ML20053E626
| ML20053E626 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 11/30/1981 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20053E618 | List: |
| References | |
| NUDOCS 8206090086 | |
| Download: ML20053E626 (23) | |
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ENCLOSURE 1 1
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NORTH ANNA UNIT 1 AND 2 i
582.8'F REACTOR COOLANT SYSTEM NSSS SAFETY EVALUATION
SUMMARY
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WESTINGHOUSE ELECTRIC CORPORATION November, 1981 1
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... '.4 TABLE OF CONTENTS A.
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B.
Conclusions 1
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2.
Comparison of Parameters
-' i 3.- NSSS Accident Analyses 4.
NSSS System Impact 5.
NSSS Component Impact 6.
NSSS/B0P Interface 7.
Technical Specification Revisions 08190:1 1
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OBJECTIVE
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l The objective of this report is twofold as described below:
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- 1) To provide a description of the proposed change in the North Antia Unit 1 and 2 reactor coolant system average temperature (i.e.,
580.3*F-+- 582.8*F) and to assess,the associated impact on the NSSS.
This information is to be utilized by the Virginia Electric and Power Company in their request for amendment of the' plant operating licenses in accordance with the requirement of 10CFR50.90.
- 2) To provide a technical basis for the determination that the proposed change in reactor coolant system average temperature does not involve an ~unreviewed safety question in s:cordance with the requirement of 10CFR50.59.
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CONCLUSION The proposed change in reactor coolant system average temperature has been reviewed and evaluated in detail with respect to the following:
- 1) NSSS Accident Analyses
- 2) NSSS System Adequacy
- 3) NSSS Component Integrity
- 4) NSSS/ Balance of Plant Interfaces
- 5) Technical Specification Impact Based on the fact that the proposed change does not result in violation of any NSSS system or equipment design criteria, and that it is not necessary to revise any of the plant operating procedures it was con-cluded that:
- 1) The probability of accidents or malfunctions of equipment previously evaluated in the FSAR is not ine-aased.
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- 2) The possibility of an accident'or equipment malfunction of a dif-ferent type than any previously evaluated in the FSAR is not created.
- 3) The consequences of accidents or malfunctions of equipment evaluated in the FSAR are not increased.
- 4) The rnargins of safety as defined in the bases to the plant Technical Specifications are not reduced.
Therefore this change does not reduce the plant safety margins and involves no unreviewed safety question as defined by 10CFR50.59.
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I 1 INTRODUCTION i
Ine virginia Eiectric and Power Company (VEPCO) is pursuing a perform-ance optimization program to maximize the electrical output of North I
..... ;;.. : ".. * :nd 2.
One of the areas being investigated is the impact of increased Nuclear Steam Supply System (NSSS) steam generator outlet steam pressure on the electrical generation of the units.
Increasing l
the NSSS steam generator outlet steam pressure at the full thermal load
- o. 50: an is accomplished by increasing the NSSS reactor coolant i
system average temperature.
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A test was conducted at North Anna Unit 2 on October 29,1980, to l
determine the effect of varying steam generator outlet steam pressure on gross electrical generation. The test revealed that at 2785 MWt a 4*F variation in MSSS reactor coolant system average temperature resulted in a 25 psi change in steam generator outlet pressure and a 2.5 MYA variation in gross electrical generation. This increase in gross i
electrical geri: ration was solely attributable to the additional energy l
contained in the higher pressure steam. There was no increase in the i
i thermal output of the NSSS.
A review of the North Anna FSAR Section 15.1.2.2.reveeled that the cur-
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rently docketed accident analyses, accounting for measurement and con-trol uncertainties, are based on a reactor coolant system average temperature which is 2.5'F greater than that presently allowed by the Technical Specifications for the North Anna units.
Based on the data i
obtained from the gross generation versus steam pressure test, increasing the reactor coolant system average temperature by 2.5'F would l
result in approximately an 18 psi increment in steam pressure which corresponds to approximately a 1.8 MVA improvement in gross electrical l
gcneration per unit.
Changing the North Anna 1 and 2 Technical Specifications to reflect the value of the reactor coolant system average temperature currently I
utilized in the docketed FSAR accident analyses would result
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I in imreased electrical g:;neration from th3 units with essentially no i ::::t on the currently approved licensing documentation and therefore should require minimal regulatory review.
The Westinghouse scope of effort in this report is as follows:
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Verify margins for NSSS accident analyses.
- 2) Confirm NSSS systems adequacy.
- 3) Confirm NSSS component integrity.
4)
Identify any Balame of Plant (B0P) interface revisions.
- 5) Cite recommended revisions to the North Anna 1 and 2 Technical Specifications.
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e 2 COMPARISON OF PARAMETERS The NSSS steam generator outlet steam pressure is determined by the
'-,----'"- ; and temperature dif ferential s between the primary and secondary side of the steam generator tubes. The control systems of the plant are based on producing essentially " dry" steam for input to the turbine, therefore, the steam pressure for practical purposes is fixed
. trgeratures on the primary side of the steam generator tubes.
The steam generator steam flow is primarily dependent on the NSSS thermal power. The NSSS thermal rating will r'emain at its currently licensed value of 2785 MWt (Core heat output = 2775 MWt: Reactor Coolant Pump net heat input = 10 MWt). At rated thermal load, increasing the average temperature on the primary side of the steam generator tubes by 2.5'F will increase the temperature of the steam on the secondary side by approximately 2.4*F, which corresponds to an 18 psi increase in steam pressure.
Table 2.1 contains a comparison of the current and proposed Reactor Coolant System (RCS) temperatures and flow rates at rated thermal From the table it can be seen that-the RCS thermal rating, power.
pressure, core inlet volumetric flow rate and "no load" temperature remain at the current values. The core inlet mass flow rate has decreased slightly due to the reduction in coolant density associated with the increased core inlet temperature. The RCS' average temperature has been increased from 580.3*F to 582.8'F to reflect the additional 2.5'F contained in the docketed accident analyses. The variations in inlet temperature and temperature rises are attributable to the thermo-dynamic properties of compressed liquid water. Figure 2-1 graphically depicts the reactor vessel cold leg, average and hot leg fluid tempera-tures as a function of power level for both the, current and proposed operating conditions. The variations in temperature between operation at the proposed and current parameters decrease to zero as power i:,
reduced from rated thermal load (i.e., 2785 MWt) to no load conditions.
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it.'c ::;t:,mation presented in Table 2.1 and Figure 2-1 represent only the
'.f impac t on the plant parameters and operating character-t i stic s.
The remainder of this report reviews the impac t of the 2.5*F
' rr:::: in P.CS average temperature on NSSS accident analyses, NSSS systems, NSSS components, NSSS/ BOP interfaces and Technical Spec i fic ations.
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e TABLE 2.1 COMPARISON OF REACTOR COOLANT SYSTEM PARAMETERS Thermal and Hydraulic Design Parameters Design Conditions Current Proposed USSS Pcwor, MWt 2785 2785 Reactor Core Heat Output, MWt 2775 2775 System Pressure, Nominal psia
- 2250 2250 System Pressure, Min., Steady State, psia 2220 2220 Total Core Inlet Thermal Flow Rate, gpm 278,400 278,400 6
6 Total Core Inlet Thermal Flow Rate, lbm/hr 105.1 x 10 104.7 x 10 Core Effective Flow Rate for Heat Transfer, 6
6 lbm/hr 100.4 x 10 100.0 x 10
. Reactor Coolant System Temperatures, *F Nominal Reactor Vessel / Core Inlet 546.9 549.5 Average Rise in Vessel 66.9 66.6 l
Average Rise in Core 69.7 69.4 Average in Core 583.6 586.1 Average in Yessel 580.3 582.8 No Load 547.0 547.0 i
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FIGURE 2-1 NORTH AriNA I AND II REACTOR C00LAfli TEMPERATURES
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r,OF RATED THERMAL LOAD f REACTOR VESSEL
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HOT LEG TEMPERATURE
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REACTOR VESSEL AVERAGE f
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TEMPEM E RE
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REACTOR VESSEL
- COLD LEG TEMPERATURE
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Tavg = 580.3 Current Tavg = 582.8 Proposed 530 i
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'20 40 60 80 100
- f. RATED THERMAL LOAD 9
c 3 NSSS ACCIDENT ANALYSES A review was performed to assess the impact of a 2.5'F increase in RCS average temperature (i.e., 580.3*F to 582.8'F) on the docketed North Anna Unit I and 2 postulated accident analyses.
Section 15.1.2.2 of the F5Ai< inoicates that the original design bases for the accident analyses included a 2.5'F additional allowance on temperatures.
"An additional allowance of 2.5'F has been made in the core t.cr.perature to allow for steady state operation at nominal average temperatures up to 2.5'F greater t,han the design value of 580.3*F without invalidat.ing any accident analyses. Thus, all accident analyses are performed at either the design average temperature of 580.3*F + 6.5'F or at 580.3*F - 4*F, which ever is more conservative."
The docketed FSAR accident analyses and design bases were reviewed (e.g., nuclear design, DNB, non-LOCA and LOCA) to verify that the tem-peratures assumed correspond to an RCS average temperature of 580.3*F
+6. 5*F or -4*F.
An uncertainty of + 4*F is required to envelope the temperature measurement accuracies and the "deadband" associated with the automatic Rod Control Cluster control system. Therefore, the existing FSAR analyses are adequate for operation at 582.8'F + 4*F.
For transients postulated to initiate at "no load" conditions the docketed temperature of 547*F remains unchanged.
A detailed review of the Technical Specification reactor protection setpoints revealed that, with the exception of the overpower and over-temperature aT setpoints, the docketed setpoints are appropriate for the RCS average temperature of 582.8'F. The overtemperature and overpower AT trip setpoints were determined for the revised nominal RCS average temperature. An analysis was then perfomed to verify that these new setpoints provide adequate protection against DNB. Additional discus-sion of recommended revisions to the Technical Specifications is provided in Section 7 In summary the docketed FSAR accident analyses envelope NSSS full power
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operation at 2785 MWt with an RCS average ~ temperatura of 582.8'F.
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4 NS$5 SYSTEMS IMPACT
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n evelvation was performed to determine the influence of increasing the Reactor Coolant System average temperature by 2.5*F to 582.8*F on the USSS tystems. The systems reviewed were:
Reactor Coolant System Chemical and Volume Control System Residual Heat Removal System Safety Injection System Reactor Control and Protection System Each system evaluation verified the adequacy of the system design during normal and postulated transient operating conditions.
In addition any revisions to system functional requirements and/or design criteria associated with the increased temperature were identified and reviewed by the cognizant component design func tion.
The 2.5*F temperature increase has no impact on either the Residual Heat Removal or Safety Injection Systems. These systems are not functional during normal operation. The function of the RHR system is to cool the NSSS from approximately 350*F. to a cold shutdown condition. The prime function of the SI system is to provide cooling flow to the core in the event of severe transients. The cooldown and ECCS functional require-ments and design criteria are not impacted by this normal operating temperature change.
With respect to the Chemical and Volume Control System (CVCS) the 2.5*F increased temperature has a minor impact on the operating temperatures of the CVCS heat exchangers. The need for revised functional require-ments was evaluated relative to the existing heat exchangers (i.e.,
regenerative, non-regenerative, excess letdown and seal water).
It was detennined that the installed North Anna CVCS heat exchangers are in compliance with the original system design bases when operated at the elevated temperatures. Operation of the CVCS heat exchangers with an RCS average temperature of 582.8*F will result in minor increases in heat load to the Component Cooling Water System (CCWS) relative to 0819Q: 1 11 4
1 current operation at 580.3*F. The ircreased CCWS heat loads are ci::::,cd in Section 6.
Th2 corponents downstream of the CVCS heat
"*=ny" (e.g., piping, tanks, filters, pumps) are unaf fected by the increased temperatures.
E valuation of the Reactor Coolant System at the increased temperature revealed that no significant revisions in functional requirements and design bases are necessary. For example, with respect to postulated tactor Coolant System transients for systems and equipment design te.g., neat-up, cooldown, reactor trip, etc.) the original design bases reflected temperatures and core thermal power in excess of those associated with the proposed change. The revised nominal operating temperatures were forwarded to the cognizant RCS systems and component design functions for review. The result of the equipment evaluation is presented in Section 5.
Cased on this evaluation the increased RCS average temperature was found to be within the original design envelope of the Reactor Coolant System.
In addition to the evaluation of the increased temperatures on the Reactor Protection System setpoints which is discussed in Section 7, the Reactor Control System was reviewed to provide a sunnary of recommended revisions to control functions. An example of this type of review is the development of the RCS average temperature as a fum tion of ther-mal power based on the increased full thermal load temperature with the no-load temperature remaining at the current value of 547'F. The major-ity of the control system information is contained in the North Anna Precautions, Limitations and Setpoints (PLS) document.
Recommended modifications to the Reactor Control System software have been forwarded to VEPCo via revisions to the North Anna PLS. The North Anna Reference Operating Instructions were reviewed and it was determined that no revisions were required for an RCS average temperature of 582.8'F.
In summary, the review of the NSSS systems indicates that the proposed Reactor Coolant System average temperature of 582.8*F is enveloped by the original design bases, critaria and furetional requirements. Where appropriate, revisions to the North Anna Precautions Limitations and Setpoints document and Balarte of Plant interfaces have been provided to VEPCo for review and implementation.
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5 NSSS COMPONENT IMPACT
.. -....__... ::: 5 component design functions determined the impact of a 582.8*F RCS average temperature on the NSSS equipment. Tb2 evaluation
' "-' the standards and design criteria applied to the original license were applicable for the proposed change. The system information of prime importance to the equipmerit designer is the design conditions (temperature, pressure, etc.), seismic response, LOCA loadings,
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' avents and nominal temperatures. The proposed increase in RCS average temperature has no impact on the design conditions and the seismic response. An increase in fluid temperature results in a decrease in LOCA loadings, and therefore the original LOCA loads are conservative for the proposed 582.8'F RCS average temperature. This is because the decompression wave which hydraulically loads the vessel internals, loop and component supports is proportional to the difference between normal operating pressure and saturation pressure.
Operation at higher temperatures causes a reduction in this differential pressure and thus reduced loads.
During the original design an evaluation was performed to determine the type and number of occurrences of plant operational transients which constitute the bases for analyzing and evaluating the cyclic behavior of the NSSS components. A description of these transient events is provided in Section 5.2 of the FSAR. The transient evaluation was performed based on temperatures, pressures and power levels in excess of the actual plant rating. The transient analyses provide each component design group with information regarding variations in temperature, power, pressure and flow rate.of the unit for use in structural integrity evaluations of the NSSS components. As discussed in Section 4, review of the transients applicable to North Anna revealed that the original design basis transients remain conservative for an RCS average temperature of 582.8*F.
The nominal operating cold leg, hot leg and average temperature of the RCS have increased. As a result it was necessary to verify that each installed component and the associated analyses are in compliance with the design codes, standards and criteria in effect at the time of the original license for the revised nominal operating conditions.
In a few 13 0819Q:1
v instances it was necessary to revise the documented analyses to account for the t ry:reased RCS terperatures. Three levels of ef fort were utilirad for this review. Each of the three levels and the components I
in eacn level are discussed below.
A.
The first level of effort was to identify for which f1SSS systems and associated components no change in the original design bases and functional requirements was required. For these components and/or systems, which are listed below, no additional effort was required with respect to the RCS average temperature increase.
- 1) Residual Heat Removal System
- 2) Safety Injection System
- 3) Pressurizer Spray, Power Operated Relief and Safety Valves
- 4) Chemical and Volume Control System except for Heat Exchangers B.
The second level of effort was to identify for which 11555 components the Increased RCS temperatures were bounded by analyses performed for a generic design or where a unit with the identical component was evaluated at duty ratings equal to or greater than those associ-ated with the proposed change. The components in this category are:
- 1) Reactor Coolant Pumps
- 2) Control Rod Drive Mechanisms
- 3) Reactor Coolant Loop Piping
- 4) Reactor Coolant Loop Isolation Valves
- 5) CYCS Heat Exchangers
- 6) Pressurizer
- 7) Upper React'or Internals Assembly C.
The third level of effort was to confirm analytically compliance with the applicable design codes, standards'and criteria for specific instances where the increased RCS temperatures were not bounded by analyses performed for a generic design or for a unit with the identical components at duty ratings equal to or greater 08190:1 14
i than those astociated with the proposed change. The three compon-ents in this category and a description of the associated effort is
. discussed below:
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Reactor Vessel Review of the reactor vessel documentation revealed that the existing analyses bound the p'roposed change with the exception of the unit loading transient. For this particular transient minus T I II
- the proposed temperature change (TH0T N0 LOAD no load to full load conditions is approximately 2.6*F greater than originally evaluated.
Analyses were performed for an even greater temperature change than proposed.
These analyses indi-cate that the change has an insignificant impact on the stress range and fatigue life of the vessel, and therefore is accept-able.
- 2) Lower Reactor Internals Assembly i
The North Anna 1 and 2 lower reactor internals assembly were reviewed relative to other 3 loop internals at duty ratings in excess of those associated with the subject temperature increase. With the exception of the item discussed below, the North Anna Unit 1 and 2 lower reactor internals are geomet-rics11y identical to higher duty rated structures.
Therefore, with the exception of the item in the following discussion, the analyses performed for higher duty 3 loop plants bound the proposed North Anna 1 and 2 increased temperatures. With respect to the one geometrical variation, the following discus-sion demonstrates that the increased temperatures have no significant impact on either the existing component or the associated analyses.
Thermal Shield - The higher duty rated structures which were reviewed all utilize neutron panels in lieu of the thermal ihield employed on the North Anna design.
The prime concerns relative to the thermal shield are related to flow loadings and 08190: 1 15
the potential for flow induced vibrations. A 2.5*F increase in temperature has an insignificant impact on the flow rate, and therefore creates no hardware concern.
In addition, a 2.5*F variation in temperature is within the accuracy of the existing thermal analyses. Therefore, the installed North Anna 1 and 2 lower reactor internals assemblies are in compliance with the design codes, standards and criteria applied at the time of the original ifcense.
- 3) Steam Generator Review of the steam generator documentation revealed that the current documentation bounds the proposed pressure and temperature changes with the exception of specific evaluations for the steam generator shell in the vicinity of the upper lateral supports and the feedwater nozzle. For these exceptions it was necessary to perform detailed analyses to verify compliance with design criteria for the increased pressure loading at NSSS full thermal power. These analyses consider loads imparted simultaneously to the components due to supports, weight, temperature, flow, pressure and seismic. The analyses performed assumed the original design basis loadings on the component remain unchanged, with the exception of the pressure loadings. A pressure value of 1020 PSIA was assumed with full thermal load steain flow in the generator. This pressure corresponds to the no load value and represents an upper bound value for the evaluation. The results of the analyses demonstrate that the steam generators remain in compliance with the applicable design criteria.
In sumary, the review of majority of the NSSS components indicates that the proposed Reactor Coolant System average temperature of 582.8'F is enveloped by either the original North Anna analyses or analyses for other 3 loop plants with identical structures at a higher duty rating.
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i Fo'r :pe:ific components where additional analyses were n:cessary, it tras
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6*crmf.921 that the structures remain in compliance with the design codes, standards and criteria applied at the time of the original j
license.
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.m 6 HSSS/3]P INTERFACE in an ef fort to coordinate the NSSS review with the related review of the Balan:e of Plant (30P), a program was est3blished to identify areas in which the 2.5'F increase in Reactor Coolant System (RCS) average t?mperature could have an impact on the BOP design. This temperature
'n:rease corresponds approximately to an 18 psi increase in steam
- nerator outlet steam pressure at the full thermal loading of 2785 MWt.
During the course of the evaluation it was determined that a number of NSSS/B0P interfaces were not impacted by the 2.5'F increase in RCS average temperature. The interfaces in this category and a brief explanation of why the existing interface data is not impacted are dis-cussed below:
- 1) Mass and Energy Release Data - The original Loss of Coolant Accident data for containment integrity evaluations was based on an NSSS rating of 2910 MWt/850 psi steam pressure.
This data bounds the proposed temperature increase on the RCS at 2785 MWt. The Main Steamline Break data was based on the event occurring at the no load condition which is on: hanged by the increased nominal operating temperatures.
- 2) Auxiliary Feedwater System - The original auxiliary feedwater system requirements were based on an "SSS rating of 2910 MWt/850 psi steam pressure. The 2.5*F RCS average temperature increase at 2785 MWt is bounded by the original analyses. Therefore, the auxiliary feedwiter system requirements are unchanged.
3)
Source Terms for Offsite Dose Evaluations - The data currently in the FSAR are based on a core power of' 2900 MWt. These source terms are essentially a function of core power and burnup only The impact of the 2.5*F temperature increase on burnup is negli-gible. Therefore, the current FSAR data remail unchanged.
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f 4) 0;ent Fuel Pit Decay Haat Loads - The decay heat of the fusi in the spent fuel pit is a function of core power and burnup and is independent of the 2.5'F average temperature increase. The data employed in the original design evaluation remain unchanged.
5)
Steam System Design Transient - The steam system transients provided in the Westinghouse Steam Systems Design Manual are unchanged.
o) RCS Loop Pipe Loads, Thermal Displacements and Design Data -
Based on a detailed review it was determined than any changes in loadings and piping / support thermal displacements and other design data are within the bounds of the original evaluation.
- 7) Flux Mapping System - It was determined that the effect of the increased temperature is within the accuracy of the original design data.
A number of areas have been identified for which it is recommended that the Balance of Plant systems be reviewed to evaluate the effects of the proposed change. These areas are:
- 1) Condensate and Feedwater Systems - Review the system performance and component adequacy at the increased pressure to ensure satisfactory operation.
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- 2) Main Steam System - Perform the same review as outlined for the I
condensate and feedwater system.
3)
Component Cooling Water System - The increased RCS inlet tem-i perature does increase the heat loadings on the Component i
Cooling Water System (CCWS) due to the Chemical and Volume Control System heat exchangers. Table 6.1 provides a summary of the heat loadings on the CCWS for the current RCS cold leg 0819Q:1 19
temperature of 546.8'F and a value of up to 555.5'F. The heat loads are based on a maximum CCWS inlet temperature of 105'F to each heat exchanger. The impact of the increased CYCS heat exchanger heat loads on the CCWS should be reviewed to assure snat the maximum inlet temperature of 105'F remains unchanged.
An additional area of review was the Westinghouse supplied turbine generators.
Based on a detailed review, it was determined that the assumptions, analyses and evaluations (e.g. turbine missiles) performed to verify the operating characteristics and structural integrity of the turbine generator at the current operating parameters bound the conditions resulting from the 2.5'F RCS average temperature increase to 582.8'F.
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TABLE 6.1 INCREASED CCWS HEAT LOADS RESULTING FROM CYCS OPERATION UNDER INCREASED RCS TEMPERATURES 6
Heat Exchanger / Condition Heat Transfer Rate (BTU /HR x 10 )
RCS Cold leg Temp.
RCS Cold Leg Temp.
546.8'F 555.5*F Non-Regenerative:
Normal Operatien 5.10 5.28 Maximum Purification 11.1 11.1 Heat Up (Design) 16.1 16.1 Excess Letdown:
3.26 Design 3.08 3
Seal Water Return:
Normal Operation 0.92 1.02 Design 1.25 1.37
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7 TECHNICAL SPECIFICATION REVISIONS A review of the North Anna Unic 1 and 2 Technical Specifications was vs. fusiued to establish revisions necessary to reflect the 2.5'F increase in Reactor Coolant System average temperature to 58?.8'F at full thermal load of 2785 MWt.
i.*! 2. '.he exception of the overtemperature and overpower AT setpoints (Table 2.2-1) and Table 3.2-1, all of the Technical Specification data are appropriate for an RCS average temperature of 582.8'F. The calcula-tion of the currently licensed overpower and overtemperature aT set-points and associated constants was based on a nominal RCS average tem-perature of 580.3*F at 2785 MWt. Analyses were perfonned to determine the overpower and overtemperature aT setpoints for an RCS average tem-perature o f 582.8*F. Confirmatory analyses were performed to verify that the revised constants and resulting setpoints are appropriate. The confirmatory analyses, consisted of,evaluati.ng the rod withdrawal at power transient over a bounding range of reactivity insertion rates.
.This transient is limiting with respect to these setpoints.
The only modification required for-the Reactor Protection System is incorporation of the Technical Specification revisions identified in 22 1311Q: 1
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