ML20195J157
| ML20195J157 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 06/09/1988 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20195J141 | List: |
| References | |
| NUDOCS 8806290064 | |
| Download: ML20195J157 (26) | |
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l ATTACHMENT 1 ,
Proposed Technical Specification License Nos. DPR-39 and DPR-48 Appendix A, Sections 3.2, 3.4, 3.8 and 3.9 Pages Modified 66 127 138 139 145 185 194 207b Pages Added 127a Pages Deleted 166-Leave Intentionally Blank 4
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la order to ensure solution solubility at the boron concentration is measured and the predicted boric acid concentration in the system, a minimum curve is adjusted to this point. As power temperature of 145'F is required. The operation operation proceeds, the measured boron ,
of one boric acid transfer pump provides a backup concentration is compared with the predicted I supply (besides RWST) of boric acid to the concentration and the slope of the curve relating charging pump suction. One of the two channels of burn-up and reactivity is compared with that heat tracing is normally operating and is predicted.
sufficient to maintain the line temperature of 145'F. The second channel of heat tracing This process or normalization shall be completed provides backup for continuous plant operation early in a core life. Thereaf ter, actual boron when one channel is inoperable. Shouta both concentration can be compared with prediction.
channels of heat tracing become inoperative, a Any reactivity anomaly greater than 1% would be reactor can be easily borated before acid unexpected, and its occurrence would be precipitative temperature. thoroughly investigated and evaluated. The calue of 1% is considered a safe limit since a shutdown When the boron concentration of a Reactor Coolant margin of at least 1% with the most reactive rod System is to be changed, the process must be in the fully withdrawn position is always uniform to prevent sudden reactivity changes in a maintained.
reactor. Mixing of the reactor coolant will be sufficient to maintain a uniform boron Design criteria have been chosen for Condition I concentration if at least one reactor coolant pump and II events which are consistent with the fuel or one residual heat removal loop is operating integrity analyses. These relate tu fission gas while the change is taking place. A residual heat release, pellet temperature and c: adding removal pump will circulate the equivalent of the mechanical properties. Also the minimum DNBR in reactor coolant system volume in approximately one the core must not be less than the applicable half hour. design limit DNBR in normal operation or in short term transients.
To eliminate possible errors in the calculations of the initial reactivity of a core and the In addition to conditions imposed for Condition I reactivity depletion rate, the predicted relation and 11 events, the peak linear power density must between fuel burn-up and the boron concentration not exceed the limiting Kw/ft values which result recessary to maintain adequate control from the large break loss of coolant accident characteristics, must be adjusted (normalized) to analysis based on the ECCS acceptance criteria accurately reflect actual core conditions. When limit of 2200*F.
full power is reached initially, and with the control rod groups in the desired positions, the This is required to meet the initial conditions assumed for loss of coolant accident. To aid in Amendment Nos. 83 Unit 1 specifying the limits on power distribution the 73 Unit 2 following hot channel factors are defined.
12330/12340 66 e
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e_ l LIMITING CON 01T10N FOR OPERAT10N SURVEILLANCE REQUIREMENT 3.4 SAFEGUARDS INSTRUMENTATION AND CONTROL 4.4 SAFEGUARDS INSTRONENTATION AND CONTROL Applicability: Applicability:
Applies to safeguards instrumentation and Applies to the testing and calibration of safeguards control channels per unit. Instrumentation and control channels per unit.
Objective:
I Objective: t' To establish the limiting conditions of To establish the testing and surveillance ,
operation for saf eguards instrumentation and requirements for safeguards instrumentation ind i controls. control channels.
Specification: Specification:
- 1. The setpoints for the engineered safeguards 1. Not Applicable, systems are presented in Table 3.4-1. Table 2. The instrument channel check, instrument and 3.4-2 designates the primary sensors for the control channel functional check and instrument safeguard channels. channel calibration frequency requirements for the various safeguards instrumentation and ,
For on line testing or instrumentation control channels are specified in Table 4.4-1.
2.
failure unit operation shall be permitted to The safeguards equipment shall either be actuated continue as follows: or continuity checked to the final actuating device quarterly (See Foctnotes 1 & 2) as specified in Table 4.4-2. Performance of any
- a. In accordance with Table 3.4-1 surveillance test in this specification is not
- b. Only one channel of a particular required if the unit is in Mode 5 or 6 provided protection set shall be tested at a time. that prior to exceeding Mode 5 the specified tests have been performed.
- c. Failed channels or channels being tested shall be placed in the tripped mode with Valves with an
- or ** in Table 4.4-2 shall the exception of the containment Hi-H1 be placed in the required position indicated and Pressure channels. shall have the power removed at their motor control center. Each month the breakers for these valves shall be verified off. Power may be restored for plant testing, plant evolutions requiring operation of these valves, or in Mode 5 or 6.
Amendment No. 20 & 17 12330/12340 127
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LIMITING CONDITION FOR OPERAT10N SURVEILLANCE REQUIREMENT ,
Motor operated valves with a # in Table 4.4-2 shall' be placed in the required position indicated and shall have power removed at their motor control center. Each month the breakers for these valves shall be verified of f until removal of the Boron Injection Tank (Bli) under modifications M22-1-88-21, work package 00132, for Unit 1; and M22-2-88-21, work package 00134, for Unit 2. On -
completion of the modification, these motor operated valves ,
shall be restored to cperability and shall be subject to the ,
quarterly acutation testing requirements of Technical :
Specification 4.4.2. {*
Air operated valves with a # in Table 4.4-2 shall be i failed to the required position indicated and have their air j supply verified isolated monthly until these valves have been permanently disabled or removed. ;
i At least once per 18 months, all valves with a # in !
Table 4.4-2 shall be verified to be in their required ,
position.
Valves with a + in Table 4.4-2 will be removed, and e deleted from the Technical Specification, on completion of , ,
modification M22-1-88-21, work package 00132, for Unit 1; and M22-2-38-21, work package 00134, for Unit 2.
e 9
(1) Valves with
- in Table 4.4-2 shall be tested only during refueling outage.
(2) Valves with a # in Table 4.4-2 shall not Amendment No. 20 & 17 be actuated.
12ra '
12330/12340 i
1 Safctv tion (Cont.) Device Actuation Safety Injection Pump 1A (2A) X Safety Injection Pump 1B (28) X R:sidual Heat Removal Pump 1B (28) X MOV-RH8716B X MOV-RHB7008 X Component Cooling Pump OE X Comporent Cooling Pump 00 X Component Cooling Pump OC X Charging Pump 1A Lube 011 Pump (2A) X Charging Pump 1A (2A) X R sidual Heat Removal Pump 1A (2A) X MOV--RH8700A X l McV-SIB 803A # OPEN X M0Y-VC81i0 X l MOV-518800C
- OPEN X 1 MOV-SIB 003B # OPEN X MOV-VC8111 X ,
McV-51880BD
- OPEN X j MOV-LCV1128 X MOV-LCV1120 X MOV-VC8105 X MOV-518808A
- OPEN X
- See Section 4.4.2 l MOV-518801A + X and Basis on page 145.
MOV-SI8804A X
- See Section 4.4.2 MOV-519011B ** CLOSED X and Basis on page 145.
MOV-5189238 X MOV-518806 *0 PEN X . # See Section 4.4.2 MOV-518802 *0 PEN X and Basis op page 145.
MOV-SIB 8000 X MOV-SI8923A X + Valves with this MOV-518800A X designation will be MOV-SI9010A X deleted from the A0V-518880 X Technical Specification MOV-RH8716C X on completion of MOV-RH9000 ** CLOSED X Modification P.22-1-88-21 MOV-RH8703 *0 PEN X work package 00132, for MOV-RH8716A X Unit 1; and M22-2-88-21 Component Cooling Pump OA X work package 00134, for Component Cooling Pump 08 X Unit 2.
Engineered Safeguards Equipment Actuation Test Amendment No. 20 & 17 TABLE 4.4-2 (Sheet 3 of 8) 138 12330/12340
A Contiruity Check At Final I' Safety ., fction (Cont.) Device Actuation ) Actuating Devico ,
l A0V-SIB 870A+, # CLOSED X l A0V-518883+, i CLOSED X '
MOV-LCV112C X MOV-LCV112E X MOV-VC8106 X MOV-5188088
- OPEN X MOV-519010B X ,
MOV-SIB 807B X MOV-SI8804B X MOV-518809A **0 PEN X
- See Section 4.4.2 X and Basis on page 145.
l MOV-518801B +
I ADV-518870B+, # CL) SED X MOV-518800B X ** See Section 4.4.2 MOV-5188128
- OPEN X and Basis on page 145.
MOV-SI8812A
- OPEN X
- See Section 4.4.2 MOV-519011A ** CLOSED X [
MOV-SIB 807A X and Basis on page 145.
MOV-SI8809B ** OPEN X MOV-SIB 000C X + Valves with this ;
designation will be +
F) Emergency Fan Coolers deleted from the ,
Technical Specification Component on completion of modification M22-1-88-21 Vent Fan 1A low speed (2A) X work package 00132, for Vent Fan IB low speed (28) X Unit 1; and M22-2-88-21 Vent Fan 1C low speed (2C) X work package 00134, for Vent Fan ID low speed (2D) X Unit 2.
Vent Fan IE low speed (2E) X i G) Service Water Pump Starts and System Isolation j .
! Component Service Water Pump 1A (2A) X OA0V-SWOO20 - DA0V-SWOO21 X Service Water Pump 1B (28) X MOV-SW0001 X MOV-5WO100 X i OMOV-SW0006 X Amendment No. 20 + 17 Engineered Safeguards Equipment Actuation Test TABLE 4.4-2 (Sheet 4 of 8) 139 12330/12340
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Bases:
- Valves in Table 4.4-2 which have been designated as required to be stroked during refueling outage: These 4.4 The bases for Tables 3.4-1 and 4.4-1 is an valves are placed in the proper positions for ECCS operatlJn analysis of the engineered safeguards and have their power removed. Removal of power has been instrumentation and control system. found to be acceptable to satisfy spurious valve actuation Conservative failure rates for the criteria until analysis determines a final acceptable individual channels were employed based on solution. These valves do not change position throughout published data typical for individual LOCA. Stroking at a refueling interval will insure their components in the channels. The test and operabi li ty. Energization of these valves is permissible to calibration frequencies are therefore quite support other testing or plant evolutions such as beatup or ,
conservative. Actuation of the safeguards cooldown. The valves in Table 4.4-2 marked with ** also are equipment quarterly using the actuating placed in proper position with power removed as above.
relay (s) insures continued system lhese valves are required to be shifted when going from cold
- operabi li ty. to hot leg injection. Quarterly testing is kept to insure i operability of these valves. These valves may also be energized to support plant testing or operations. In Mode 5 I or 6, power may be returned to all valves under spurious (1) FSAR Section 7.5 valve actuation criteria. The valves in Table 4.4-2 marked j with # are also placed in proper position for ECCS operation ;
- with power removed as above. In the case of Air Operated Valves, these valves will be failed to the proper position for ECCS operation with their air supply isolated. Valves marked with # are associated with the Boron Injection Tank :
(BIT), which is no longer required and may be removed and
- replaced with piping.
On completion of modifications M22-1-88-21, work package -
00132, for Unit 1; and M22-2-88-21, work package 00134, for i Unit 2; Motor Operated Valves marked with # will be restored ,
to operability. Concurrently, valves marked with (+) will be removed and replaced with piping. Air operated valves marked with # and specified as closed may be disabled either pneumatically or mechanically, or they may be removed and their associated lines capped.
Amendment 20 & 17 12330/12340 145
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Component Name Component Number i Centrifugal Charging Pump-1A (2A) VC006-1A (2A) ~
Centrifugal Charging Pump-1B (2B) VC006-1A (28)
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Charging Header Isolation MOV-VC8105 Valves MOV-VC8106 R:: circulation Flow Isolation MOV-VC8110 Valves MOV-VC8111 Volume Control Tank Outlet MOV-VC112B Isolation Valves MOV-VC112C Emergency Suction Valves MOV-VC112D From RWST MOV-VC112E MOV-518801A
- On completion of modification M22-1-88-21, work package 00132 for Unit 1; and M22-2-88-21, work package 00134, for Unit 2; these valves shall be replaced in this specification d by valves 1(2) MOV-SI8803A and 1(2) MOV-518803B.
i i
Centrifugal Charging Pump System TABLE 4.8-1 185 12330/12340
v .s b1 Bases 3.8 and 4.8 (Continued)
The availability of the systems is demonstrated by The pressure and volume limits for the accumulators inanediately demonstrating the operability of the assure that the required amount of water is cumponents redundant to the failed one, as well as injected with the required boric acid concentration .
the operability of the inter-related systems and following a loss-of-coolant accident. The Ilmits the standby AC and DC power supplies that feed are based on the values used for the accident them. The continued availability of these analyses. (4) coup,snents during the repair period is demonstrated !
by repeating these tests daily. The five component cooling system pumps and three heat exchangers are located in the Auxiliary Assuming a reactor has been operating at full rated Building and are a shared system between Units I power fer at least 100 days, the magnitude of the and 2. The components are accessible for repair decay heat decreases after initiating MODE 3. af ter a loss-of-coolant accideat. During the Thus, the requirement for core cooling in case of a recirculation phase following a loss-of-coolant postulated loss-of-coolant accident while in the accident on a unit, only one componOnt cooling pump ODE 3 condition is significantly reduced below the and heat exchanger is required for minimum requirements for a postelated loss-of-coolant safeguards of that unit. Therefore, a minimum accident during power operation. Putting a reactor requirement of four component cooling pumps and in the MODE 3 condition significantly reduces the three heat exchangers for two operating units patential consequences of a loss-of-coolant provides sufficient redundancy. (5) accident, and also allows more free access to some of the engineered safeguards components in order to A totil of six service water pumps are installed; effect repairs. only one service water pump is required tausediately following a postulated loss-of-coolant accident.
Failure to complete repairs within the allowable (6) (See p. 195) time after going to the MODE 3 condition is considered indicative of 7 requirement for major maintenance ar1 therefore in such a case, the ,
reactor in general is to be put into the MODE 5 :
condition. j The limits for the accumulators and refueling water l
storage tank insure the required amount of water with the required boron concentration is available for injection into the primary r,oolant system following a loss-of-coolant accident and are based on the values used for the accident analysis (4).
12330/12340 194 Amendments 81 and 77 I
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) -1 VALVE NUMBER FUNCTION MOV-CS0002 Containment Spray Header Isolation -
MOV-CS0004 Containment Spray Header Isolation MOV-CS0006 Containment Spray Header Isolation MOV-FWO0l6 Feedwater to Steam Generator B MOV-FWOOl7 Feedwater to Steam Generator C MOV-FW0018 Feedwater to Steam Generator A MOV-FWOO19 Feedwater to Steam Generator D MOV-FWOOSO Aux Feed to Steam Generator B MOV-FWOO51 Aux Feed to Steam Generator B MOV-FWOOS2 Aux Feed to Steam Generator C MOV-FW0053 Aux Feed to Steam Generator C MOV-FWOOS4 Aux Feed to Steam Generator A M0Y-FWOO55 Aux Feed to Steam Generator A MOV-FWOOS6 Aux Feed to Steam Generator D MOV-FWOO57 Aux Feed to Steam Generator D
+MGV-MS0005 Steam to Auxiliary Feedwater Pump
+MOV-MS0006 Steam to Auxiliary Feedwater Pump
+MOV-MS00ll Steam to Auxiliary feedwater Pump MOV-RHB701 Residual Heat Loop Outlet
+MOV-RH9000 Hot leg Safety Injection
+MOV-S18801A & B ** ECCS Charging Discharge l Cold leg Safety Injection
+MOV-SIB 802
+MOV-S18809A & B Residual Heat Removal to Loops i
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+MOV-SI90llA & B Hot leg Safety Injection I +MOV-SW0001 Service Water to Fan Coolers
+MOV-SW0002 Service Water to Fan Coolers
+MOV-SWOOO3 Service Water to Fan Coolers
+MOV-SW0004 Service Water to Fan Coolers
- Indicates that Type C Local Leak Rate Testing is required. ;
- On completion of Modifications M22-1-88-21, work package 00132 for Unit 1; and M22-2-88-21, work package 00134 for Unit 2; replace 1(2) MOV-SI8801A & B in this table with 1(2) MOV-SI8803A and B. Note that on :
completion of these modifications, valves 1(2) MOV-SI8803A and B shall be dual function valves as defined in Specification S.9.3.A.
TABLE 3.9-3d CONTAINMENT ISOLATION VALVES OTHER 207b Amendment Nos 105 & 95 12330
ATTACHMENT 2 Proposed Technical Specification License Nos. DPR-39 and DPR-48 Appendix A, Sections-3.2, 3.4, 3.8 and 3.9 Description and Summary of Proposed Changes C
ATTACHMENT 2 Summary of Proposed Changes Zion Technical Specifications, Appendix A-Section 3.2, 3.4, 3.8, 3.9 The purpose of this Technical Specification (TS) change is to remove the Boron Injection Tank (BIT) and the piping, valves and heat tracing associated with recirculation of the BIT to Boric Acid Tank (BAT).
This change will be performed in distinct phases as outlined below:
Phase 1: (Figure 1)
Upon approval of this TS license amendment, and when operationally appropriate both units will operate in the following configuration:
- 1. BIT inlet isolation valves, MOVSI 8803A and MOVSI8803B will be deenergized OPEN
- 2. BIT to BAT isolation valve, AOVSI8870A and AOVSI8870B will be failed CLOSED
- 3. BAT pump to BIT isolation valve, AOVSI8883 will be failed CLOSED
- 4. BIT will be filled with Reactor coolant at operating boron concentration and pressurized to charging pump discharge pressure
- 5. BIT discharge vales MOVSI8801A and MOVSI8801B will be CLOSED cold in order to prevent pressure binding of double disc gate valves. Also the equalizing line between the double disks will be isolated by closing SI8871A and SI8871B
- 6. This change can be performed with the plant in any operating mode
- 7. This phase affects pgs. 66, 127, 127a, 138, 139, 145, 166 and 194
Phtro 2: (Figure 2)
During subsequent refueling outages on each unit, the following plant modifications will be performed:
- 1. BIT inlet isolation valves, MOVSI8803A and MOVSI8803B will be either electrically or mechanically disabled OPEN
- 2. BIT to BAT isolation valves A0VSI8870A and MOVSI8870B will be (either electrically or mechanically disabled CLOSED) then removed.
- 3. BAT pump to BIT isolation valve A0VSI8883 will be removed
- 4. BIT to BAT recirculation valves and piping will be removed and capped
- 5. BIT will be removed and replaced with straight pipe
- 6. After BIT is removed, BIT inlet isolation valves, MOVSI8803A and MOVSI8803B will be reenergized and restored to OPERABLE status
- 7. BIT discharge valves, MOVSI8801A and MovsI8801B will be removed
- 8. MOVSI8803A and MOVSI8803B will replace MOVSI8801A and MOVSI8801B as Containment Isolation Valves
- 9. This phase affects pgs. 127a, 138, 139, 145, 185 and 207b Summary of changes-Sorted by Technic 31 Specification Page
- 1. Page 66-removes reference to BIT recirculation from the basis to Tech S; : 3.2
- 2. Page 127-format change to consolidate footnotes
- 3. Page 127a-entire page added to Tech Specs to explain the use of (#) and (4) signs in Table 4.4-2
- 4. Page 138-incorporates use of (#) and (+)
signs into Table 4.4.-2
- 5. Page 139-incorporates use of (#) and (+)
signs into Table 4.4.-2
- 6. Pag 3 145-rsvicss basis ecction 4.4 to includa use and explanation of (#) and (+) signs
- 7. Page 166-deletes entire specification 3.8.1.E and 4.8.1.E and leave page intentionally blank
- 8. Page 185-removes valves MOVSI8803A, MOVSI8803B, AOVSI8870A, AOVSI8870B, A0VSI8883 and SI 8912 froia Table 4.8-1 and add (*)
symbol
- 9. Page 194-removes reference to BIT in Basis 3.8/4.8
- 10. Page 207b-identifies MOVSI8803A and MOVSI8803B as containment isolation valves after MOVSI8801A and MOVSI8801B are removed
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ATTACHMENT 3
'i Proposed Technical Specification License Nos. DPR-39 and DPR-48 Appendix A,' Sections 3.2, 3.4, 3.8 and 3.9 Technical Justification of the Modification
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ATTACHMENT 3 Justification of The Proposed Amendment Zion Technical Specifications, Appendix A, Sections 3.2, 3.4, 3.8, 3.9 Improvements in the analytical techniques used for FSAR accident analyses have allowed the Boron Injection Tank (BIT) concentration to be reduced or the entire BIT to be removed in Westinghouse NSSS Plants. The high boric acid concentrations currently required in the BIT impose operational and maintenance problems (minimum volumes and concentrations in boric acid system tanks, heat tracing malfunctions, BIT valve testing, and recovery from inadvertant safety injection) which adversely affect plant availability. Furthermore, the NRC has identified a safety concern involving boric acid solidification rendering emergency core cooling equipment inoperable (ie, Indian Point). Consequently, Westinghouse was contracted to evaluate BIT removal at Zion Station. The analysis allows flexibility for Zion Station to perform any action from removing the BIT entirely to having it filled with demineralized water.
The sole function of the BIT is to provide a negative reactivity insertion, via concentrated boric acid, to mitigate the consequences of the uncontrolled cooldown resulting from steamline break accidents. Although the BIT will function on various size steam breaks from any power level, the limiting cases used in the licensing bases are the hypothetical steam break, (a double-ended rupture of a main steamline), and the credible break, (the spurious opening of an atmospheric relief, steam generator safety, or steam dump valve.) The hypothetical break is a condition IV event; therefore the radiation release must remain within the llmits specified in 10CFH100. The credible break is a condition II event and releases must meet the requirements of 10CFR20. Thus, the FSAR cases described above were reanalyzed assuming the BIT remains installed with a boron concentratior. of 0 ppm to verify the limits are met. This is the most restrictive case since a slight RCS dilution would result.
The cases examined in this analysis assumed the following conditions exist at the initiation of the event:
- 1) The reactor is at end-of-life (EOL) burnup wi ti 9 minimum EOL shutdown margin, in hot shutdown v 'h .
equilibrium Xenon, and the most reactive rod (. 'R ) 'ully withdrawn. If the reactor was critical or at i ver, the reactor coolant system (RCS) contains more stori energy than at no-load, due to both energy stored in fu ' and the higher coolant temperature. Following the steam reak, this additional energy must be removed, via the c '1down,
b; fore ths no-load condition cc:umed in thD cn21y;in is reached. Furthermore, steam generator water inventory is greatest at no-load. Therefore, the magnitude and duration of the RCS cooldown is less for a steam break at power.
- 2) The moderator temperature coefficient (MTC) corresponds to the BOL (most negative) rodded core with the most reactive rod fully withdrawn. The variation of MTC with temperature and pressure is accounted for in the analysis. The core properties in the region near the affected steam generator are combined with the rest of the core to ob*.ain average core properties for reactivity feedback calculations. Further, a uniform core power distribution was assumed. This is conservative because it results in an underprediction of negative reactivity feedback in the region near the stuck rod (highest point).
- 3) Minimum capability for injection of boric acid (2000 ppm from RWST) solution corresponding to the most restrictive single failure in the safety injection system.
- 4) The Moody curve (1) for f(L/D) = 0 is used to compute steam flow. This is the same assumption used in the PSAR.
- 5) perfect moisture separation in the steam generator is assumed, yielding conservative results since carryover, which is in fact considerable, reduces the magnitude of the RCS cooldown.
- 6) The design value of th4 steam generator heat transfer coefficient is used, siaee the variation of this value has been shown to have litt13 impact on the transient.
For this steamline break analysis, system parameters such as RCS temperature, RCS pressure, steam flow, boron Concentration, and core heat flux were calculated using the LOFTRAN(2) system st9nsient analysis code, instead of the MARVEL (3) code, which was used in previous FSAR analyses.
The thermal-hydraulc behavior of the core is modeled with the THINC (4) code. THINC also performs a DNB analysis for the core conditions determined by LOFTRAN to verify the design basis is met.
The limiting case examined for the credible break is a steam flow of 247 lbm/sec at 1100 psia with offsite power available, which is the capacity of a single steam dump, relief, or safety. The steam release results in an initial increase in steam flow, which decreases during the accident as pressure falls. The removal of energy from the RCS causes a reduction of coolant temperature; due to the negative MTC, the cooldown yields a positive reactivity insertion, protection is provided from the pressurizer low pressure SI, steamline 6 P SI, OpaT trip, high flux trip, and main feedwater isolation. New acceptance criteria allow a return to criticality for this event if it is assured that no consequential fuel damage results.
l The limiting hypothstical break casa is the 4.6 square foot double-ended rupture at the steam generator exit (largest possible) with offsite power available. (Note loss of offsite powec results in reactor coolant pump trips and poorer coupling between primary and secondary, hence
,a cooldown of lesser magnitude.) The pipe rupture causes increased steam flow and a decreasing steamline pressure. 'A positive reactivity insertion results, as with the depressurization, although of much greater magnitude.
With-the most reactive rod stuck out, the return to power is.more severe.
This is a significant problem because of the high peaking factors associated with the most reactive rod withdrawn. Non-uniform core inlet temperatures,-
due to the rupture, must also be accounted for; the coldest core inlet temperature- sre assumed to occur in the region of the stuck rod. Local voiding adjacent to the stuck rod, in conjunction with the large negative MTC, partially offset the effect of the stuck assembly. Safety injection is initiated by steamline6 P. Steam release from more than one generator is prevented by main steamline check valves. Criticality is attained before 2000 ppm borated water is delivered to the RCS from the SI system. The dalay arises from actuation of the SI signal, time required to stroke open RWST isolation valves, time required to close VCT isolation valves, and time needed for the 2000 ppm boron solution to reach the RCS. (Because of hydrogen overpressure in the VCT, the charging pumps will continue to draw from the VCT until the isolation valves close. This was not considered in the original Westinghouse analysis. An evaluation was performed in which these delays were addressed and found to be acceptable. (5). Safety injection flow can then begin to decrease reactor power.
Main steam ruptures outside containment require MSIV operation to prevent blowdown of all steam generators and the SI signal is generated by the high steam flow SI logic. Therefore, they do not produce as severe a return to power transient as the in-containment case, primarily due to nozzles in the containment main steam pipe which limit the maximum steam flow for any break further downstream.
For the hypothetical break, (ie, most limiting condition IV break,)
the analysis shows the radiation reletse $s within the requirements of 10CFR100 by demonstrating the DNB design basis is met. Although preventing clad damage is not necessary for condition IV events, the analysis shows the DNB ratio remains greater than the limit. The dose calculations, which are performed assuming five percent failed fuel, demonstrate that the condition IV accident criteria are satisfied.
The credible steamline break analysis was performed without the previous Westinghouse internal criterion prohibiting return to criticality for condition II events. This is permitted under NRC and ANS criteria, which require that radiation releases remain within the limits of 10CFR20.
In order to allow substantial reduction in BIT concentration, Westinghouse now allows a return to power for condition II events, but the DNB design basis must be met in order to satisfy 10CFR20 dose requirements. DNB analyses show the design basis was met and no fuel failures were predicted.
This enslysis demonstrates that Zion Station may reduce BIT boron concentration to O ppm or remove the BIT entirely. The safety. criteria for the main steam depressurization (condition II event) and the main steamline rupture (condition IV event) accidents analyzed in chapter 14 of the PSAR are still valid.
Footnotes to :
- 1. F.H. Moody, "Maximum Flow Rate of Single Component, Two-Phase Mixture", Paper No. 64-HT-35, and ASME Publication.
- 2. Burnett, T.W.T., et el., LOFTRAN Cr e Description",
WCAP-7907-P-A, April 1984
- 3. J.M. Geets, "MARVEL - A Digital Computer Code for Transient Analysis of Multiloop PWR Systems", WCAP-7909, June 1972.
- 4. Hochreiter, L.E., Chelemer, H. and Chu, P.T., "THINC IV An Improved Program for Thermal-Hydraulic Analysis of Rod Bundle Cores", WCAP-7956, June 1973.
- 5. Letter to F.G. Lentine from Westinghouse Electric Corporation, CWE-87-191, July 22, 1987. "Zion BIT Removal Steambreak/SI Response Time Evaluation " (SECL No.87-343)
ATTACHMENT 4 Proposed Technical Specification License Nos. DPR-39 and DPR-48 Appendix A,-Sections 3.2, 3.4, 3.8 and 3.9 No Significant Hazards Consideration 1
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ATTACHMENT 4 Evaluation of Significant Hazards Consideration Proposed Changes to Zion Technical Specifications, Appendix A, Sections 3.2, 3.4, 3.8, 3.9 Description of Amendment Request An amendment to the Zion Facility Operating License is proposed to remove the Boron Injection Tank (BIT) and the piping and valves associated with the recirculation of the BIT to Boric Acid Tanks (BAT).
Background
10 CFR50.92 states that a proposed amendment will involve a no significant hazards consideration if the proposed amendment does not:
(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.
The discussion below addresses each of these three criteria and demonstrates that the proposed amendment involves a no significant hazards consideration, Basis for No Significant Hazards Consideration Determination Does the proposed amendment (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety?
l Discussion Item No. 1 The purpose of the Boron Injection Tank (BIT) is to provide a negative reactivity insertion during accidents involving accidental depressurization of the Main Steam System. Removal of the BIT will not increase the probability of such an accident because the initiating events for such accidents insalve either a steamline break or' accidental opening of a steamline safety valve or relief valve. None of this equipment is affected by this Technical Specification change.
The consequences of accidents involving accidental depressurization of the Main Steam System will not be increased by the removal of the BIT.
The revise / safety analysis concluded that event without the negative reactivity insertion provided by the BIT as currently required, normal safety injection from the RWST is sufficient to ensure that no fuel failures occur and DNB is not reached.
The probability and consequences of accidents evaluated in the PSAR will actually be decreased by this Technical Specification change for two reasons. The high concentration of Boric Acid in the BIT may contribute to the probability of failure of ECCS components due to boric acid solification. Reduction of boric acid concentration in the ECCS charging discharge flowpath will reduce this probability. Additionally, by deenergizing or disabling in their safeguards position the BIT inlet and BIT to BAT recirculation valves, the station is removing the possibility of failure of these active components during a safeguards Actuation.
Discussion Item No. 2 The proposed amendment would remove equipment which is no longer needed, and would not add any new equipment, and therefore does not create a mechanism for a previously unevaluated accident. By removing or disabling in their safeguards position valves which were required to actuate during a safety injection, this proposed change will actually reduce the possibility of active failures of this equipment. The configuration which will exist following receipt of this Technical specification change will be functionally analagous to, but simpler than, the configuration for which the plant is currently analyzed during normal operating and post accident conditions.
Discussion Item No. 3 A Departure from Nucleate Boiling (DNB) analysis was performed for the cases most critical to DRB. The most limiting case was shown to be a double ended rupture of the main steamline at the steam generator exit. For all cases, the revised safety analysis submitted showed that the DNB ratio design basis is met and no fuel failures are predicted. Therefore the margin of safety is not reduced from the previous analysis.
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ATTACHMENT 5 Proposed Technical Specification License Nos. DPR-39 and DPR-48 Appendix A, Sections 3.2, 3.4, 3.8 and 3.9 "Boron Concentration Reduction / Boron Iniection Tank Elimination for Zion Units 1 and 2" by J.C. Bass, October 1985 I
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