ML20246N377

Application for Amend to License NPF-57,revising Tech Spec Table 4.3.6.1 to Increase Channel Functional Test Surveillance Intervals for Various Control Rod Block Instrumentation

ML20246N377
Person / Time
Site:	Hope Creek
Issue date:	05/05/1989
From:	Miltenberger S Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20246N382	List: ML20246N377 ML20246N460
References
NLR-N89054, NUDOCS 8905190419
Download: ML20246N377 (10)

Text

- _-. _ _ _ _ . . _ _ _ _ _ - - _ _ _ _ _ _ _ - _ _ _ - _

,.s + e Public Serv!ce Electric and Gas Company Stiven E. Mittenberger Public Service Electric and Gas Company P.O. Box 236. Hancocks Bridge, NJ 08038 609-339-4199 vice Prewent and chief Nucle OHicer May 5, 1989 NLR-N89054 United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Gentlemen:

REQUEST FOR AMENDMENT FACILITY OPERATING LICENSE NPF-57 HOPE CREEK GENERATING STATION DOCKET NO. 50-354 i

In accordance with 10 CFR 50.90, Public Service-Electric and Gas Company (PSE&G) hereby transmits a Request for Amendment to Facility Operat,ing License NPF-57 for Hope Creek Generating Station (HCCS). This amendment request revises Technical Specification Table 4.3.6-1 to increase the channel functional test surveillance intervals for various Control Rod Block instrumentation in accordance with General Electric Company (GE)

Licensing Topical Report (LTR) NEDC-30851P-A, Supplement 1 dated October 1988. The Technical Specification changes shown in Attachment 2 would permit various channel functional tests to be conducted quarterly rather than monthly in order to reduce the potential for unnecessary plant scrams, excessive equipment test cycles, and diversion'of test personnel and resources on unnecessary testing.

Attachment 1 provides sufficient justification to demonstrate that the proposed changes follow guidance contained in the LTR and the NRC Safety Evaluation Report (SER) on the topic dated i September 22, 1988. Based upon the justification provided, PSE&G believes that the proposed changes do not involve a significant <

hazards consideration pursuant to 10 CFR 50.92. In addition, since these changes simply reflect NRC-approved, generic changes l

contained in the LTR, a detailed NRC Branch review or specialist review should not be required and thus PSE&G believes that the proposed change can be processed as a Category 2, Item 4 amendment request.  !

[

l l

l I

8905190419 890505 7 7jI PDR ADDCK 05000354' P PNU i  ; 1 l

t Document Control Desk 2 05-05-89 NLR-N89054 In accordance with the requirements of 10 CFR 50.4(b)(ii), this submittal includes one (1) signed original, including affidavit, and thirty-seven (37) copies. In accordance with 10 CFR

50. 91 (b) (1) , a copy of this request has been sent to the State of New Jersey as indicated below. Upon NRC approval, please issue a License Amendment which will be effective upon issuance and shall l be implemented within 60 days of issuance. This latitude permits appropriate procedural modifications necessary to implement the proposed changes.

Should you have any questions or comments on this transmittal, do j not hesitate to contact us. 1 l

Sincerely, l

~

nA b w /

Affidavit Attachments (2)  ;

i l

I s

l l ..

Document Control Desk 3 05-05-89 NLR-N89054 i

C Mr. C. Y. Shiraki )

USNRC Licensing Project Manager ]

Mr. G. W. Meyer USNRC Senior Resident Inspector Mr. W. T. Russell, Administrator USNRC Region I Mr. K. Tosch, Chief New Jersey Department of Environmental Protection Division of Environmental Quality Bureau of Nuclear Engineering CN 415 Trenton, New Jersey 08625 1

f l-

f. .'

l' Ref: NLR-N89054 HCGS LCR 89-08 STATE OF NEW JERSEY )

) SS.

COUNTY OF SALEM )

f Steven E. Miltenberger, being duly sworn according to law desposes and says:

I am Vice President and Chief Nuclear Officer of Public Service

. Electric and Gas Company, and as such, I find the matters set forth.on our letter dated May 5, 1989 ., concerning the Hope Creek Generating Station, are true to the best of my knowledge, information and belief.

w r

Subscribed and Sworn Jo before me this 5YL day of /f/U/ , 1989 s

Notary Public of New Jersey BLEEN M. OCHS NOTARY PUBUC OF NEW JERSEY My Commission expires on A casiananima r.-g _... 4 I

i I

'l:; .

l I

l l

ATTACHMENT 1 PROPOSED CHANGE TO THE TECHNICAL SPECIFICATIONS HOPE CREEK GENERATING STATION FACILI1T OPERATING LICENSE NPF-57 NLR-N89054 DOCKET NO. 50-354 HCGS LCR 89-08' I. DESCRIPTION OF THE PROPOSED CHANGES Revise Hope Creek Generating Station (HCGS) Technical Specification (TS) Table 4.3.6-1, Control Rod Block Instrumentation Surveillance Requirements, to increase the channel functional surveillance test interval (STI) from monthly to quarterly for the following trip functions:

1

1. Rod Block Monitor:

a. Upscale

b. Inoperative

c. 'Downscale

2. Average Power Range Monitor (APRM):

a. Flow Biased-Neutron Flux - Upscale

b. Inoperative

c. Downscale

d. Neutron Flux - Upscale, Startup

5. Scram Discharge Volume:

a. Water Level - High (Float Switch)

6. Reactor Coolant System Recirculation Flow:

a. Upscale

b. Inoperative

c. Comparator Attachment 2 contains the revised Technical Specification page.

II. REASON FOR THE PROPOSED CHANGES The proposed changes reflect those standard TS revisions contained in NEDC-30851P-A, Supplement 1 (originally subhd tted by the BWR Owners' Group on June 23, 1986) which, based upon probabilistic analyses, justifies the identified time extensions by reducing the. potential for: (1) unnecessary plant scrams, (ii) excessive equipment test cycles, and (iii) diversion.of personnel and resources on unnecessary testing. The NRC staff has reviewed and approved this Licensing Topical Report (LTR) in  !

the letter and accompanying Safety Evaluation Report (SER) from C. E. Rossi (NRC) to L. N. Grace (BWR Owners' Group) dated September 22, 1988. .

Page 1 of 5

, Control Rod Block STI Increase HCGS LCR 89-08 l:

III. JUSTIFICATION FOR THE PROPOSED CHANGES PSE&G has justified extending the generic analysis completed by I

the BWR Owners' Group to HCGS by completing the required plant specific. analysis. The following discussion provides the information requested by the NRC staff in plant-specific submittals while'Section IV contains the Significant Hazards

! Consideration Evaluation completed pursuant to 10 CFR 50.92. As

-stated within the NRC's'SER for the LTR, two issues must be addressed to apply the generic analysis to individual plants when specific facility Technical Specifications are considered-for revision.

1. Confirm the applicability of the generic analyses of NEDC-30851P, Supplement'1 to the plant.

Response

Licensing Topical Report NEDC-30851P, Supplement 1, Appendix B identities PSE&G as a participating utility in the development of the Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation. In addition, Table 3-1 identities HCGS as a plant which enforces the rod block function through the Reactor Manual Control System (RMCS). PSE&G has reviawed the assumptions and design details contained in the NEDC-30851P, Supplement 1 and concluded that the report is applicable to and bounds the design of HCGS.

2. Confirm that any increase in instrument drift due to the extended-STIs is properly accounted for in the setpoint calculation methodology. (For additional information on this issue, see letter from C. E. Rossi to R. F. Janecek dated April 27, 1988).

Response

The guidance provided in the Rossi to Janecek letter indicated that:

"... licensees need only confirm that the setpoint drift which could be expected under the extended STIs has been .

studied and either (1) has been shown to remain within the existing allowance in the...setpoint calculation or (2) that the. allowance and setpoint have been adjusted to account for the additional expected drift."

The Rod Block Monitor and APRM trip functions were reviewed to determine whether the increased functional test interval affected the setpoint drift calculation. Calculation of the drift ot )

these trip function setpoints is based upon their channel i calibration interval of 6 months which is not affected by this proposed change. Similarly, the drift of the Scram Discharge Page 2 of 5 ]

~.

Control Rod Block STI Increase HCGS LCR 89-08 Volume trip function setpoint is based upon it's channel calibration interval of 550 days which is not affected by this proposed change. Therefore, it can be concluded that the setpoint drift for these three trip functions will remain within the existing allowanca in the setpoint calculation when the channel functional test interval is ircreased from monthly to quarterly.

The drift of the Reactor Coolant System Recirculation Flow trip i tunction setpoint is based upon the channel functional test interval which does vary with time. A review of the setpoint calculation W.t*,h an increased surveillance interval, from monthly to quarterly, nas been performed. Sufficient margin exists within the setpoint calculations to concluded that revisions to the current TS setpoints are not required. The drift that is expected under the extended STI will remain within the existing allowable margins.

In conclusion, PSE&G has determined that the proposed increases in the STI for the identified trip functions do not require any corresponding changes in the Control Rod Block setpoints. This conclusion was reached because the drift characteristics for the instrumentation with extended STIs are bounded by the current setpoint calculations. Hence the assumptions used in NEDC-30851P, Supplement 1 when the functional test interval is extended from monthly to quarterly can be applied to HCGS.

IV. SIGNIFICANT HAZARDS CONSIDERATION EVALUATION The proposed changes to the HCGS Technical Specifications:

1. Do not involve a significant increase in the probability or consequences of an accident previously evaluated.

As detailed in NEDC-30851P-A, Supplement 1 the sequence of events necessary for an unmitigated rod withdrawal error includes tailure of the Local Power Range Monitor (LPRM), failure of the Average Power Range Monitor (APRM) upscale trips, failure of the APRM upscale rod block, dual channel failure of the Rod Block Monitor, and of course operator failure to recognize and respond to any of these events. The BWR Owner's Group evaluated the impact of increased STIs on the probability of control rod block failure and concluded that a rather small increase in scram frequency results. However, both the absolute and relative increase is acceptably low and offset by the benefits of extending the Reactor Protection System (RPS) test intervals (see NEDC-30851P and the PSE&G to NRC submittal dated February 6, 1989.) Therefore, it can be concluded that the proposed changes do not involve a significant increase in the probability of an accident previously evaluated.

Page 3 of 5

1 l

.\

Control Rod Block STI Increase HCGS LCR 89-08 >

The consequences of an unmitigated rod withdrawal error were also addressed in NEDC-30851P-A, Supplement 1. Specifically, such an incident is very mild compared to the limiting reactivity l accident - a control rod drop accident. The BWR Owner's Group l indicated that the severity of a control rod drop accident bounds )

a rod withdrawal error due to the higher rate of reactivity j addition. HCGS Updated Final Safety Analysis Report (UFSAR) {

Section 15.4.9 and Tables 15.4-15 and 15.4-21 provides the plant j specific evaluation of the control rod drop accident. As a (

result the consequences of a rod withdrawal error were shown to be substantially leos than those associated with a dropped rod accident and less than 1% of the specified site boundary dose limits. This low consequence combined with the low probability of an unmitigated rod withdrawal error results in a negligible increase in risk which is offset by decreased risks associated with reduced testing of rod block and RPS instrumentation.

2. Do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The increased Control Rod Block surveillance test intervals do not alter the function of the instrumentation nor involve any type of plant modification. Additionally, no new modes of plant operation are involved with these changes. Therefore, it can be concluded that the proposed changes do not create the possibility of a new or different kind of accident than any accident previously evaluated.

3. Do not involve a significant reduction in a margin of safety.

The NRC staff has reviewed and approved the generic study contained in Licensing Topical Report NEDC-30851P-A, Supplement 1 and has concurred with the BWR Owners' Group that the proposed changes do not significantly affect the reliability or availability of the Control Rod Block instrumentation. Hence it can be concluded that the proposed changes do not adversely .

atfect plant safety margins. I V. CONCLUSION As discussed above, PSE&G has concluded that the proposed changes to the Technical Specifications do not involve a significant hazards consideration since the changes: (i) do not involve a significant increase in the probability or consequences of an accident previously evaluated. (ii) do not create the possibility of a new or different kind of accident from any accident previously evaluated, and (iii) do not involve a significant reduction in a margin of safety.

Page 4 of 5

Cont,rol Rod Block STI Increase HCGS LCR 89-08 In addition, PSE&G has addressed the two issues, In Item III above, which the NRC staff has indicated are necessary in order

-to. implement on a plant-specific basis the generic Technical Specification changes identified in NEDC-30851P-A, Supplement 1.

I I

Page 5 of 5 l

l

_ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ I

... ? n'

-;3 ji

p ,

e ATTACHMENT 2 HCGS LICENSE CHANGE REQUEST 89-08, NLR-N89064 CONTROL ROD BLOCK SURVEILLANCE TEST INTERVAL INCREASE The following Technical Specification page has been revised to reflect the proposed change:

Page 3/4 3-60 Table 4.3.6-1 p

i I

---

- - - _ - _ _ . - _ _ _ . _ _ _ - - - - - _ _ - . _ - - _ - - - _ _ _ - - - - - - _ - . _