L-MT-10-054, Response to Requests for Additional Information Concerning Extension of Permanent Relief from Inspection Requirements of 10 CFR 50.55a(g) for Volumetric Examination of Reactor Pressure Vessel Shell Circumferential Welds for the Renewed O

Response to Requests for Additional Information Concerning Extension of Permanent Relief from Inspection Requirements of 10 CFR 50.55a(g) for Volumetric Examination of Reactor Pressure Vessel Shell Circumferential Welds for the Renewed Oper

ML102710108
Person / Time
Site:	Monticello
Issue date:	09/27/2010
From:	O'Connor T Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-10-054
Download: ML102710108 (11)

Text

September 27, 2010 Monticello Nuclear Generating Plant 2807 W County Road 75 Monticello, MN 55362 L-MT-10-054 10 CFR 50.55a U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Monticello Nuclear Generating Plant Docket 50-263 Renewed Facility Operating License No. DPR-22 Response to Requests for Additional Information Concerning Extension of Permanent Relief from Inspection Requirements of 10 CFR 50.55a(g) for Volumetric Examination of Reactor Pressure Vessel Shell Circumferential Welds for the Renewed Operating License Term

References:

1)

NSP Letter to NRC, "Relief Request No. 17, Extension of Permanent Relief from Volumetric Examination of Reactor Pressure Vessel Circumferential Shell Welds for the Renewed Operating License Term," (L-MT-10-014) dated March 12, 201 0.

2)

NRC e-mail from P. Tam, to R. Loeffler, "Monticello - Draft RAI on Relief Request dated 311 211 0 (TAC ME3526)," dated July 9, 201 0.

3)

NRC e-mail from P. Tam, to R. Loeffler, "Monticello - Second set of draft RAI re. Relief Request dated 3/12/10 (TAC ME3526)," dated July 14, 2010.

On March 12,2010, the Northern States Power Company - Minnesota (NSPM) submitted Relief Request No. 17, Reference 1, requesting authorization for an alternative to performing volumetric examination of the Monticello Nuclear Generating Plant (MNGP) reactor pressure vessel circumferential shell welds for the term of the Renewed Facility Operating License.

The U.S. Nuclear Regulatory Commission (NRC) staff requested additional information in two e-mail requests (References 2 and 3, respectively) to support their review. provides the NSPM response to these requests.

Document Control Desk L-MT-10-054 Page 2 of 2 Summary of Commitments This letter proposes no new commitments and does not revise any existing commitments.

ions regarding this letter, please contact Mr. Richard Loeffler at President, Monticello Nuclear Generating Plant Northern States Power Company - Minnesota Enclosure cc:

Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce

L-MT-10-054 Page 1 of 9 Response to RAls for a 10 CFR 50.55a Request Concerning Volumetric Examination of RPV Shell Circumferential Welds For the Term of the Renewed Facility Operating License On March 12, 2010, the Northern States Power Company - Minnesota (NSPM) submitted Relief Request No. 1.7, "Extension of Permanent Relief from Volumetric Examination of Reactor Pressure Vessel [RPV] Circumferential Shell Welds for the Renewed Operating License Term," (Reference 1) for the Monticello Nuclear Generating Plant (MNGP). The proposed alternative allows the permanent deferral of performance of volumetric examinations of the RPV shell circumferential welds as required by 10 CFR 50.55a(g) for the 20-year duration of the renewed operating license.

The U.S. Nuclear Regulatory Commission (NRC) staff requested additional information (RAI) in two e-mail requests (References 2 and 3, respectively) to support review of the requested alternative. The NRC RAls are shown in boldface text with the responses immediately following.

Part I I) Please describe the fluence calculations performed that support this relief request. Address their adherence to Regulatory Guide 1.190 recommendations, and if applicable, reference the NRC-approved methodology document that was used to perform the calculations.

Response

The NRC-approved and Regulatory Guide 1.I90 compliant General Electric - Hitachi (GEH) fluence methodology, NEDC-32983P-A, "Licensing Topical Report, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations" (Reference 4) was used to perform the fluence calculations for the MNGP. The fluence calculations for the MNGP Extended Power Uprate (EPU) were performed and are documented in a GEH report entitled "Monticello Neutron Flux and Fluence Evaluation for Extended Power Uprate" (Reference 5).

L-MT-10-054 Page 2 of 9 Response to RAIs for a 10 CFR 50.55a Request Concerning Volumetric Examination of RPV Shell Circumferential Welds For the Term of the Renewed Facility Operating License

2)

Please describe the fast neutron flux source used to determine the 120-percent original licensed thermal power [OLTP] level neutron fluence.

Discuss whether the uprated fluence projection is based on extrapolation of previous fluence values, or on a new source analysis. If it is based on extrapolation, please provide additional information justifying the extrapolation in terms of its conservatism relative to expected peripheral flux values during uprated operation.

(It was discussed during the July 20, 2010, conference call that the fast neutron flux source was not based upon an extrapolation of the OLTP neutron fluence, and it was indicated that the second part of the question was not applicable.)

Response

Flux calculations were performed applying the considerations of the NRC approved GEH fluence methodology described in NEDC-32983P-A. The neutron source distribution used for the EPU fluence calculation is based on the EPU equilibrium core, not on an extrapolation of the original licensed thermal power source distribution.

L-MT-10-054 Page 3 of 9 Response to RAls for a 10 CFR 50.55a Request Concerning Volumetric Examination of RPV Shell Circumferential Welds For the Term of the Renewed Facility Operating License Part I Criterion 2 of Generic Letter 98-05 requires that the licensee implement sufficient procedures andlor operator training to ensure that the probability of a cold overpressure event is minimized. To satisfy this criterion, the licensee has provided analyses of the potential high-pressure injection sources, administrative controls, and operator training, that help to minimize the risk of cold overpressure events. The licensee noted that it examined a number of conditions that may be precursors to cold overpressurization; however, the licensee did not discuss any of the conditions.

(1) Provide a discussion on the following systems regarding minimizing a cold overpressure event: e.g. flow rates; inadvertent injections; overfill; etc.

High Pressure Core Spray Injection Reactor Core Isolation Cooling System Feed Water System Control Rod Drive and Reactor Water Cleanup Systems Standby Liquid Control System RPVPressureTesting

Response

The NRC final Safety Evaluation (SE) for the Boiling Water Reactor Vessel and lnternals Project (BWRVIP) for BWRVIP-05 licensing topical report (LTR) (Reference 6) and the Safety Evaluation Report (SER) for Section 4.2.6.2, "Staff Evaluation," of the MNGP License Renewal NUREG-1 865 (Reference 7), indicate that an acceptable alternative has to include implementation of operator training and established procedures that limit the frequency of cold overpressure events to the frequency specified in the staff's SER. The NSPM committed to and revised plant procedures and upgraded operator training in accordance with our previous application for this alternative (see References 8 and 9). These changes minimize the potential for a cold overpressure event in conjunction with previously receiving approval to apply this alternative through the end of the full-term operating license.

Section 2.6.1.I, "Frequency Estimation of Cold Over Pressurization Events from Inadvertent Injections," of the NRC SE for the BWRVIP-05 LTR and the follow-on documents for this subject, (Reference 6), concluded that although water can be

L-MT-I 0-054 Page 4 of 9 Response to RAls for a 10 CFR 50.55a Request Concerning Volumetric Examination of RPV Shell Circumferential Welds For the Term of the Renewed Facility Operating License supplied to the reactor pressure vessel (RPV) by various systems,(') not all of these injection systems can be considered as a possible initiator of a cold over pressure condition. The NRC SER for the BWRVIP-05 report stated the following:

Specifically, the BWRVIP stated that most of the... systems do not contribute to the potential for cold over pressurization events because:

( 1 the system shutoff. head is low for several systems (e.g., the LPCS 1 LPCI pumps) so that the vessel remains within the acceptable limits of the pressure-temperature (P/T) curves even at shutdown temperatures (i.e.,

these systems can be activated only under low pressure),

(2) overfilling and pressurization to the shutoff head is very unlikely because of automatic trip of the system on high water level (e.g., HPCS),

(3) the system is steam driven and is not in use during cold shutdown conditions (e.g., RCIC and HPCI), and (4) operation of the system, such as standby liquid control, requires a series of deliberate operator actions like manual pump activation and is unlikely to happen without adequate monitoring.

Thus, RCIC, HPCI, feedwater, HPCS, LPCI, LPCS, and the SLC systems were considered to have a negligible impact on the risk of a cold over pressure event.

The NRC SE (Reference 6) for the BWRVIP-05 LTR indicates that the major contribution to low temperature over-pressure (LTOP) event frequency, as determined by the BWRVIP, results from unmitigated injections from the Condensate or Control Rod Drive (CRD) Hydraulic Systems and a failure to properly realign the Reactor Water Cleanup (RWCU) System following a reactor trip at low temperatures.

1.

The NRC SER lists the following systems; feedwater, condensate, control rod drive [CRD],

standby liquid control [SLC], reactor core isolation cooling [RCIC], High Pressure Coolant Injection [HPCI], High Pressure Core Spray [HPCS], Low Pressure Core Spray [LPCS],

and I or Low Pressure Coolant Injection [LPCI] mode of Residual Heat Removal [RHR].

L-MT-10-054 Page 5 of 9 Response to RAls for a 10 CFR 50.55a Request Concerning Volumetric Examination of RPV Shell Circumferential Welds For the Term of the Renewed Facility Operating License A discussion on the HPCI, RCIC, Feedwater, CRD, RWCU and SLC Systems is provided below. The answer to the second RAI provides a restatement of changes to the operating procedures and the improvements in operator awareness of the potential for an LTOP.

This was previously accomplished by enhancing the operating procedures and operator training.

High Pressure Coolant Injection and Reactor Core Isolation Cooling Svstems The High Pressure Coolant lnjection(*)(HPCI) and Reactor Core Isolation Cooling (RCIC) systems are sources of high-pressure injection to the RPV. Both the HPCI and the RCIC systems are steam-turbine driven. During reactor cold shutdown conditions, no steam is available for operation of these systems. Therefore, it is not plausible for these systems to contribute to an over pressurization event while the unit is in cold shutdown.

Feedwater and Condensate Svstems For the reactor feed pumps, an inadvertent injection with the vessel water level greater than +48 inches is controlled by a high water level interlock. During outages, reactor vessel level is maintained greater than +48 inches, preventing a feedwater pump from starting unless the bypass switch is placed in bypass. Switch position is procedurally controlled to prevent injecting into the vessel. During the reactor vessel pressure test, the bypass switch and the feedwater pumps are isolated and tagged.

For the condensate pumps, precautions are provided in the operating procedures, which indicate to the operators that they need to monitor reactor water level closely when the pumps are supplying feed to the reactor vessel, in order to prevent an overfill event. However, since the shutoff head of the condensate pumps is only approximately 350 psig, this does not represent a significant challenge to the RPV.

Procedural and caution steps are provided in operating procedures requiring monitoring of RPV pressure and temperature against the PIT curves for evolutions where LTOP could be of concern, i.e., startup, shutdown and pressure 1 hydrostatic testing. There are also high reactor water level and high reactor pressure alarms in the control room that warn operators when level I pressure limitations are being exceeded.

Control Rod Drive and Reactor Water Cleanup Svstems The CRD is another high-pressure water source to the RPV. The nominal flowrate of each CRD pump is 74 gpm at a discharge pressure of 1625 psig. Each RWCU pump is capable of delivering 160 gpm of normal cleanup flow. During a reactor

2.

The High Pressure Core Spray System is not used in the MNGP BWR/3 design.

L-MT-I 0-054 Page 6 of 9 Response to RAls for a 10 CFR 50.55a Request Concerning Volumetric Examination of RPV Shell Circumferential Welds For the Term of the Renewed Facility Operating License vessel pressure test, the flow rate to the vessel varies and is dependent on the rate of flow being discharged through RWCU System. The operators maintain RPV pressure by balancing CRD and RWCU flow.

The response to the second RAI (see Items 3 and 4) discusses procedural changes and operator training implemented for reactor vessel pressure testing to address reactor vessel stratification and provide recovery measures following a scram.

Standby Liquid Control System The Standby Liquid Control (SLC) System is another high-pressure water source to the RPV. There are no automatic starts associated with this system. SLC injection requires an operator to manually start the system from the control room (via a keylock switch manipulation). Additionally, the injection rate of a SLC pump is approximately 28.5 gpm at a discharge pressure of 1500 psig, which provides ample time for an operator to control reactor pressure in the event of an inadvertent injection.

(2)

Provide a brief summarization of the Operator Training and Operating Procedures as they pertain to minimizing a cold overpressure event.

Response

As discussed in the NRC SE for the BWRVIP-05 LTR (Reference 6), cold pressurization events were reviewed by both the industry, courtesy of the BWRVIP, and the NRC to identify the critical operator actions that were assumed to occur to mitigate these events. Then these results and conclusions were reviewed by each BWR licensee for applicability to their plant. Enhancements were made to operating /test procedures and training to ensure that operators were aware of the potential LTOP for applicable situations and, that operator actions would occur with a high degree of certainty so that the LTOP event frequency is maintained less than that determined acceptable by the NRC as stated in their SE. Following are results of the MNGP specific review for situations where cold over pressurization events could occur.

1.

From Inadvertent lniections The evaluation provided in Section 2.6.1.1 of the BWRVIP-05 SE (Reference 6) was determined to be applicable to the MNGP with one exception. The evaluation considered the availability of automatic trips of high pressure injection systems on high water level. Review of the MNGP procedures identified that during performance of reactor feedwater pump (RFP) testing during cold shutdown, the high reactor water level trip is bypassed. Measures are taken

L-MT-10-054 Page 7 of 9 Response to RAls for a 10 CFR 50.55a Request Concerning Volumetric Examination of RPV Shell Circumferential Welds For the Term of the Renewed Facility Operating License procedurally to close valves to prevent water from getting to the vessel. NSPM enhanced the Condensate and Reactor Feedwater System operations procedure to further assure the isolation of flow to the vessel.

2.

From Condensate lniection The evaluation provided in Section 2.6.1.2 of the BWRVIP-05 SE (Reference 6) is applicable to the MNGP. Operating procedures provide precautions which indicate that reactor water level is to be closely monitored when starting a condensate pump. This aids in assuring that an overfill event, which could lead to an LTOP event, does not occur. In order to assure that operations personnel understand that an overfill event has the potential to lead to an LTOP event, NSPM enhanced the Condensate and Reactor Feedwater System operations procedures to identify an LTOP event as a potential consequence of an overfill event. Procedural and caution steps are provided in operating procedures requiring monitoring of RPV pressure and temperature against the PIT curves for evolutions where LTOP could be of concern, i.e., startup, shutdown and pressure I hydrostatic testing. There are also high reactor water level and high reactor pressure alarms in the control room that warn operators when level I pressure limitations are being exceeded.

3.

From CRD lniection The evaluation provided in Section 2.6.1.3 of the BWRVIP-05 SE (Reference 6) is applicable to the MNGP. The evaluation notes that the risk of cold over pressurization due to CRD injection may be higher if a loss of station power were to occur during reactor vessel pressure testing. NSPM revised the reactor vessel pressure testing procedures to provide precautions that ensure proper response to a loss of station power (i.e., RWCU and recirculation pumps are restored along with restoration of CRD).

4.

From Loss of RWCU The evaluation provided in Section 2.6.1.4 of the BWRVIP-05 SE (Reference 6) is applicable to the MNGP. NSPM has procedures in place to provide guidance for recovery measures following a scram. In the event that a scram occurs that results in a RWCU isolation, procedural guidance is provided which consists of restoring the RWCU System as soon as the cause of the isolation is identified and resetting the reactor scram as soon as possible in order to limit cold water injection into the vessel. Also, procedural guidance is provided for dealing with recirculation loop or reactor vessel stratification so that an excessive amount of cold water is not distributed throughout the reactor vessel during the restart of a tripped recirculation pump(s). NSPM added a precaution in the reactor scram

L-MT-10-054 Page 8 of 9 Response to RAls for a 10 CFR 50.55a Request Concerning Volumetric Examination of RPV Shell Circumferential Welds For the Term of the Renewed Facility Operating License procedure for RWCU restoration in order to further inform the operations personnel of the potential of an LTOP event occurring during SCRAM recovery.

5.

An Actual Cold Over Pressurization Event Section 2.6.1.5 of the BWRVIP-05 SE (Reference 6) discusses an actual cold over pressurization event that occurred at a foreign BWR. The NSPM assessment of the event indicated that the likelihood of a similar event occurring at the MNGP is very low. Procedures require that the reactor vessel remain vented at all times during cold shutdown except as permitted by approved procedures. The reactor vessel pressure test procedure allows the vent valves to be closed during cold shutdown. During the pressure test, strict procedural guidance is provided for administratively monitoring reactor vessel pressure and temperature while controlling CRD injection and RWCU reject in order to assure a smooth, controlled method of increasing or decreasing pressure while the reactor vessel temperature is being maintained above the required P-T limits, If reactor pressure exceeds the specified limits, during the test, the CRD pump is immediately tripped. In addition to the above mentioned procedural guidance, a requirement is included to perform an "Infrequent Test or Evolution BriefingJ1 with all essential personnel. This briefing details the anticipated testing evolution with special emphasis on conservative decision making, plant safety awareness, lessons learned from similar in-house or industry operating experiences, the importance of open communications, and the process in which the test would be aborted if plant systems responded in an adverse manner.

6.

Operator Training The MNGP has procedures in place that guide operators in controlling and monitoring reactor pressure during all phases of operation, including cold shutdown. Use of these procedures minimizes the potential for LTOP events and is reinforced through operator training. Licensed operator training addresses LTOP events. Training is provided to operations personnel on the specific scenarios and events evaluated in (Sections 2.6.1.I through 2.6.1.5 of Reference 6). This training includes features of system design and procedural controls that prevent such events at the MNGP.

The above indicates that the system design and operating /testing procedures, after implementation of the original enhancements, when this alternative was first approved, minimize the probability of LTOP events at the MNGP.

L-MT-10-054 Page 9 of 9 Response to RAls for a 10 CFR 50.55a Request Concerning Volumetric Examination of RPV Shell Circumferential Welds For the Term of the Renewed Facility Operating License REFERENCES

1.

NSP Letter to NRC, "Relief Request No. 17, Extension of Permanent Relief from Volumetric Examination of Reactor Pressure Vessel Circumferential Shell Welds for the Renewed Operating License Term," (L-MT-10-014) dated March 12, 2010.

2.

NRC e-mail from P. Tam, to R. Loeffler, "Monticello - Draft RAI on Relief Request dated 3/12/10 (TAC ME3526)," dated July 9, 201 0.

3.

NRC e-mail from P. Tam, to R. Loeffler, "Monticello - Second set of draft RAI re.

Relief Request dated 3/12/10 (TAC ME3526)," dated July 14, 201 0.

4.

GEH Report NEDC-32983P-A, Revision 2, "Licensing Topical Report, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations," dated January 2006.

5.

GEH Report 0000-0076-7052-R0, "Monticello Neutron Flux and Fluence Evaluation for Extended Power Uprate, dated December 2007.

6.

NRC Letter to BWRVIP, "Final Safety Evaluation of the BWR Vessel and lnternals Project BWRVIP-05 Report (TAC No. M93925)," dated July 28, 1998.

7.

NUREG-1865, "Safety Evaluation Report Related to the License Renewal of the Monticello Nuclear Generating Plant," dated October 2006.

8.

NMC letter to NRC, "Request for Relief No. 12 for the Third 10-Year Interval lnservice lnspection Program," dated October 10, 2000.

9.

NMC letter to NRC, "Response to NRC Request for Additional Information for Request for Relief No. 12 for the Third 10-Year Interval lnservice lnspection Program," dated May 3, 2001.