ML20236N703

Forwards Request for Addl Info Re Licensee Amend Request, Dtd 980227.Response Requested within 30 Days of Date of Ltr

ML20236N703
Person / Time
Site:	Prairie Island
Issue date:	07/02/1998
From:	Connaughton K NRC (Affiliation Not Assigned)
To:	Richard Anderson NORTHERN STATES POWER CO.
References
TAC-MA1675, TAC-MA1676, NUDOCS 9807150221
Download: ML20236N703 (5)

Text

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J Mr. Roger O. Anderson, Director July 2, 1998 Nuclear Energy Engineering Northern Stater Power Company 414 Nicollet Mall Minneapolis, Minnesota 55401

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION ON PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2, LICENSE AMENDMENT REQUEST DATED FEBRUARY 27,1998, "(TAC NOS. MA1675 AND MA1676)

Dear Mr. Anderson:

By letter dated February 27,1998, Northern States Power Company (NSP) committed to modify the Anticipated Transient Without Scram (ATWS) Mitigating System Actuating Circuitry (AMSAC) on Prairie Island, Units 1 and 2, with the addition of a diverse scram system (DSS).

Additionally, NSP requested amendments to facility operating licenses DPR-42 and DPR-60 to reflect this modification. The proposed modification and amendment request resulted from an NSP evaluation of the capability of Prairie Island, Units 1 and 2, to mitigate a complete loss-of-feedwater A1WS. The preliminary results of this evaluation were discussed with the NRC staff in a meeting held on July 9,1997, and the final results were submitted to the NRC staff by letter dated July 23,1997. As discussed in a meeting between the NRC staff and NSP on April 29,1998, the February 27,1998, submittal provided only a portion of the information that will be necessary to support the requested license amendments. Supplementary submittals are expected as DSS detailed functional and physical design information is finalized. -

Based upon its review of the information provided to date, the staff has identified a number of

. areas where supplemental information should be provided as soon as it is available in order for the staff to continue to support the requested schedule for amendment issuance. Accordingly, the NRC requests that you respond to the enclosed request for additional information within 30 days of the date of this letter.

Sincerely, ORIGINAL SIGNED BY Kevin A. Connaughton, Project Manager Project Directorate lil-1 p8g g S M 82 Division of Reactor Projects - Ill/IV P- PDR Office of Nuclear Reactor Regulation ,

Docket Nos. 50-282,50-306

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Enclosure:

Request for Additional Information I

cc w/ encl: See next page j DISTRIBUTION.

Docket File - ACRS EAdensam PUBLIC OGC JMcCormick-Barger, Rlll i PD3-1 RF KKavanagh Q (( {@}] fy%f !

" DOCUMENT NAME:G:\WPDOC\ PRAIRIE \PIDSS.RAI Ta receive a copy of this (gcuplent. Indicate in the box: "C" = Copy without enclosures "E" = Copy with enclosures "N" = No copy OFFICE PM:PD3 K'F E LA:PD3-1 l_E, SRXB HICB_ , f PD:PD3-1 ,

NAME KConnatigiton:db CJamerson & RCaruso //f/) JMaueltf(// CACarpentercde l DATE 06/74 /98 O(/ A /98 I Ql)/ 7 /98 /h@U98 06/a /98 OFFitlAl/ RECORD COPY a . . . _ _ _ _ . _ - . .

- s Mr. Roger O. Anderson, Director Prairie Island Nuclear Generating Northern States Power Company Plant cc:

J. E. Silberg, Esquire Site Licensing Shaw, Pittman, Potts and Trowbridge Prairie Isiand Nuclear Generating 2300 N Street, N. W. Plant Washington DC 20037 Northem States Power Company 1717 Wakonade Drive East Plant Manager Welch, Minnesota 55089 Prairie Island Nuclear Generating Plant Tribal Council Northem States Power Company Prairie Island Indian Community 1717 Wakonade Drive East ATTN: Environmental Department Welch, Minnesota 55089 5636 Sturgeon Lake Road Welch, Minnesota 55089 Adonis A. Nebiett Assistant Attomey General Office of the Attomey General 455 Minnesota Street Suite 900 St. Paul, Minnesota 55101-2127 U.S. Nuclear Regulatory Commission Resident inspector's Office l 1719 Wakonade Drive East j Welch, Minnesota 55089-9642 i Regional Administrator, Region 111 .

U.S. Nuclear Regulatory Commission 801 Warrenville Road l Lisle, Illinois 60532-4351 l

Mr. Stephen Bloom, Administrator J Goodhue County Courthouse Box 408 Red Wing, Minnesota 55066-0408 Kris Sanda, Commissioner Department of Public Service l 121 Seventh Place East Suite 200 St. Paul, Minnesota 55101-2145 June 1998 l

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.4 REQUEST FOR ADDITIONAL INFORMATION ON ATWS ANALYSES TO SUPPORT THE DESIGN OF THE DIVERSE SCRAM SYSTEM (DSS)

PROPOSED BY NSP FOR PRAIRIE ISLAND UNITS 1 AND 2 (DOCKET NOS. 50-282 AND 50-305)

The proposed design modification of AMSAC/ DSS shown in an initial submittal dated 4 February 27,1998, is considered a major design change because it involves changes in the l entire AMSAC system design, including inputs, outputs, logic, power supplies, system hardware, system software, operator interface, and the resulting procedures. As a result, the NRC staff will review the modified AMSAC system as a new design. Accordingly, the staff requests that, in addition to a detailed description of the proposed design modification, the licensee should address the following in its final submittal for the proposed modification.

1. DSS DESIGN AND PREOPERATIONAL TEST REQUIREMENTS. INCLUDING f1AJQC

1. Responses to the Prairie Island plant-specific questions contained in the NRC safety evaluation (SE) of WCAP-10858P-A, ("ATWS Mitigating System Actuation Circuitry Generic Design Package"), Revision 1(Proprietary information. Not publicly available.),

need to be provided for the new AMSAC design. Please note that the technical attributes required to be discussed include Diversity, Logic Power Supplies, Safety-Related Interface, Quality Assurance, Maintenance Bypasses, Operating Bypasses, Means for Bypassing, Manual initiation, Electrical Independence, Physical Separation, Environmental Qualification, Testability at Power, Completion of Mitigative Action, and Technical Specifications.

2. Description of pre-operational test methods and criteria for the systems hardware and software testing.

3. Please confirm the following in your final submittal:

' A. That in no circumstances, the stipulations of IEEE 279 are violated since the proposed design modification interfaces with existing safety-related equipment.

B. That the quality of hardware, software, installation, testing, documentation, and record maintenance is according to the guidance provided in Generic Letter 85-06

(" Quality Assurance Guidance for ATWS Equipment That is Not Safety-Related").

C. That guidance of ISA 67-04,1992 and Regulatory Guide 1.105 (" Instrument Seipoints for Safety-Related Systems") Rev.2 is followed for setpoint calculations including the calculations for the steam generator (SG) level setpoints.

D. That the new solid state logic card designed to perform the reactor scram to mitigate an ATWS event is adequately immunized for conducted and radiated ENCLOSURE

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! *t electromagnetic interference /radiofrequency interference (EMI/RFI) and will not become a source of I. armful EMI/RFI that could affect operation of other l safety-related equipment.

E. That the proposed AMSAC/ DSS meets all requirements as stated in 10 CFR 50.62 i including the reliability requirements for both hardware and software designs.

F. That the software and setpoints relating to the AMSAC/ DSS will be subjected to l adequate administrative control such that no unauthorized changes to these features can be performed.

l ll. SAFETY ANALYSIS

1. NSP has used the DYNODE-P and VIPER-1 codes to perform the ATWS analyses in support of the adequacy of the diverse scram system (DSS) design. Both codes were l approved by the NRC for use in the design-basis analysis for the Prairie Island plant.

l However, the codes have not been specifically submitted by the licensee and approved

by the NRC for ATWS analyses. Address the acceptability of the DYNODE-P and

VIPER-1 codes for use in the ATWS analysis for the Prairie Island plant by showing that the system responses and thermal-hydraulic conditions of the ATWS analysis are within the applicable ranges of the approved codes.

2 As a result of the findings from the Maine Yankee Lessons Leamed Task Force, the l staff has taken a position that it requires licensees to verify conformance with the NRC SE for a topical report (TR) whenever the methodologies discussed in the TR are to be used for licensing applications. Accordingly, NSP is required to evaluate its compliance with the conditions spacified in the SEs for TRs referenced in the ATWS analysis to support the DSS design and show that the SE conditions for the TRs have been met.

Provide the results of these assessments.

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3. In the A1WS analysis, the DSS was credited for accident mitigation. Provide a probabilistic safety assessment to show that the total core damage frequency (CDF) for the ATWS cases with the proposed A'lWS Mitigating System Actuation Circuitry (AMSAC) design is very low, and the addition of the DSS will significantly reduce the CDF to justify that its use of the DSS in the ATWS analysis for accident mitigation is acceptable.

4. Identify any assumptions that are different from the original ATWS enalysis performed by Westinghouse for satisfying the ATWS rule and show that changes from the original generic analysis are adequate.

5. In accordance with the staff's technical review procedures on the ATWS analysis for existing pressurized-water reactors, the staff uses the following acceptance criterion to assess the ATWS analysis:

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.M The ATWS analysis must show that the unfavorable exposure time (UET), given the cycle design (including the MTC), will be less than ,

5 percent, or equivalently, that the ATWS pressure limit will be met for at I least 95 percent of the cycle. The UET is the time during the cycle when reactivity feedback is insufficient to maintain pressure under 3200 psi for a given reactor state.

1 Provide a discussion of the bases for selection of the MTC [ moderator temperature coefficient] used in the ATWS analysis and address its compliance with the acceptance criterion (5 percent UET) stated above.

6. Section 2.6 of the submittal dated February 27,1998, summarizes the results of ATWS analyses for six cases. Four cases (loss of normal feedwater, loss of load, loss of ac, and loss of condenser vacuum) can be characterized by rapid reductions in heat removal capability of the SGs. The loss of heat removal results in a rapid rise in the SGs' secondary pressure and temperature and subsequent increase in the reactor coolant system (RCS) pressure and temperature. The severity of these events increases (resulting in a higher maximum RCS pressure) if the primary-to-secondary power mismatch increases. The maximum pressures for the four cases varies by about 100 psi. However, no information was provided to explain what causes the differences in the primary-secondary power mismatch that result in different maximum pressures.

Provide information (such as valve closure time, auxiliary feedwater (AFW) actuation time and reactor coolant pump (RCP) trip time, etc.) that explains the causes for the difference in peak RCS pressures for four cases resulting from a loss of heat removal capability of the SGs.

7. In the ATWS analysis, NSP considered the effects of the SG tube plugging limit of -

15 percent of total tubes on the calculated maximum RCS pressure and minimum DNBR [ departure from nucleate boiling ratio]. Discuss the values of the input parameters ( such as initial RCS flow, RCP coastdown flow, and SG heat transfer area, etc.) reflecting the SG tube plugging and justify that the effects of the SG tube plugging are appropriately considered.

- 8. NSP has stated that the addition of the DSS removes the need to trip the AFW pumps during ATWS conditions. Provide a discussion of the SG response and the AFW pump low-pressure-switch design to show that no trip of the AFW pumps is needed to avoid the AFW pump run-out conditions durin an ATWS event.

9. Section 2.3 of the submittal dated February 27,1998, breaks down the events that are I not analyzed into three categories. Category 2 events are the events that do not require ,

reactor trip to mitigate the consequences of the event with the less restrictive i assumptions applicable to the analysis of"non-design basis" ATWS event. Clarify the definition of Category 2 events and provide a list for the events that belong to this category..

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