ML20112A702
| ML20112A702 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/11/1985 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton, John Miller Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20112A704 | List: |
| References | |
| 85-086, 85-86, NUDOCS 8503180422 | |
| Download: ML20112A702 (53) | |
Text
'
WauerL Snwr Nuckw qrmtkens D7wtnuest V::eIWsuksst itst OfutlitaXfM NuukwQnurunas Onejanus Ravritua
~ =e Ruhnerakt Virginia B261 March 11, 1985 VIRGINIA POWElf Mr. Harold R. Denton, Director Serial No.85-086 Office of Nuclear Reactor Regulation ESC / JOE:acm Attn:
Mr. James R. Miller, Chief Docket Nos. 50-338 Operating Reactors Branch No. 3 50-339 Division of Licensing License Nos. NPF-4 U. S. Nuclear Regulatory Commission NPF-7 f
Washington, D. C. 20555 f
Gentlemen:
VIRGINIA POWER AMENDMENT TO OPERATING LICENSES NPF-4 AND NPF-7 NORTH ANNA POWER STATION UNIT NOS. 1 AND 2 PROPOSED TECHNICAL SPECIFICATION CHANGES Pursuant to 10CFR50.90, Virginia Power requests an amendment, in the form of changes to the Technical Specifications, to Operating License Nos.
NPF-4 and NPF-7 for the North Anna Power Station Units 1 and 2.
The proposed changes and the supporting safety evaluation are enclosed.
There are many problems associated with the high boric acid concentration which must be maintained in the boron injection tanks and concentrated boric acid system. During normal operation reactor coolant letdown is concentrated and recycled to the boric acid tanks via the Boron Recovery System resulting in high radiation levels. These high radiation levels compound the maintenance problems caused by the high boric acid concentrations since the boric acid is a highly corrosive fluid. Leakage has led to the degradation of carbon steel components and the failure of heat tracing which is required to maintain solution solubility. Failure of heat tracing results in additional maintenance problems such as boron plateout and potential line plugging as the solution temperature drops.
The increased maintenance causes increased personnel radiation exposures.
A reduction in boric acid concentration would reduce maintenance requirements and the associated exposures to plant personnel.
This submittal is part of Virginia Power's ongoing effort to reduce the maintenance requirements and the radiation exposure to plant personnel. A similar program for our Surry Unit Nos. 1 and 2 resulted in the h7C's approval (SPC letter dated February 24, 1984) for removal of the Boron Injection Tank (BIT) at those two units. provides the detailed justification for a proposed reduction in the minimum baron concentration at North Anna in the BIT and in the boric acid system from 11.5 wt % to 7.4 wt %.
A general description of the current design of the Boron Injection Tank and Boric Acid System is given.
The proposed changes are described and operational and maintenance h
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benefits are discussed. There are no physical changes required to any of the affected systems to implement this change.
Analyses have been performed to determine the impact of the proposed changes on the appropriate North Anna licensing bases, including. a reanalysis of the steamline break accidents discussed in Chapter 15 of the North Anna Updated Final Safety Analysis Report (UFSAR). The analysis has been performed by Virginia Power, using the RETRAN Computer Code and the reactor system transient analysis methodoloby described in our topical report which was transmitted by letter dated April 14, 1981 (Serial No.
215).
The methodology, assumptions and results of the analysis are discussed in detail in Attachment 1.
This documentation will be incorporated into the North Anna UFSAR at the next annual update following NRC approval.
The changes to the Technical Specifications associated with the proposed boron concentration reduction are provided in Attachments 2 and 3.
This request has been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Safety Evaluation and Control staff. It has been determined that this request does not involve any unreviewed safety questions as defined in 10CFR50.59 or a significant hazards consideration as defined in 10CFR50.92.
We have evaluated this request in accordance with the criteria in 10CFR170.12. A check in the amount of $150, is enclosed as an application fee.
Very truly yours, h4V W. L. Stewart
Enclosures:
(1) Safety Evaluation for Boron Reduction (2) Proposed Technical Specification Changes, Unit 1 (3) Proposed Technical Specification Changes, Unit 2 (4) Voucher Check for $150
i na cc:
Dr. J. Nelson Grace Regional Administrator Region II Mr. Leon B. Engle NRC Project Manager - North Anna Operating Reactors Branch No. 3 Division of Licensing Mr. M. W. Branch NRC Resident Inspector North Anna Power Station Mr. Charles Price Department of Health 109 Governor Street Richmond, Virginia 23219
.\\
s CC9900NWEALTH OF VIRGINIA )
)
CITY OF RICHMOND
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The foregoing document was acknowledged before me, in and for the City and Commonwealth aforesaid, today by W.
L.
Stewart who is Vice President Nuclear Operations, of the Virginia Electric and Power Company.
He is duly authorized to execute and file the foregoing document in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.
Cfhe
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Acknowledged before me this
// day of
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My Commission expires:
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's PAGE 1
s ATTACHMENT 1 SAFETY EVALUATION FOR REDUCTION IN BOROM CONCENTRATION IN THE BOROM INJECTION TAMK AMD CONCENTRATED BORIC ACID SYSTEM MORTH AMMA POWER STATION UNITS 1 AMD 2 J1
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w PAGE 2
A.
INTRODUCTION 1.
Objective The nuclear industry has experienced problems with the high boron concentrations which have traditionally been maintained in various station systems, such as the Boron Injection Tank (BIT),
associated Safety Injection (SI) lines and the concentrated Boric Acid System.
Boric Acid is a corrosive fluid and the leakage of boric acid can lead to the degradation of carbon steel components (such as supports and bolts, etc.) and the failure of line heat tracing which is required to keep the solution temperature above the solubility limit.
The failure of heat tracing can in turn lead to boron precipitation and line plugging as the solution temperature drops.
At many power
- stations, these problems have resulted in extensive maintenance and excessive radiation exposures to station personnel.
At North Anna power Station the radiation levels associated with the BIT and Concentrated Boric Acid System have resulted from the concentrating and recycling of reactor coolant letdown to the Boric Acid Tanks (BAT) via the Boron Recovery System.
Like several other utilities, Vepco has been looking at ways to reduce the boric acid concentrations to help alleviate the maintenance and ALARA concerns.
The following sections describe the design functions of the Boron Injection Tank and the concentrated Boric Acid System and provide the justification for reducing the minimum boron concentration requirements in I
k)
PAGE 3
order to reduce these maintenance problems and thus also reduce the associated personnel radiation exposures.
The proposed reduction consists of a change i-n the Boron Injection Tank (BIT) and Boric-Acid System concentration from a range of 11.5% - 13%
to 7.4%
9.0X boric acid solution (weight percent).
The reduction in BIT concentration can be achieved for the steam line break accident analyses while still meeting all applicable acceptance criteria.
The reduction in boric acid system concentration can be accomplished by increasing the minimum allowable Boric Acid Tank inventory associated with each unit-from 4450. gallons to 6000
- gallons, thereby preserving the capability for cold safe shutdown at any time in life with the most reactivo control rod assembly withdrawn from the core.
Section A.2 provides a general description of the current design of the Boron Injection Tank and Boric Acid System and describes the proposed setpoint changes to each system.
The operational and maintenance benefits of the proposed changes are discussed in Section A.3.
Evaluations and analyses were performed to assess the impact of the proposed boron concentration reduction upon the existing North Anna accident analyses.
It was determined that a boron concentration reduction in the BIT only affects the results of the steamline break transient, the accidental depressurization of the main steam system and the spurious operation of the SI system at power.
A detailed discussion of R'
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PAGE 4
the supporting analyses and evaluations performed for these
' transients is provided in section B of this attachment.
The proposed boron concentration ' reduction in the BAT does not impact any-of the conclusions of the accident analyses presented in Chapter 15 of the UFSAR.
Section C
presents an evaluation of the impact of the proposed plant modifications on plant operations and the results of a review of the North Anna UFSAR.
A.2 Changes to Current System Operations A.2.1 Boron Injection Tan't The Boron Injection Tank (BIT) is a 900 gallon carbon steel tank which is internally clad with stainless steel and is part of the Safety Injection Systems it contains boric acid solution at a concentration of 11.5X to 13.0% by weight.
Redundant tank heaters and line heat tracing are provided to maintain a minimum solution temperature at the Technical Specifications limit of 145 degrees F, thus preventing boron precipitation.
Recirculation from the BIT to the Boric Acid -Tank is maintained continuously via a Boric Acid Transfer Pump to ensure that the BIT is full of concentrated boric acid at all times and to prevent cold spots and stratification within the tank.
The BIT is isolated from the Reactor Coolant
~
System and the Charging Pumps during normal plant operation by two sets of parallel isolation valves.
Figure 1 illustrates the system design as-described above.
A
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The purpose of the BIT is to provide injection of highly concentrated boric acid to the' Reactor Coolant System to mitigate the. positive reactivity addition resulting from the RCS cooldown during a main steam line break accident. Operation of the BIT, which takes place upon actuation of a Safety Injection Signal, does not impact.any of the accident analysis results presented in Chapter 15 of the Updated Final Safety Analysis Report other than the steamline
- break, accidental depressurization of the main steam system and the spurious'SI transients.
During Safety Injection, the suction of the high head safety injection / charging pumps is diverted from the normal suction at the Volume Control Tank (VCT) to the Refueling Water Storage Tank (RWST).
The Safety Injection flow path through the BIT is established by the opening of. the redundant parallel isolation valves upon a Safety Injection signal.
Concurrent with the opening of the BIT isolation valves is the closing of the redundant isolation valves in the recirculation line to the BAT.
Flow from the safety injection / charging Pumps is introduced into the BIT via a sparger type inlet that distributes the incoming boric acid solution in a 360 degree fan as it enters the tank.
This prevents channeling and also ensures radial homogeneity of the boric acid solution.
Current plans are to reduce the required boric acid j
-concentration in the BIT to a range of 7.4 to 9.0% uith no physical modifications to the plant.
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PAGE 6
A.2.2 Boric Acid System The concentrated boric acid system is a part of the Chemical and Volume control System described in Section 9.3.4 of the Updated Final Safety Analysis Report.
The purpose of the system is to provide an inventory of concentrated boric acid for (1) chemical shim reactivity control, (2) providing makeup to the Reactor Coolant
- System, Refueling Water Storage Tank, spent fuel pit and refueling l
cavity as necessary and (3) recirculation of boric acid through the BIT via the boric acid transfer pumps.
The system is shared between the two North Anna units.
It consists of three Boric Acid Tanks, four boric acid transfer pumps, one batch tank, boric acid filters and associated
- piping, valves, heat
- tracing, controls and instrumentation.
The Boric Acid Tanks are sized to provide sufficient boric acid to bring both reactors to cold shutdown conditions assuming a stuck control rod.
A simplified schematic of the system is shown in Figure 2.
The three Boric Acid Tanks (BAT), which serve as the reservoirs for boric acid inventory, are 7500 gallon stainless steel tanks and are designed for atmospheric pressure.
A boric acid solution of i
11.5%
to 13%
by weight is currently maintained at all times.
The upper concentration limit of 13%
is established to ensure concentrations low enough to remain soluble at a 145 degrees F
~
minimum temperature, which is maintained by redundant tank immersion heaters and line heat tracing.
During normal operation, boric acid is supplied to the boric l
PAGE 7
acid tanks 'from the Boric Acid Batching Tank via the Boric Acid Transfer pumps-or from the Boron Recovery System to maintain a minimum volume of 4450 gallons dedicated to each unit.
A reduction in the boron concentration requirements for the Boric Acid System to a range of 7.4 to 9.0% uill require increasing the minimum volume of boric acid stored for each operating unit to 6000 gallons.
This can be accomplished by resetting the existing level instrumentation and alarms for the new minimum low level. With this increase in volume requirement for each unit, the capability to bring the units to cold shutdown conditions, with the most reactive control rod assembly withdrawn from the core, is preserved.
BAT heater controls and the system heat tracing controls will be reset to maintain a minimum of >115'T to maintain solution solubility.
In summary, the proposed reduction in boric acid concentration will require the following:
1.
Reset Boric Acid Tank Level instrumentation and alarms for a minimum volume.of'6000 gallons.
2.
Reset Tank heater and heat tracing controls to maintain a minimum solution temperature of >115'T
.A.3 Benefits
~
Reducing the minimum required boric acid concentration for the BIT and the Boric Acid System will improve heat tracing system performance by reducing the temperature that must be maintained to js-
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j PAGE 8
prevent boron precipitation and by reducing system heat losses. This will result in a reduction in anticipated maintenance requirements (i.e.
less heat tracing f ail ures,-less line blockages due to boron precipitation), an increase in system reliability and a reduction in future radiation exposures for station personnel.
The reduced boron concentrations should also lead to slightly lower radiation levels associated with the BIT and Boric Acid System.
The reduction in BIT boron concentration will also reduce the RCS dilution required for a return to power in the event of an inadvertant Safety Injection.
This will reduce the amount of letdown which must be processed by the Boron Recovery and Waste Handling systems.
In
- summary, the proposed reductions in BIT and Boric Acid system concentrations offer significant benefits to Vepco in terms of increased operational reliability, reduced maintenance costs and decreased personnel radiation exposures.
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PAGE 11 B.
ACCIDENT ANALYSIS AND EVALUATION B.1 Accidents Evaluated The existing North Anna accident analyses presented in the UFSAR were-evaluated for potential impact from the proposed boron concentration reduction.
Those analyses having a potential impact are
'the Spurious Operation of the Safety Injection System at power, the Main Steamline Break and Accidental Depressurization of the Main Steam System.
An evaluat' ion of the MSSS response to an Inadvertant Operation of the SIS at power was performed for the reduced boron concentration. This ovaluation -incorporated the assumptions in the exising UTSAR analysis along with the revised boron concentration.
It was determined that the lower boron concentration affects only the timing involved for this transient.
As noted in the original transient evaluation, this change produces a minor effect*on-the overall transient by reducing the rate at which the core reactivity is changed.
This effect is similar to the offect that would be noted if the conditions chosen for the current
-onalysis were at a different time in the core lifetime which was less responsive to this transient.
Thus steady state operation at power remains the bounding condition for this transient.
The proposed boron. concentration reduction will impact the existing
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calculated results for the MSSS response to the range of Main Steamline Breaks.
The existing casas of this analysis presented in the UFSAR were reanalyzed.
The results are presented in Section B.2 in the UFSAR
PAGE 12 format.
The Accidential Depressurization of the Main Steam System analysis in the UFSAR' considers two tr ansie nts'; the inadvertent opening of the the largest capacity single steam dump, relief, or safety valve, and a failure of the -decay heat release piping.
Since the results of these two transients are very similar, only the inadvertent opening of the largest _ capacity valve was reanalyzed for this set of conditions.
Results of this analysis are presented in rection B.3.
The severity of the results of these two cases is bounded by the results of the main steamline break analyses that were performed.
The containment response following a Main Steamline Break could be impacted ~if the proposed changes affect the amount of energy released from the steam generators.
A conservative evaluation of the mass and energy releases following the limiting steamline break case was performed accounting for the affects of the boron concentration reduction.
The containment impact was evaluated by comparing these releases to those in the existing UFSAR analyses. It was concluded that the mass and energy releases, and thus the containment response, of the exising analysis remain bounding.
O 1
'o PAGE 13 B.2 Main Steamline Break Updated UFSAR Accident Analysis Section 15.4.2 MAJOR SECONDARY SYSTEM pipe RUPTURE 15.4.2.1 Rupture of a Main Steam Line 14.5.2.1.1 Identification of Causes and Accident Description The steam release arising from a rupture of a main steam pipe would result in an initial increase in steam flow, which decreases as the steam pressure falls. The energy removal from the reactor coolant system causes a reduction of coolant temperature and pressure. In the presence
~
of a negative moderator temperature coefficient, the cooldown results in a
reduction of core shutdown margin. If the most reactive rod cluster control assembly is assumed stuck in its fully withdrawn position after reactor
- trip, there is an increased possibility that the core will become critical and return to power. A return to power follouing a steam Pi e rupture is a potential problem mainly because of this high power P
peaking factors that
- exist, assuming the most reactive rod cluster control assembly to be stuck in its fully withdrawn position. The core is ultimately shut down by the boric acid injection delivered by the safety injection system.
The analysis of a
main steam pipe rupture is performed to demonstrate that the following criteria are satisfied 1.
Assuming a stuck rod cluster control assembly, with or without offsite power, and assuming a single failure in I
9 pAGE 14 the engineered safeguards, there is no consequential damage to the primary system, and the core remains in place and intact.
2.
Energy release to containment from the worst steam pipe break does not cause failure of the containment structure.
I Although departure from nucleate boiling and possible clad perforation following a
steam pipe rupture are not necessarily unacceptable, the following analysis shows that no departure from nucleate boiling occurs for any rupture, assuming the most reactive assembly stuck in its fully withdrawn position.
The following functions provide the necessary protection against a steam pipe rupture:
1.
Safety injection system actuation from any of the followings a.
Two out of three low pressurirer pressure signals.
b.
High differential pressure signals between steam lines.
c.
High steam-line flow in two main steam lines (one out of two per line) in coincidence with either lou-lou l
reactor coolant system average temperature (two out of three) or lou steam-line pressure in any two lines.
d.
Two out of three high containment pressure.
2.
The overpower reactor trips (neutron flux and delta T) and lI
pAGE 15 the reactor trip occurring in conjunction with receipt of the safety injection signal.
3.
Redundant isolation of the main feedwater lines.
Sustained high feedwater flow would cause additional cooldown.
Therefore, in addition to the normal control action that will close the main feedwater valves, a safety injection signal will rapidly close all feedwater control valves, trip the main feedwater Pumps, and close the feeduater pump discharge valves.
4.
Trip of the fast-acting main steam trip valves (designed
~
to close in less than 5 sec after receipt of the signal) on a.
High steam flow in two main steam lines (one out of two per line) in coincidence with either lou-low reactor coolant system average temperature or low steam-line pressure in any two lines.
b.
Two out of three intermediate high-high containment pressure.
Each steam line has a fast-acting trip valve with a downstream nonreturn valve.
These valves prevent blowdown of more than one steam generator for any break location even if one valve fails to close. For
- example, in the case of a break upstream of the trip valve in one line, closure of either the nonreturn valve in that line or the trip valves in the other lines will prevent blowdown of the other steam generators.
Steam flow is measured by monitoring pressure difference between Pressure taps in the steam drum and downstream of the steam-line flow
~
s PAGE 16 restrictor nozzles.
These
- nozzles, which are of considerably smaller diameter than the main steam pipe, are located in the steam lines inside the containment near the steam generators to limit the maximum steam flow for any break further downstream.
15.4.2.1.2 Analysis of Effects and Consequences 15.4.2.1.2.1 Method of Analysis.
The analysis of the steam pipe rupture has been performed to determine 1.
The core heat flux and reactor coolant system temperature and pressure resulting from the cooldown following the steam-line break. The RETRAM code has been used.
2.
The thermal and hydraulic behavior of the core following a steam-line break.
A detailed thermal and hydraulic digital-computer calculation, COBRA has been used to determine if departure from nucleate boiling occurs for the core conditons computed in 1 above.
The following conditions were assumed to exist at the time of a main-steam-line-break accident:
1.
End-of-life shutdown margin at no-load, equilibrium xenon conditions, and the most reactive assembly stuck in its fully withdrawn position.
Operation of the control rod banks during core burnup is restricted so that addition of positive reactivity in a steam-line-break accident will not lead to a more adverse condition than the case analyzed.
I I ]
- o PAGE 17 2.
The negative moderator coefficient corresponding to the end-of-life rodded core with the most reactive rod in the fully withdrawn position.
The variation of the coefficient with a
temperature and pressure has been included.
The k-effective vs temperature at 1000 psi corresponding to the negative moderator temperature coefficient used plus the Doppler temperature effect, is shown in Figure 15.2-57.
The effect of power generation in the core on overall reactivity is shown in rigure 15.4-24.
The core properties associated with the sector nearest the affected steam generator and those associated with the remaining sector were conservatively combined to obtain average core properties for reactivity feedback calculations.
This causes underprediction of the reactivity feedback in the high-power region near the stuck rod.
To verify the conservatism of this method, the reactivity as well as the power distribution was checked for the statepoints shown on Table 15.4-8.
These core analyses considered the Doppler reactivity from the high fuel temperature near the stuck rod cluster control assembly, moderator feedback from the high enthalpy water near the stuck i
rod cluster control assembly, power redistribution, and non-uniform core inlet temperature effects.
For cases in which
~
steam generation occurs in the high-flux regions of the core, the effect of void formation was also included.
It was determined that the reactivity used in the kinetics analysis t
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PAGE 18 was always larger than the true value for all statapoints in Table 15.4-8, verifying conservatism; i.e.,
underprediction of negative reactivity feedback from power generation.
3.
Minimum capability for injection of high-concentration boric acid (12950 ppm) solution corresponding to the most restrictive single failure in the safety injection system.
This corresponds to the flow delivered by one charging pump delivering its full flow to the cold-leg header.
No credit has been taken for the low-concentration boric acid that must he swept from the safety injection lines downstream of the
~
boron injection-tank isolation valve prior to the delivery of.high-concentration boric acid to the reactor coolant loops.
The assumed single failure for the steam-line-break analysis is the failure of one safeguards train to function, thus resulting in the maximum delay time for boron to reach the core.
Other failures that could affect the severity of the transient are as'follows:
1.
Main' Steam trip valve.
l 2.
Feedwater control valve.
3.-
Main steam safety valve, atmospheric dump valve, or steam dump valve.
The failure of any main steam trip valve would result in no more than.one steam generator blowing down after line isolation and l
n-o PAGE 19 would not affect the severity of the transient.
i There is a backup feedline isolation valve in series with the feedwater control valve.
The failure of either of these valves would not affect the severity of the transient.
The failure of a main steam safety valve, atmospheric dump valve, or main steam dump valve would result in a small increase in steam flow that would be compensated for by full operation of the safety injection system, greatly reducing the delay of boron reaching the core.
4.
Four combinations of break sizes and initial plant conditions have been considered in determining the core power and reactor coolant system transients:
a.
Complete severance of a pipe outside the containment,
' downstream of the steam flow measuring nozzle, with the plant initially at no-load conditions, full reactor coolant flow with offsite power available, b.
Complete severance of a pipe inside the containment at the outlet of the staan-generator with the plant initially at no-load conditions with offsite power available, c.
Case (a) above, with loss of offsite power simultaneous
~
with the initiation of the safety injection signal.
Loss-of offsite power results in coolant pump coastdown.
l, d.
Case (b) above, with the loss of offsite power simultaneous u'w
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5.
Power peaking factors corresponding to one stuck rod cluster control assembly and nonuniform core inlet coolant temperatures are determined at end of core life.
The coldest core inlet temperatures are assumed to occur in the sector with the stuck rod.
The power peaking factors account for the effect of the local void in the region of the stuck control assembly during the return-to-power phase following the steam-line break.
This void, in conjunction with the large negative moderator coefficient, partially offsets the effect of the stuck assembly.
The power peaking factors depend upon the core power, temperature, pressure, and flow, and thus are different for each case studied.
A conservative thermal design flow rate was assumed for the steam-line-break analysis.
This flow rate is lower than either the mechanical design flow rate or the measured flow rate.
Using a high core flow rate may result in slightly higher peak heat fluxes but would also increase the minimum DNBR.
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PAGE 21 The values used for three of the four steam-line-break accidents analyzed are given in Table 15.4-8.
The three cases are selected on the basis of hot-channel factors, core power, and reactor coolant pressure.
The fourth case is less severe relative to DNBR.
The core parameters used for each of the three cases correspond to values determined from the respective transient analysis.
Five statepoints have been chosen for each case.
All the cases above assume initial hot shutdown conditions at time zero, since this represents the most pessimistic initial condition.
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 overpower protection system when power level reaches a trip point.
Following a trip at power, the reactor coolant system contains more stored energy than at no-load, the average 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 cooldown caused by the steam-line break before the no-load conditions of reactor coolant system temperature and shutdown margin assumed in the analyses are reached.
After the additional stored energy has been removed, the cooldown and reactivity insertions proceed
~
in the same manner as in the analysis that assumes no-load conditions at time zero.
However, since the initial steam-generator water inventory is I
PAGE 22 greatest at no-load, the magnitude and duration of the reactor coolant system cooldown are less for steam-line breaks occurring
.at power.
6.
In computing the s team flow during a steam-line break, the Moody curve for fL/D =0 is used.
'7.
perfect ~ moisture separation in the steam generator is assumed.
The assumption leads to conservative results since, realisti-cally, _ considerable water would be discharged.
Water carryover would reduce the magnitude of the RCS energy removal and system cooldown.
p 15.4.2.1.2.2 Results.
The results presented are a
conservative indication-of the events that would occur assuming a' steam-line' rupture, cince it is postulated'that all of the conditions described above occur simultaneously.
Cone power and Reactor Coolant System Transient i
Figure 15.4-25 shows the reactor coolant system transient and core heat flux following a main steam pipe rupture (complete severance of a pipe)
I outside the containment, downstream of the flow-measuring nozzle at initial no-load conditions (case a). The break assumed is the largest i,
break that can occur anywhere outside the containment, either upstream or downstream of the trip valves. Offsite power is assumed available so
~that full reactor coolant flow exists. The transient shown assumes a Ji steam release from.only one steam generator. Should the core be critical
~
PAGE 23 at near mere power when the rupture occurs, the initiation of safety injection by high differential pressure between any steam line and the remaining steam lines, or by high steam, flow signals in coincidence with either low-low reactor coolant system temperature or low steam-line
- pressure, will trip the reactor. Steam release from more than one steam generator will be prevented by automatic trip of the fast-action trip valves in the steam lines by the high steam flow signals in coincidence with either low reactor coolant system temperature or low steam-line pressure.
Even with the failure of one valve, release is limited to no more than 10 sec for the other steam generators while the one steam generator blows down.
The steam-line trip valves are designed to be fully closed in less than 5 sec after receipt of closure signal with no flow through them.
With the high flow existing during a steam-line rupture, the valves will close considerably faster.
The steam flow in Figures 15.4-25 through 15.4-28 represents steam flow from the faulted steam generator only. In addition, all steam generators were assumed to discharge through the break for the first 10 sec.
As shown in Figure 15.4-25, the core attains criticality with the rod cluster control assemblies inserted (with the design shutdown assuming sone stuck assembly) before boron solution at 12,950 ppm enters the reactor coolant system from the safety injection system. The delay time consists of the time to receive and actuate the safety injection signal
~
and the time to completely open valve trains in the safety injection lines.
The safety injection pumps are then ready to deliver flow. At this stage a
further delay time is incurred before 12,950-ppm boron l
I
PAGE 24 solution can be injected to the reactor coolant
- system, due to low-concentration solution being swept from the safety injection lines.
A peak core power well below the nominal full-power value is attained.
The calculation assumes that the boric acid is mixed with, and diluted by, the water flowing in the reactor coolant system prior to entering the reactor core.
The concentration after mixing depends upon the relative flow rates in the reactor coolant system and in the safety injection system. The variation of mass flow rate in the reactor coolant system due to water density changes is included in the calculation, as is the variation of flow rate from the safety injection system due to changes in the reactor coolant system pressure. The safety injection system flow calculation includes the line losses in the system as well as the pump head curve. The accumulators would provide an additional source of borated water when the reactor coolant system pressure decreases to belou 550 psia. The integrated flow rate of borated water from the safety injection system for each of the four cases analyzed is shown in Figure 15.4-29.
~
rigure 15.4-26 shows case b, a steam-line rupture at the exit of a steam generator at no-load.
The sequence of events is similar to that described above for the rupture outside the containment, except that criticality is attained earlier due to more rapid cooldown, and a higher peak core average power is attained.
Figures 15.4-27 and 15.4-28 show the response of the salient parameters for cases c
and d, which correspond to the cases discussed above with additional loss of offsite power at the time the safety injection signal
PAGE 25 is generated. The safety injection system delay time includes 10 see to otart the diesel and.10 sec for the safety injection pump to reach full speed.
In each case, criticality is achieved later and the core power increase is slower than in the similar case with offsite power available.
The ability of the emptying steam generator to extract heat from the reactor coolant system is reduced by the decreased flow in the reactor coolant system.
For both these
- cases, the peak core power remains well below nominal full power value.
It should be noted that.-following a steam-line break, only one steam generator bleus down completely. Thus the remaining steam generators are ctill available for dissipation of decay heat after the initial transient is over.
In the case of loss of offsite power, this heat is removed to the atmosphere via the main steam safety valves, which have been sized to cover this condition.
i A
steam-line break assuming an isolated loop is less severe than the case.analyced above. Although operation with an isolated loop results in a
reduced primary coolant inventory, this condition is offset by the increased. shutdown margin available due to the reduced power defect.
I The sequence of events is shown in Table 15.4-9.
l The steam-line break analysis adequately addresses the NRC's concerns i
oxpressed in IE Bulletin 80-04.
i-I l
(!.
PAGE 26 Margin to critical Heat Flux Using the transients shown in Figur'es 15.4-25 through 15.4-28 the Westinghouse W-3 correlation was used in conjunction with the Vepco version of the COBRA core thermal hydraulics code to determine the cargin to DNB.
Carefully chosen points from each transient were oxamined and the results showed that all cases had a minimum DNBR greater than 1.30.
The power and flow conditions are shown together with pressure and core inlet temperatures in Table 15.4-8 for the three
-cases most critical to departure from nucleate boiling.
[
i l
T PAGE 27 TABLE 15.4-8 CORE PARAMETERS USED IN STEAM BREAK DNB ANALYSIS Case a Time Points Parameter 1
2 3
4 5
Reactor vessel inlet tempera-ture to sector nearest affected steam generator,
'T 418 411 407 388 365 Reactor vessel inlet tempera-
-ture to remain-ing sector.
'T 499 495 491 461 428 RCS pressure, 1447.
1422.9 1401.0 1279.2 1194.1 psia RCS, flow %
100 100 100 100 100 Heat flux. %
13.02 13.08 12.80 9.59 9.42 Time, sec 46 58 70.5 156 267 W
I I
PAGE 28 TABLE 15.4-8 (continued)
CORE PARAMETERS USED IN STEAM BREAK DNB ANALYSIS Case b Reactor assel inlet tempera-ture to sector nearest affected steam generator,
'T 372 363 359 357 355 Reactor vessel inlet tempera-ture to remain-ing sector,
'T 518 513 508 505 500 RCS pressure, 1345.7 1162.8 998.8 941.9 909.7 psia RCS flow, X 100 100 100 100 100 Heat flux, X 23.38 23.72 22.92 22.32 21.~80 Time, sec 33.75 45.25 56.75 61.5 70.0 l
PAGE 29 TABLE 15.4-8 (continued)
CORE PARAMETERS USED IN STEAM BREAK DMB ANALYSIS Casa d Parameter 1
2 3
4 5
Reactor vessel inlet tempera-ture to sector nearest affected steam generator,
'T 311 244 237 232 226 Reactor vessel inlet tempera-ture to remain-ing sector,
'T 523 514 512 510 508 RCS pressure, Psia 1463.9 1393.5 1399.0 1403.2 1407.6 RCS flow, X 18.90 8.94 8.26 7.74 6.99 Heat flux, X 9.456 6.406 6.128 5.691 5.042 Time, sec 50 135 152 168 195
=
. l
PAGE 30 TABLE 15.4-9 TIME SEQUENCE OF EVENTS FOR MAJOR SECONDARY SYSTEM FILE RUPTURE Accident Event Time (sec)
Major secondary system Pipe rupture 1.
Case a Steam-line ruptures 0
Criticality attained 13 Pressurizer empty 10 12250 ppm boron reaches loops 32 2.
Case b Steam-line ruptures 0
Criticality attained 11 Pressurizer empty 10 12250 ppm boron reaches loops 32 3.
Case c Steam-line ruptures O
Criticality attained 18 Pressuri=er empty 11 12250 ppm boron reaches loops 41 4.
Case d Steam-line ruptures 0
Criticality attained 16 Pressurizer empty 11 12250 ppm boron reaches loops 41 l
' LAPS UFSAR Figure 15.4-2 PAGE 31
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NAPS UFSAR Figura 15.4-28 PAGE 35
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NAPS UFSAP Fioure 15.4-29 i
PAGE 36 Case a 15 l
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PAGE 37 B.3 Accidental Main Steam System Depressurization Updated UFSAR Accident Analysis Section 15.2.13 ACCIDENTAL DEPRESSURIZATION OF THE MAIN STEAM SYSTEM 15.2.13.1 Identification of Causes and Accident Description The most severe core conditions resulting from an accidental depressuri=ation of the main steam syster are associated with an inadvertent opening of a single steam dump, relief, or safety valve. The analyses performed assuming a rupture of a main steam pipe are given in Section 15.4.
The steam release as a consequence of this accident results in an initial increase in steam flow that decreases as the steam pressure falls.
The energy removal from the reactor coolant system causes a reduction of coolant temperature and pressure.
In the presence of a negative moderator temperature coefficient, the cooldown results in a reduction of core shutdoun margin.
The current analysis is performed to demonstrate that the following criterion is satisfied assuming a stuck rod cluster control assembly and a single failure in the engineered safety features, there will be no departure from nucleate boiling in the core for a
steam release equivalent to the spurious
- opening, with failure to close, of the largest of any single steam dump, relief, or safety valve.
The following systems provide the necessary protection against an accidental depressuri=ation of the main steam systems l
I PAGE 38 1
1.
Safety injection system actuation from any of the followings a.
Two out of three low-low pressuri=er pressure signals.
b.
High differential pressure signals between steam lines.
2.
The overpower reactor trips (neutron flux and delta T) and the reactor trip occurring in conjunction with receipt of the safety injection signal.
3.
Redundant isolation of the main feeduater lines - Sustained high feeduater flow would cause additional cooldown.
Therefore, in addition to the normal control action that will close the main feedwater valves following reactor trip, a safety injection signal will rapidly close all feeduater control valves, trip the main feedwater pumps, and close the feedwater pump discharge valves.
15.2.13.2 Analysis of Effects and Consequences 15.2.13.2.1 Method of Analysis The following analyses of a secondary system steam release are performed for this section8 1.
The core heat flux and reactor coolant system temperature and pressure resulting from the cooldown following the inadvertent opening of the largest capacity valve listed above.
The RETRAM code has been used.
~
l
e.
PAGE 39 2.
The thermal and hydraulic behavior of the core during this event.
A detailed thermal and hydraulic digital-computer calculation, COBRA has been used to determine if departure from nucleate boiling occurs for the core conditions computed
.in 1 above.
The following conditions are assumed to exist at the time of a secondary-system-break accident:
1.
End-of-life shutdown margin at no-load equilibrium xenon conditions, and with the most reactive rod cluster control assembly stuck in its fully withdrawn position.
Operation
~
of rod cluster control assembly banks during core burnup is restricted so that addition of positive reactivity in a secondary-system-break accident will not lead to a more adverse condition than the case analyzed.
2.
A negative moderator coefficient corresponding to the end-of-life rodded core with the most reactive rod cluster control assembly in the fully withdrawn position.
The variation of the coefficient with temperature and pressure is included.
The Kaff versus temperature at t
1000 psi corresponding to the negative moderator temperature coefficient used plus the Doppler tempera-ture effect is shown in Figure 15.2-57.
3.
Minimum capability for injection of high-concentration boric acid solution corresponding to the most restric-I
a.
PAGE 40 tive single failure in the safety injection system.
This corresponds to the flow delivered by one charging pump delivering its full. contents to the cold-leg header.
No credit has been taken for the low-concen-tration boric acid (2000 ppm) that must be swept from the safety injection lines downstream of the boron injection tank isolation valves prior to the delivery of high-concentration boric acid (12.950 ppm) to the to the reactor coolant loops.
4.
The case studied is an initial total steam flow of 262 lb/sec at 1020 psia from all steam generators, with offsite power available.
This is the maximum capacity of any single steam dump or safety valve.
Initial hot shutdown conditions at time zero are assumed, since this represents the most pessimistic initial condition.
Should the reactor be just critical or operating at power at the time of a steam release, the reactor will be tripped by the normal overpower protection when power level reaches a trip point.
Following a trip at power, the reactor coolant system contains more stored energy than at no-load, the average 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
PAGE 41 cooldown caused by the steam-line break before the no-load conditions of reactor coolant system temperature and shutdown margin assumed in the analyses are reached.
After the additional stored energy has been removed, the cooldown and reactivity insertions proceed in the same manner as the analysis, which assumes no-load condition at time zero.
However, since the initial steam generator water inventory is greatest at no-load, the magnitude and duration of the reactor coolant system cooldoun are less for steam-line breaks occurring at power.
5.
In computing the steam flow, the Moody Curve for fl/D
= 0 is used.
6.
perfect mositure separation in the steam generator is assumed.
7.
In the original analysis of the steam-line break incident, which is a depressuri=ation transient, credit was taken for coincident low pressuri=er pressure and level for safety injection actuation following a credible break (accidental depressurization of the main steam system).
Since that analysis was performed, the low-level coincidence requirement has been removed from the plant protection circuitry.
Thus, safety injection actuation can occur on a low Pressuri=er pressure.
I
PAGE 42 15.2.13.2.2 Results The results presented are a conservative indication of the events that would occur assuming a secondary system steam release, since it is postulated that all of the conditions described above occur simultaneously.
Figure 15.2-59 shows the transients arising as the result of a steam release with an initial steam flow of 262 lb/sec at 1020 psia with steam release from one condensor dump valve. The assumed steam release is typical of the capacity of any single steam dump or safety valve. In this case safety injection is initiated automatically by low pressurizer Pressure.
Operation of one centrifugal charging pump is considered.
Boron solution at 12,950 ppm enters the reactor coolant
- system, providing sufficient negative reactivity to limit the return to power to a
level below 4% of the rated nominal power.
With the reactor coolant pumps still providing full flow, the minimum departure from nucleate boiling ratio is well above the limit for Condition II acceptance criteria.
The reactivity transient for the case shown in Figure 15.2-59 is more severe than that of a faulted steam-generator safety or relief
- valve, which is terminated by steam-line differential pressure. The transient is quite conservative with respect to cooldoun, since no credit is taken for the energy stored in the system metal other than that of the fuel elements.
Since the transient occurs over a period of about 5 min, the neglected stored energy is likely to have a significant effect in slowing the cooldoun.
~
l
PAGE 43 TABLE 15.2-1 (continued)
TIME SEQUENCE OF EVENTS FOR CONDITION II EVENTS Accident Event Time (sec)
Accident depressuri-Inadvertent opening of one 0
=ation of the reactor RCS safety valve coolant system Reactor trip 21.6 Minimum DNBR occurs 24.0 Accidental Inadvertent opening of one 0
depressuri=ation of steam safety or relief valve the main steam system pressuri=er empties 106 12950 ppm boron reaches core 233 Decay heat release line break Hot =ero-power break Break occurs 0.0 pressuri=er empties 150 Safety injection starts 172 20,000 ppm reaches core 215 Minimum shutdown margin 215 reached
. I
PAGE 44 TABLE 15.2-1 (continued)
TIME SEQUENCE OF EVENTS FOR CONDITION II EVENTS Accident Event Time (sec)
Hot full-power break Low-low steam-genertor 0.0 level trip Auxiliary feedwater starts 60 pressurizar empties 564 Safety injection starts 566 20,000 ppm reaches core 613 Minimum shutdown margin 617 reached Inadvertent operation Charging pumps begin inject-0 of ECCS during power ing borated water operation.
Lou pressure trip point 64 reached Rods begin to drop 66 I
ygg gym yigure 15.2-59 e,
a
- PAGE 45
.J 2500
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TRANSIENT RESPONSE FOR A STEAM LINE BREAK EQUlVALENT TO 252 LB/SEC AT 1020 PSIA WITH OUTSIDE POWER AVAILABLE
n pAGE 46 C.
OPERATIONS /FSAR REVIEW 1.
Operations Impact An evaluation of effects on plant operations was made to determine all positive or negative implications of reducing boron concentrations in the concentrated Boric Acid System and in the Boron-Injection Tank.
A summary of those impacts is provided belous a.
Increase in time to borate under normal and emergency operating conditions-station curve Book nomographs for boron addition will be revised for the decreased minimum concen-tration of 7.4N.
All increases in times re-quired to borate or makeup were found to be satisfactory from a plant operational standpoint.
b.
Increase in minimum volume of boric acid.
The, increase in the minimum volume of boric acid stored for each unit to 6000 gallons (previously 4450) was evaluated relevant to overflow considerations.
It was determined that sufficient tank capacity is available to replenish the tank in standard batch volumes without overflowing the tank.
lI
.o PAGE 47 c.
The boron evaporators will be operated to produce a nominal boric acid concentrate of 8% if reuse is desired or 12% if disposal (via solid waste) is desired.
d.
Setpoint, chemistry, and operating procedure changes have been identified and will be re-vised in accordance with approved procedures.
2.
UFSAR Revieu The Updated Final Safety Analysis Report for North Anna has been reviewed and the required changes, other than the changes in the accident analysis section previously addressed, have been identified.
Upon approval of this submittal, these changes will be submitted with the normal yearly UFSAR update.
D.
CONCLUSIONS The reduction in boric acid concentration for the Boron Injection Tank and the Concentrated Boric Acid System offers significant benefits to Virginia Electric and power Company.
These benefits include increased operational reliability, reduced maintenance costs and decreased personnel radiation exposure.
Additionally, a detailed operational revieu was conducted it has been concluded that the plant can continue to be operated in a safe and efficient manner following the change.
.I
PAGE 48 Analyses of the Main Steam pipe Rupture and Accidental Depressurization of the Main Steam System incorporating the proposed changes.have been performed to demonstrate that these transients meet the Condition II transient acceptance criteria.
As such, it can be concluded that the change in boron concentration will not cause any safety limits to be exceeded for any incident and consequently no unreviewed safety questions as defined in 10CFR50.59 exist as a
result of these proposed changes.
The results of this evaluation can be stated as follows.
i 1.
No increase in the probability of occurrence or consequences of an accident will result from these proposed changes.
The systems will undergo no physical changes for the reduction in boron concentration and therefore no change in the associated probabilities is expected.
2.
Since the proposed changes cause no other system changes (e.g.,
alterations in plant configuration), and given that the effects upon system accident response are fully described by the parameters evaluated, operation with these proposed changes does not create the possibility of an accident of different type than any evaluated previously in the safety Analysis Report.
3.
The margin of safety as defined in the basis for the Technical Specifications is not reduced.
The calculated safety parameters for the affected transients are all well within the allowable limits for the acceptance criteria for condition II transients.
It has also been determined that the reduction of the boron concentration in the boron injection tank and the boric acid tanks does not pose a significant hazards consideration.
This is based on Example vi of those examples of amendments that are considered not likely to involve significant hazards considerations.
Example vi partially states "A change which either may result in some increase l
PAGE 49 to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all, acceptable criteria with respect to the system or component specified in the Standard Review Plan."
The analyses do show that the reduction of the boron concentration in the boron injection tank and the boric acid tanks allow a slight increase in the accident consequences (i.e.,
a small return-to-power for the accidental depressurization of the main steam system).
The results of these analyses clearly show that all of the acceptance criteria for these types of transients are met as shown in chapter 15 of the Morth Anna UFSAR and the Standard Review Plan.
J