ML070811166

Request for Additional Information Re. Amendment for Thermowell-Mounted Temperature Detectors

ML070811166
Person / Time
Site:	Cook
Issue date:	03/27/2007
From:	Tam P NRC/NRR/ADRO/DORL/LPLIII-1
To:	Nazar M Indiana Michigan Power Co
References
TAC MD3462, TAC MD3463
Download: ML070811166 (8)

Text

March 27, 2007 Mr. Mano K. Nazar Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106

SUBJECT:

D. C. COOK NUCLEAR PLANT (DCCNP), UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION REGARDING PROPOSED AMENDMENT INVOLVING THERMOWELL-MOUNTED TEMPERATURE DETECTORS (TAC NOS. MD3462 AND MD3463)

Dear Mr. Nazar:

In a letter dated November 3, 2006, Indiana Michigan Power Company requested an amendment to the Technical Specifications of the units to reflect a plant modification that will replace the Reactor Coolant System resistance temperature detectors and associated bypass piping with fast response thermowell detectors mounted directly in the primary loop piping. The staff determined that additional information is need to complete its review. Accordingly, by e-mail dated January 16 and March 7, 2007 (publicly available at Accession Nos. ML070160275 and ML070660581, respectively), the staff provided draft questions to your personnel.

On March 15, 2007, the staff discussed those questions with your personnel by telephone.

Participants in the phone call agreed to modify or retain as-is the draft questions, resulting in the enclosed Request for Additional Information (RAI). Please respond by June 29, 2007. Feel free to contact me if you need clarification of this RAI.

Sincerely,

/RA/

Peter S. Tam, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316

Enclosure:

Request for Additional Information cc w/encl: See next page

ML070811166 OFFICE LPL3-1\PM LPL3-1/LA EICB/BC SRSB/BC LPL3-1/BC NAME PTam PTam for AHowe* GCranston LRaghavan THarris DATE 3/27/07 3/27/07 3/6/07 3/27/07 3/27/07

• Questions transmitted by memo of 3/6/07.

REQUEST FOR ADDITIONAL INFORMATION D.C. COOK NUCLEAR PLANT, UNITS 1 AND 2 (DCCNP-1, -2)

AMENDMENT REGARDING THERMOWELL-MOUNTED TEMPERATURE DETECTORS

Reference:

Application for amendment from Indiana Michigan Power company, November 3, 2006 Reactor Systems Branch Questions A.1 The TS changes associated with the resistance temperature detector bypass line elimination have already been approved for DCCNP-1 (Amendment No. 296, dated October 6, 2006). As with DCCNP-1 (see licensees letter dated May 31, 2006; Accession No. ML061600449), the response time for the OTT trip in DCCNP-2 will be maintained at 8 seconds or less.

The following events could lead to a reactor trip when the calculated OTT trip setpoint is reached:

1. Loss of electrical load/turbine trip

2. Uncontrolled rod cluster control assembly (RCCA) bank withdrawal at power

3. Chemical Volume Control System (CVCS) malfunction that results in a decrease in the boron concentration in the reactor coolant

4. Inadvertent opening of a pressurizer safety or relief valve The licensee has provided evaluations of the first three events for both Cook units (licensees May 31 and November 3, 2006, letters). The fourth event, the inadvertent opening of a pressurizer safety or relief valve, is not in the licensing basis of either Cook unit. This event, like the uncontrolled RCCA bank withdrawal at power, could erode thermal margin (i.e., an OTT trip could occur as thermal margin is decreased by a reduction in reactor coolant system (RCS) pressure as well as by an increase in power generation). Both the uncontrolled RCCA bank withdrawal at power event, and the inadvertent opening of a pressurizer safety or relief valve event are important in the determining the constants and coefficients in the OTT trip setpoint equation and of the OTT trips dynamic time response characteristics.

The OTT trip is designed to protect the plant against departure-from-nucleate-boiling (DNB) during uncontrolled RCCA bank withdrawal at power events that insert reactivity slowly. The high nuclear flux trip provides protection when reactivity is inserted more rapidly. The OTT trip, and the low pressurizer pressure trip, protect the plant against DNB during the inadvertent opening of a pressurizer safety or relief valve events. The effectiveness of the OTT trip is verified by showing that the reactor trip signal is generated in time to prevent DNB, without taking credit for a reactor trip from the low pressurizer pressure trip logic. This has not been done for either of the Cook units, since these units do not include the inadvertent opening of a pressurizer safety or relief valve event in their licensing bases.

Table 1 shows that the inadvertent opening of a pressurizer safety or relief valve event was specified in the Standard Format in October, 1972 (RG 1.70, Revision 1).

Table 1: Standard Format and Content of Safety Analysis Reports Event analyses that are not in the Licensing Bases of RG 1.70 RG 1.70 RG 1.70 DCCNP-1 and -2 R0 - 2/72 R1 - 10/72 R1 - 10/72 Reactor coolant pump shaft break locked rotor locked rotor T15-1 (3.4)

Single RCCA withdrawal T15-1 (3) T15-1 (3) T15-1 (4.3)

Inadvertent loading and operation of a fuel assembly in T15-1 (18) T15-1 (15) T15-1 (4.7) an improper position Inadvertent actuation of the emergency core cooling T15-1 (32) T15-1 (5.1) system that increases RCS inventory Inadvertent actuation of the CVCS that increases RCS T15-1 (4) T15-1 (4) T15-1 (5.2) inventory Inadvertent opening of a pressurizer PORV T15-1 (13) T15-1 (6.1)

Radiological consequences of failure of small lines T15-1 (26) T15-1 (22) T15-1 (6.2) carrying primary coolant outside containment Table 2 shows that DCCNP-1 and DCCNP-2 were licensed 2 and 5 years, respectively, after RG 1.70 incorporated the inadvertent opening of a pressurizer safety or relief valve event.

Table 2: Chronology June, 1966 A Guide for the Organization and Contents of Safety Analysis Reports July, 1967 Proposed General Design Criteria February, 1971 10 CFR 50, App A, General Design Criteria July, 1971 10 CFR 50, App A, General Design Criteria February,1972 RG 1.70, Rev 0, Standard Format and Content October, 1972 RG 1.70, Rev 1, Standard Format and Content August, 1973 ANSI N18.2-1973, Nuclear Safety Criteria for Design of PWRs October, 1974 DCCNP-1 was licensed January, 1975 DCCNP-1 achieved initial criticality September, 1975 RG 1.70, Rev 2, Standard Format and Content November, 1975 NUREG-75/087 SRP December, 1977 DCCNP-2 was licensed March, 1978 DCCNP-2 achieved initial criticality November, 1978 RG 1.70, Rev 3, Standard Format and Content October, 1986 DCCNP-2, Cycle 6 SAR August, 1989 Analysis of D.C. Cook Unit 2, Cycle 8 Reload

Table 2 also shows that DCCNP-1 and -2 were licensed more than a year after the issuance of ANSI N18.2-1973, Nuclear Safety Criteria for Design of PWRs. This standard categorizes the analyzed events according to expected frequency of occurrence, and lists the inadvertent opening of a pressurizer safety or relief valve event as an example of a Condition II (an event of moderate frequency) event. One year after DCCNP-2 was licensed, another revision of RG 1.70 and the Standard Review Plan were issued. Both contained the inadvertent opening of a pressurizer safety or relief valve event. Nevertheless, the licensee continued to maintain that this event, along with six others listed in Table 1, were not in the licensing bases of DCCNP-1 and -2. The NRC staff accepted this position as recently as 1989 (letter from J. G. Giitter, August 3, 1989). At that time, the licensee and Westinghouse asserted that the seven events were analyzed or evaluated by ANF in response to NRC staff questions regarding the use of ANF methodology in the licensing of ANF-supplied fuel. They were not part of the licensing basis for Westinghouse-supplied fuel. As such, they were not to be considered as part of the licensing basis when the Cook fuel supply contracts reverted to Westinghouse.

The seven events of Table 1 are not in the current licensing bases of DCCNP-1 and -2.

Yet, when issuing an amendment, the NRC staff needs to be able to make the statement that there is reasonable assurance that the activities authorized by [the]

amendment can be conducted without endangering the health and safety of the public (i.e., the absence of an issue in the current licensing basis is not a cause prohibiting the staff from reviewing that issue where safety may be affected by the proposed amendment). The fact that the current DCCNP-1 and -2 licensing bases do not include the aforementioned seven event evaluations or analyses should not prevent the NRC staff to question whether there is a significant reduction in a margin of safety related to one of these events. The subject amendment application would result in a change to the OTT trip. Accordingly, the staff requests an analysis, or equivalent, to provide reasonable assurance that the modified OTT trip will not significantly reduce a margin of safety (e.g., thermal margin) during an inadvertent opening of a pressurizer relief or safety valve, an event that could demand a reactor trip through the OTT trip logic.

A.2 [Original draft question deleted per telephone discussion of March 15, 2007.]

A.3 Please verify that despite the proposed changes to the allowable values for the OPT and the OTT trip set points, the UFSAR analysis limits will be maintained.

A.4 The calculations performed by the NRC staff show that the changes in the allowable values are within a fraction of a degree Fahrenheit. Show that the new thermowell resistance temperature detectors (RTDs) have the capability to measure this difference in the allowable value.

Instrumentation and Controls Branch Questions B.1 Enclosure 2, Section 4.0, discusses in general, the instrument uncertainty considerations for the calculations of the allowable value for OTT and OPT and Enclosure 3 provides the generic D. C. Cook Nuclear Plant setpoint methodology found acceptable by the NRC. Please provide the detailed calculations, including all actual values used for uncertainties, that show justification for the increase in allowable values for OTT and OPT. Also, please provide the source and/or justification for each uncertainty value used in the calculation.

B.2 Enclosure 2, Section 3.0, states that the three hot-leg scoops in each reactor coolant system (RCS) loop will be modified to accept the new thermowells, which will contain the new, fast-response RTDs and that a hole will be drilled through the end of each scoop to facilitate flow past the RTD. How large is the drilled exit hole in comparison to the scoops water-entry cross-section size? How was it determined that this exit hole size was sufficient to not cause reduced flow through the scoop that could potentially add to a delay in the response time of the measurement of RCS temperature changes or even introduce another uncertainty in the measurement?

B.3 [Original draft question deleted per telephone discussion of March 15, 2007.]

B.4 Enclosure 2, Section 3.0, describes in general, the arrangement whereby the three RTDs in an RCS loop will be electronically averaged to obtain a single hot-leg RCS temperature for that loop. Please describe the averaging function. Can a failure of one of the three RTDs in an RCS loop be automatically identified and taken out of the averaging equation by the new electronic averaging circuit?

B.5 [This question was not discussed in the March 15, 2007, phone call.] Enclosure 3 states that the NRC concluded that the DCCNP allowable value calculation methodology is acceptable in a letter dated June 1, 2005 (Reference 6). However, based on the staff concerns identified in the NRC Regulatory Issue Summary (RIS) 2006-17 (Reference 4),

please provide a statement confirming that the setpoints for OTT and OPT are Limiting Safety System Settings for the variables on which a Safety Limit (SL) has been placed.

B.6 [This question was not discussed in the March 15, 2007, phone call.] The NRC letter to the Nuclear Energy Institute, Setpoint Methods Task Force, dated September 7, 2005 (Reference 1), describes setpoint-related technical specifications (SRTS) that are acceptable to the NRC for instrument settings associated with SL-related setpoints.

Specifically, Part A of the Enclosure to the letter provides limiting condition of operation notes to be added to the TS, and Part B includes a check list of the information to be provided in the TS Bases related to the proposed TS changes.

a. Describe whether and how you plan to implement the SRTS suggested in the September 7, 2005, letter. If you do not plan to adopt the suggested SRTS, then explain how you will ensure compliance with 10 CFR 50.36 by addressing items b and c, below.

b. As-Found Setpoint Evaluation: Describe how surveillance test results and associated TS limits are used to establish operability of the safety system. Show that this evaluation is consistent with the assumptions and results of the setpoint

calculation methodology. Discuss the plant corrective action processes (including plant procedures) for restoring channels to operable status when channels are determined to be inoperable or operable but degraded. If the criteria for determining operability of the instrument being tested are located in a document other than the TS (e.g. plant test procedure), explain how the requirements of 10 CFR 50.36 are met.

c. As-Left Setpoint Control: Describe the controls employed to ensure that the instrument setpoint is, upon completion of surveillance testing, consistent with the assumptions of the associated analyses. If the controls are located in a document other than the TS (e.g. plant test procedure), explain how the requirements of 10 CFR 50.36 are met.

References

1. Letter from P. L. Hiland, NRC, to NEI Setpoint Methods Task Force, "Technical Specification for Addressing Issues Related to Setpoint Allowable Values," dated September 7, 2005 (Accession No. ML052500004).

2. Letter from B. A. Boger, NRC, to A. Marion, "Instrumentation, Systems, and Automatic Society (ISA) S67.04 Methods for Determining Trip Setpoints and Allowable Values for Safety-Related Instrumentation," dated August 23, 2005 (Accession No. ML051660447).

3. Letter from J. A. Lyons, NRC, to A. Marion, NEI, "Instrumentation, Systems, and Automation Society S67.04 Methods for Determining Trip Setpoints and Allowable Values for Safety-Related Instrumentation," dated March 31, 2005 (Accession No.

ML050870008).

4. NRC Regulatory Issue Summary 2006-17, NRC Staff Position on the Requirements of 10 CFR 50.36, Technical Specification, Regarding Limiting Safety System Setting During Periodic Testing and Calibration of Instrument Channels, dated August 24, 2006 (Accession No. ML051810077).

5. Technical Specification Task Force (TSTF) recommendation for Standard Technical Specification (STS) changes, TSTF-493, Clarify Application of Setpoint Methodology for LSSS Functions, Revision 0, January 27, 2006 (Accession No. ML060270503).

6. Letter from J. Donohew, NRC, to M. Nazar, I&M, dated June 1, 2005, Paragraphs G.1.2.a, G.1.2.b, and G.3.2 of the Safety Evaluation for the conversion of the CNP TS to Improved Technical Specifications (Accession No. ML050620034).

Donald C. Cook Nuclear Plant, Units 1 and 2 cc:

Regional Administrator, Region III Michigan Department of Environmental U.S. Nuclear Regulatory Commission Quality Suite 210 Waste and Hazardous Materials Div.

2443 Warrenville Road Hazardous Waste & Radiological Lisle, IL 60532-4351 Protection Section Nuclear Facilities Unit Attorney General Constitution Hall, Lower-Level North Department of Attorney General 525 West Allegan Street 525 West Ottawa Street P. O. Box 30241 Lansing, MI 48913 Lansing, MI 48909-7741 Township Supervisor Lawrence J. Weber, Plant Manager Lake Township Hall Indiana Michigan Power Company P.O. Box 818 Nuclear Generation Group Bridgman, MI 49106 One Cook Place Bridgman, MI 49106 U.S. Nuclear Regulatory Commission Resident Inspector's Office Mark A. Peifer, Site Vice President 7700 Red Arrow Highway Indiana Michigan Power Company Stevensville, MI 49127 Nuclear Generation Group One Cook Place Kimberly Harshaw, Esquire Bridgman, MI 49106 Indiana Michigan Power Company One Cook Place Bridgman, MI 49106 Mayor, City of Bridgman P.O. Box 366 Bridgman, MI 49106 Special Assistant to the Governor Room 1 - State Capitol Lansing, MI 48909 Susan D. Simpson Regulatory Affairs Manager Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106