ML20076K323

Forwards Response to 830610 Request for Addl Info Re Methodology Used to Perform Safety Analyses Supporting Cycle 6 Operation.Analyses Performed Based on C-E Owners Group Rept CEN-199 in Response to NUREG-0737,Item II.K.2.17

ML20076K323
Person / Time
Site:	Millstone
Issue date:	07/01/1983
From:	Counsil W NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	Clark R Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.17, TASK-TM A03323, A3323, TAC-49798, NUDOCS 8307080400
Download: ML20076K323 (9)

Text

gg General Offices

• Seldon Street Berlin, Connecticut v.g cowcecut vo , mo eona ==

P.O. BOX 270 2, 7.7.0, ',*

• H ARTFORD. CONNECTICUT 45141-0270 (203) 66 & 6911 L t J C' O';',C,",",

July 1,1983 Docket No. 50-336 A03323 Director of Nuclear Reactor Regulation Attn: Mr. Robert A. Clark, Chief Operating Reactors Branch //3 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

References:

(1) R. A. Clark letter to W. G. Counsit, dated June 10,1983.

(2) R. A. Clark letter to W. G. Counsil, dated January 12,1982.

(3) W. G. Counsil letter to R. A. Clark, dated April 13,1983.

(4) W. G. Counsil letter to R. A. Clark, dated June 22,1983.

Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 Additional Information in Support of Cycle 6 Operation In Reference (1), the NRC Staff requested additional information regarding the methodology utilized in performing the safety analyses supporting Cycle 6 operation of Millstone Unit No. 2. The attached information represents the Northeast Nuclear Energy Company (NNECO) response to that request.

As a general comment, NNECO is perplexed that the Staff continues to question the methodology utilized by our fuel vendor in performing the safety analyses to support plant operation. The NRC Staff has reviewed the Basic Safety Report (BSR) submitted in support of Westinghouse reloads of Millstone Unit No. 2 and has documented the results of their review of the transient analyses in Ref erence (2).

Reference (2) documents the acceptability of the transient analyses presented in the BSR. The analyses performed for Cycle 6 operation were performed utilizing equivalent methodologies. The request transmitted by Reference (1) suggests that the Staff is questioning the correctness of this prior approval. We believe that such a concern is not technically justified.

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The attached responses address Questions 400.1-400.3 of Reference (1). NNECO intends to provide the response to Question 400.5 concerning the Steam Generator Tube Rupture analysis by July 15,1983.

As identified in Reference (4), NNECO has identified failed fuel at Millstone Unit No. 2. The need to replace defective fuel assemblies with new assemblies or previously discharged fuel from Cycle 1 operation willimpact the core loading pattern and core physics parameters used as input to the safety analyses docketed in Reference (3). The ef fect of a new loading pattern on the ongoing review of the Reference (3) material is not fully apparent at this time; however, additional transient analyses may be required if significant loading pattern changes are needed. My staff will promptly apprise you of any forthcoming revisions to the Reference (3) safety analyses or technical specifications as soon at the results are available.

We trust you find the attached information responsive to the Reference (1) request.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY UAlh W.'G.'Counsit Senior Vice President

Docket No. 50-336 Attachment 1 Millstone Nuclear Power Station, Unit No. 2 Additional Information in Support of Cycle 6 Operation July,1983

Question 440.1 The steam line break analysis submitted in support of Cycle 6 reload was at no load conditions with of fsite power available. CEN-199 was submitted by the C-E Owners Group in response to TMI Action item II.K.2.17 of NUREG-0737. This reporc states that the limiting DNBR steam line break event for Millstone 2 occurs at full power conditions with loss of offsite power at the time of break initiation. Address why the full power event, with offsite power not available, is no longer limiting. Justify this with an appropriate analysis. In addition, since the Westinghouse methodology is being applied instead of the CE methodology, provide another calculation which assumes offsite power available. This additional analysis is needed because the limiting steam line break event analyzed by Westinghouse for their plants occurs when of fsite power is available.

When analyzed by Combustion Engineering, the limiting event occurs when of fsite power is lost. The staff, at this time, is unable to determine if the difference between results calculated by the two vendors is due to differences in design or calculational methodology. For all calculated steam line break events, provide the minimum DNBR as a function of time.

Response

The Combustion Engineering Owner's Group (CEOG) submitted CEN-199 in support of NUREG-0737 Item II.K.2.17, Reactor Coolant System voiding. Item II.K.2.17 required an evaluation of the potential for voiding in the reactor coolant system (RCS) during transients. These evaluations were considered necessary due to the course representation of the reactor vessel upper head region transient response by analysis models.

The evaluations presented in CEN-199 were performed utilizing assumptions which maximize the potential for RCS voiding. These assumptions include full power operation and RCS operating temperatures followed by a steam line rupture and loss of offsite power. The assumption of full power operation will maximize the energy stored in the RCS including the upper head and f uel. The assumption of loss of offsite power and attendant loss of forced reactor coolant flow is conservative in that reactor vessel upper head flow is essentially reduced to zero.

The results of the evaluation presented in CEN-199 demonstrate that void formation during a steam line rupture is not great enough to impair reactor coolant circulation or core coolability. Additionally, CEN-199 concluded that the consequences of non-uniform mixing in the reactor vessel upper head is of secondary importance when compared to the consequences from not explicitly modeling the reactor vessel upper head region.

The Cycle 6 safety analysis of the steam line break submitted in Reference (1) represents the licensing basis for Millstone Unit No. 2. The analysis was performed to maximize the RCS cooldown and potential for core criticality and return to power. A return to power following a steam pipe rupture is a potential occurrence mainly because of the high power peaking factors which exist assumind the most reactive control element assembly to be stuck in its fully withdrawn position.

The analyzed case 4.ssumes initial hot shutdown conditions at time zero since this represents the most pessimistic initial condition with respect to plant cooldown

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l and return to power. Should the reactor be just critical or operating at power at the time of a steam line break, the reactor will be tripped by the normal high power protection system when power level reaches a trip point. Following a trip at power the RCS contains more stored energy than at no load, the average i coolant temperature is higher than at no load and there is appreciable energy stored in the fuel. Thus, the additional stored energy is removed via the 4 cooldown caused by the steam line break before the no load conditions of RCS temperature and shutdown margin assumed in the analyses are reached. Af ter the additional stored energy has been removed, the cooldown and reactivity i

insertions proceed in the same manner as in the analysis which assumes no load

condition at time zero. However, since RCS stored energy is lower and the i initial steam generator water inventory is greatest at no load, the magnitude and i duration of the RCS cooldown are less for steam line breaks occurring at power and, therefore, the RCS cooldown and potential for return to power is maximized

{ at no load conditions.

The steam line break analysis submitted in Reference (1) includes assumptions i reviewed and approved by the NRC Staff. Specifically, the Basic Safety Report (BSR), submitted in support of Westinghouse reloads of Millstone Unit No. 2, l included the steam line break at no load conditions with offsite power available as the licensing basis for Millstone Unit No. 2. The NRC Staff documented their I acceptance of the analysis in Reference (2).

l The following qualitative discussion of a steam line break at no load conditions I

with loss of offsite power is provided for information. The loss of offsite power affects the steamline break transient principally due to the effects on the rate i of cooldown, reactivity feedback, and the rate of steam generator blowdown. An additional delay would be assumed following the generation of a safety injection signal to start the diesel generators and to commence loading the safety j injection equipment onto them. Since the reactor coolant pumps are coasting down with the loss of offsite power, the ability of the emptying steam generator to extract heat from the reactor coolant system is reduced. The closest i approach to criticality would occur later in the transient and the core power increase would be slower than in the similar case with offsite power available.

Although the steamline break transient is classified as a Condition IV event, the Condition 11 acceptance criterion of no DNB following a return to power has been i applied. This is highly conservative since the occurrence of DNB in small regions 1 of the core would not violate applicable NRC acceptance criteria. For the

! steamline break with loss of offsite power, no violation of the DNB criterion for 1

Millstone Unit No. 2 is anticipated.

Question 440.2 Westinghouse analyzed the steam line break event for the Cycle 6 reload.

However,' Millstone 2 is a Combustion Engineering designed plant. Justify the  ;

use of the Westinghouse mixing factors (used for simulating asymmetric thermal- i hydraulics) on a plant designed by Combustion Engineering. Provide the mixing factors used in the calculation.

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Response

For the BSR analysis, a sensitivity study was performed to assess the impact of different vessel inlet and outlet mixing assumptions on the steamline break transient results. The case study included very good (almost perfect) mixing and poor mixing. The resultant statepoints demonstrate that higher heat flux, higher coolant temperatures, and slightly higher pressures were calculated for the good mixing case. This is due to the fact that Millstone can incur a very high uncontrolled steam release due to its main steam piping design which results in a break area of 6.5 square feet. This higher steam release causes a greater cooldown and depressurization in the reactor coolant system. The safety injection system does not terminate the reactivity excursion, but merely dampens it. These two ef fects, the fast depressurization and the failure of the safety injection system to render the core subcritical, lead to the higher heat flux generation during a steam line break with very good mixing.

In addition, when mixing is good, the upper head temperature tends to fall at a slower rate, due to the f act that more flow from the cold loop is allowed to mix with the hot loop flow in the inlet and outlet of the vessel. Part of the inlet flow is routed to cool the upper head. Since this water is warmer than in the poor mixing case, the upper head temperature would be higher. When the pressurizer empties, the upper head becomes an effective pressurizer and the reactor coolant system pressure is the saturation pressure of the upper head temperature. Therefore, the reactor coolant system pressure is higher with better mixing which impedes the insertion of borated water, allows a higher reactivity level and more heat flux generation. A statepoint evaluation indicates that the impact of the higher heat flux and coolant temperatures more than offset the slightly higher system pressure. Thus, the use of the perfect mixing assumption produces a more conservative transient in terms of DNB.

Question 440.3 In support of continued plant operation with 15.3% plugged steam generator tubes, the licensee analyzed the following events: (1) steam line breaks; (2) loss i of coolant flow; (3) CEA withdrawal; and (4) steam generator tube rupture. Tube l plugging reduces core flow, as it provides added resistance to flow. Tube I plugging also decreases the ef ficiency of the steam generators to remove core heat. It is not obvious that other events need not be analyzed when a significant number of tubes are plugged. For the remaining Chapter 15 events (i.e.,

inadvertent opening of an atmospheric dump valve, feedwater system pipe breaks, reactor pump shaf t seizure, rod ejection accidents, etc.), which have not been documented, describe why tube plugging would not alter previously analyzed conclusions.

Response

NNECO plugged a relatively large number of steam generator tubes (approximately 1000 tubes) during the last refueling outage. To support continued plant operation, an evaluation of the effects of tube plugging on the docketed safety analyses was provided in Reference (3).

To support Cycle 6 operation with up to 15.3% of the steam generator tubes plugged, all licensing basis non-LOCA transients presented in the BSR were reevaluated to determine which could potentially be affected by the fuel reload or by the potential steam generator tube plugging. NNECO notes that only those transients described in the Millstone Unit No. 2 FSAR constitute the licensing basis for the plant. Tube plugging in sufficient numbers results in three effects:

Reactor coolant flow is reduced due to increased steam generator flow resistance.

The primary flow and steam generator heat transfer area are reduced.

Thus to maintain guaranteed steam flow, Tavg must be increased or steam pressure reduced.

Primary reactor coolant mass inventory is reduced.

The basic approach used was to identify the important parameters for each accident, determine which of these parameters were affected by the higher steam generator tube plugging levels, and then determine how the impacted parameters affected the accident analysis. The resulting impacts were determined by either evaluating the accident to qualitatively demonstrate that the accident is not limiting or by reanalyzing the affected accident (if the accident was found to be limiting or very sensitive to the impact of higher steam generator tube plugging levels). The evaluations were consistent with the following assumptions:

Maximum core thermal power, MWt 2700.

Thermal design flow, gpm (plant total) 350000 S.G. tube plugging level, percent 15 Core inlect temperature, OF 549.

RCS pressure, psia 2250 Fr 1.565 Tnoload, OF 532 Fq maximum 2.64 The impact of reduced radial peaking factor (F r) and thermal design flow on the non-LOCA accident analyses presented in the Millstone Unit No. 2 BSR plus subsequent reanalysis has been assessed. In general, all of the transients are sensitive to the steady state primary flow. A study was rnade of each currently applicable accident analysis to identify margins to safety limits which could be used to offset penalties due to reduced primary flow. The reduction in radial peaking factor from the Cycle 5 limit is a benefit in DNB calculations and more than offsets the penalty due to the reduced flow.

For the licensing basis events presented in the Millstone Unit No. 2 FSAR and analyzed in the BSR, NNECO submitted the results of reanalysis of the control element assembly (CEA) withdrawal, loss of coolant flow, steam line break and steam generator tube rupture events to support Cycle 6 operation. A discussion of the impact of the reload and additional steam generator tube plugging on the seized rotor and CEA ejection events is provided below.

i Sgized Rotor The current analysis shows that the most severe Seized Rotor Accident is an instantaneous seizure of a reactor coolant pump rotor at 100 percent power.

Following the incident, reactor coolant system temperature rises until shortly af ter reactor trip.

The impact on the Seized Rotor Accident of increased steam generator tube plugging will be primarily due to the reduced flow. These impacts will not affect the time to DNB since DNB is conservatively assumed to occur at the beginning of the transient. The flow coastdown in the affected loop due to the Seized Rotor is so rapid that the time of reactor trip (Iow flow setpoint is reached) is essentially identical to the current analysis.

It is estimated that the change in peak pressure reported in Reference (4) will be insignificant based on 2250 psia operation plus 30 psia uncertainty. This pressure is significantly below the pressure at which vessel stress limits are exceeded. The 15 percent reduction in steam generator tubes would result in approximately a 5 percent reduction in primary mass which decreases the heat capacity of the RCS by the same amount. This would not result in higher peak temperatures or pressures since the peak values are reached in considerably less than the one loop transport time constant of approximately 12.8 seconds.

Thus operation at reduced flow will not cause safety limits to be exceeded for a seized rotor accident.

CEA Election The rupture of a control element assembly mechanism housing which allows a control element assembly to be rapidly ejected from the core would result in a core thermal power excursion. This power excursion would be limited by the Doppler reactivity effect as a result of the increased fuel temperature and would be terminated by a reactor trip activated by high nuclear power signals.

The CEA ejection transient is analyzed at full power and hot zero power conservatively combining the most limiting BOC and EOC conditions (Section 5.3.14 of the BSR). Reduced core flow is the primary impact resulting from increased levels of steam generator tube plugging. This impact would result in a reduction in heat transfer to the coolant which would increase clad and fuel peak temperatures. The effect of reducing flow by 5 percent is primarily to increase l peak clad temperatures by approximately 500F. The current analysis shows that for all cases a value of approximately 3000F can be accommodated before peak clad limits are reached (27000F).

The fuel temperatures will also increase, however the increase will be much less l than the increase in the clad temperature due to the rapid nature of the l transient. Thus, the conditions at the hot spot fuel rod do not exceed the clad

! damage threshold of 200 cal /gm and the conclusions presented in the BSR are still valid.

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In summary, to assess the effect of non-LOCA accident analyses on operation of Millstone Unit No. 2 with additional steam generator tube plugging, a safety evaluation was performed. The safety evaluation was based in part on reanalyses and in part on qualitative assessments where more detailed investigations were not required.

The transients reanalyzed were control element assembly withdrawal at power, steamline break, steam generator tube rupture and loss of flow. In addition, an evaluation was performed to identify the effect of a flow reduction on the remaining transients and to quantify margins available to offset penalties. Based on this evaluation, operation with an approximate 5 percent reduction in thermal design flow and a maximum of approximately 15 percent effective steam generator tube plugging level will not result in violation of safety limits for the transients evaluated.

References (1) W. G. Counsil letter to R. A. Clark, dated April 13,1983.

(2) R. A. Clark letter to W. G. Counsil, dated January 12,1982.

(3) W. G. Counsil letter to R. A. Clark, dated February 23,1982.

(4) W. G. Counsil letter to R. A. Clark, dated November 17, 1981.