Power Plant. These recommendations were a direct result of the staff's recent resolution of Unresolved Safety issue (USI) A-47,

• Safety implications of Control Systems in LWR Nuclear Power Plants."

a Generic Letter 8919 (GL 8919) also requires that the Authority submit, under oatt: or affirmation, a statement to the NRC as to whether the recommendationt, of t @n .re 2 to GL 89-19 will be implemented, a schadule for the implementation of any modificatgr ;, or changes, and th9 basis for that schedule.

This attachment constitutes the Authority's response.

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1.1 BWROG on Hardware Change Recommendations The Authority participated in the BWROG committee on GL 8919. The committee prepared a report which has been submitted to the NRC staff (Reference 3). The Authority has reviewed this report and has determined that it is applicable to the FitzPatrick plant. The Authority endorses this report. 1 The BWROG report summarized the design of Installed overfill protection at several BWRs, estimated the cost of hardware modifications recommended by GL 89-19, and reviewed relevant l operational experience for BWRs throughout the world. The BWROG report concludes that the recommended modifications are expensive and do not contribute significantly to overall plant safety. Operational experience does not suggest that improvements to the existing system are warranted. No feedwater control system common mode failures were identified by the BWROG that might have resulted in reactor vessel overfill.

Based on this report, the Authority believes that the existing overfill protection system is adequate.

No hardware modifications will be installed at this time.

1.2 Incorporation into FitzPatrick IPE/PRA Program The Authority recognizes the importance of a reliable overfill protection system. To confirm the conclusions of the BWROG report, the Authority will expand the FitzPatrick IPE/PRA to quantify the risks associated with vessel overfill and feedwater control system fallures. If the results of the IPE/PRA indicate that the existing overfill protection system is a significant contributor to overall  !

risk, modifications will be considered. '

. Attachment I Response to NRC Generic Letter 89-19

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Although the Authority did not identify that it would resolve USI A 47 as part of the RtzPatrick '

IPE/PRA in Reference 6, we estimate that this additional work can be completed without delaying l the IPE/PRA.

i 1.3 Plans and Schedules l If, as result of the IPE/PRA, the reactor overfill protection system is determined to be a significant ,

contributor to overall risk, the Authority will submit plans and schedules for appropriate modifications one hundred and twenty days after the IPE/PRA results have been sent to the NRC. .

Th!s report will include a description of the proposed modification (s), along with plans and .

schedules for

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plant modification (s).

l operator training, surveillance procedures test procedures,  ;

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plant operating procedures, and

. technical specification changes.  ;

, As described in the Authority's 60-day IPE/PRA response (Referer.ce 6), the Authority will submit the FitzPatrick IPE/PRA by the end of this year (December 30,1990). Based on this date, plans i

! and schedules for any necessary overfill protection system modifications will be submitted by P April 30,1991.

The Authority's IPE/PRA program was recently approved by the NRC in Reference 10.

2.0 Une-by Une Response to GL 8919 -

The paragraphs below respond line by line to the five specific lasues detailed in GL 89-19 for the l FitzPatrick plant.

2.1 Section 1(a) Evaluation of Overfill Protection Group Classification (1, ll, Ill): Class I. Automatic overfill protection trips the main feedwater pumps in a *2 out-of 3" logic system. .

Significant Deviations from Recommendations: The existing overfill protection system is not ,

electrically independent or physically separate from the main feedwater control system. One of (

the three reactor vessel level transmitters is devoted exclusively to overfill protection; the two l remaining transmitters also provide clgnals to the main feedwater control system.

l l Implementation Schedule: No hardware modifications to the existing overfill protection system

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y are warranted at this time. 3 p

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,, Attachment i

. . Response to NRC Generic Lett7 8919 Review of Plant Operating Procedures L Existing Procedures: Plant operating procedures for the High Pressure Coolant injection (HPCI),

Reactor Core Isolation Cooling (RCIC) and the Condensate /Feedwater Systems were reviewed to l assure that they adequately address reactor vessel overfi1l protection. Based on their design and flow capabilities, only these three systems could potentially overfill the reactor vessel (The i

Control Rod Drive Hydraulic System is a low flow system and the low pressure emergency core cooling systems are low head systems.)

i A caution statement will be added to FOP 65,

• Start up and Shutdown Procedure,' (Ref.11), This operating procedure provides detailed instructions for reactor start up and shutdown.

This new cautlon will alert operators of the potential of over filling the vessel with the condensate l booster pumps during startup and shutdown while the reactor vessel is at reduced pressure. The remaining plant procedures adequately address existing reactor vessel overfill protection.

l Implementation Schedule: This caution statement will be added to FOP 65 not later than June 30,1990.

i Operator Training Existing Training Program: Existing operator training programs adequately address reactor vessel overfill protection. Simulator training for licensed operators already includes feedwater control system problem scenarios. Emergency Operating Procedure training scenarios include situations when the reactor is at reduced pressure and the condensate booster pumps are operating. During EOP training, operatars practice controlling reactor level and compensating for ,

incteasing level, implementation Schedule: No changes to the operator tralning program are required.  ;

i 2.2 Section 1(b) Surveillance and Test Procedures Existing Procedures: Plant surveillance and test procedures already include provisions to

! periodically demonstrate overfill protection operability. Procedures are also already in place to assure that the existing overfill protection system is operable during power operation. ,

Implementation Schedule: No changes to the existing plant surveillance and test procedures i are required.

Technical Specif; cations Existing Specifications: The LCOs and surveillance requirements identified below periodically verify the operability of the existing overfill protection and assure that automatic overfill protection is available to mitigate main feedwater overfill events during reactor power operation.

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. Attachment i 1

. Response to NRC Generic Letter 8919  :

i The following portions of the FitzPatrick Technical Specifications address existing automatic overfill protection:  ;

i Umiting Conditions for Operation l Umiting Conditions for Operation (LCO) 3.2.B,

• Core and Containment Cooling Systems .

Initiation and Control," page 49.

Table 3.2 2, *lnstrumentation That initiates or Controis the Core and Containment Cooling Systems", item 3, *High Reactor Water Level," page 66.  !

Surveillance Tests l Surveillance Requirement 4.2.B, " Core and Containment Cooling Systems Initiation and i Control," page 49. '

Table 4.2 2,

• Minimum Test and Calibration Frequency for Core and Containment Cooling .

Systems,' Item 1, ' Reactor Water Leve!," page 79. '

i implementation Schedule: No changes to the existing specifications are necessary at this time. .

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.- ,. Responsa to NRC Generic Letter 8919 3.0 References

1. NRC Generic Letter 8919,

• Request for Action Related to Resolution of Unresolved Safety lasue A 47, ' Safety implication of Control Systems in LWR Nuclear Power Plants,' Pursuant to 10 CFR 50.54(f)," dated September 20,1989,

2. NYPA letter, J. C. Brons to USNRC, dated March 19,1990 (JPN 90-014/lNP 90-024) regarding Generic Letter 89-19.

3. BWROG letter, S. D. Floyd to J. G. Partlow, dated April 2,1990 (BWROG 9048) transmits BWROG report EDE 07 0390/DRF A00-03773, *BWROG Response to GL8919, Enclosure 2, Hardware Change Recommendations,' Rev. 0.

4. NRC Generic Letter 88 20, " Individual Plant Examination for Severe Accident Vulnerabilities," dated November 23,1988.

5. Supplement 1 to GL 88 20, dated August 29,1989 regarding initiation of the individual plant examination for severe accident vulnerabilhies 10 CFR 50.54(f).

6. NYPA letter, J. C. Brons to USNRC, (JPN 89-089/IPN 89-066) dated October 27,1989.

Submits 60-day response to GL 88 20 for both FitzPatrick and Indian Point 3.

7. NUREG 1217,

• Evaluation of Safety implications of Control Systems in LWR Nuclear Power Plants, Technical Findings Related to USl A 47, Final Report," published June 1989.

8. NUREG 1218, " Regulatory Analysis for Proposed Resolu'lon of USI A 47, Safety implications of Control Systems, Draft Report for Comment," published April 1988.

9. NUREG/CR-4262, " Effects of Control System Failures on Transients and Accidents at a General Electric Bolling Water Reactor," including both Vols.1 and 2, main report and ,

appendices, respectively.

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NRC letter, D. E. LaBarge to J. C. Brons, dated January 30,1990 regarding review of 60-day  !

response to GL 88 20.

Concludes that the FitzPatrick IPE/PRA approach, methodology and schedule are acceptable.

11. FitzPatrick Operating Procedure FOP 65, " Start up and Shutdown Procedure.'

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