ML20148K398
| ML20148K398 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 07/08/1975 |
| From: | Vandenburgh D YANKEE ATOMIC ELECTRIC CO. |
| To: | Purple R Office of Nuclear Reactor Regulation |
| References | |
| WYR-75-70, NUDOCS 8012040444 | |
| Download: ML20148K398 (44) | |
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, , Proposed Change No. 117-4: / '
" Supplement No. 3 Telephene 617 366 90ll
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YANKEE.- ATOMIC ELECTRIC COMPANY urR n -,0
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20 Tumpike Road Westborous h, Masso:hussrts 01581
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July 8, 1975 United States Nuclear Regulatcry Comnission Washington, D. C. .20553 TWS DOCUMENT CONTAlWS w P00R' QUn.UTY PAGES Attention : Robert A. Purple, Chief /--- Operating React: ors 13 ranch No.1 Division of Reactor Licensing
Subject:
Order for :ECCS Re-atalyst.s % . References (1) License No. DPR 3 (Docket No, 50-29) (2) " Order for Modification of Lit:ense." From R. A. Perple (NRC) to Yanken Atomic Sluctric Conpany, Attention: i G. C. Andognini, Decerber 27, 19'; 4. (3) Preposed Change No. 117., "Revi sed FCCS Analysis,' submitted to DCL/AEC on July 31, 1974. (4) Proposed Change No. 117, Supplement No. 1, " Revised ECCS Analysis," submitted to DCL/AEC on De: ember 12, 1974. (5) Preposed Chan;e No 117, supplement No. 2, " Revised ECCS Anatlysis," submi':ted to D0L/AE.7 on Decerrber 30, 1974. (6) Letter fro'.n R. A . Purple DOL /N30 t.o Yan).ee Ator.ic Electric Company, Attention: G. C. Andogniti,, June 13, 1975. l (7) Change No. 97 to License No. DPR-3, Docket No. 50-29. (0) Letter frcm D. E. Var.denburgh 1%EC to DOL /AEC, "Containrent Data,
- Nove:rber 26, 1974.
Dear Sir:
1 Pursuetnt to section 53.59 of the Cotunission's Rules and Regulations a .d in response t.o your Order fer Modification of License (Reference ?) , t.:e Yankee Atonic Electric Ccmpany hereby submits the following Supple, ment ,
' c . 3 to Proposed Change No .117 (Reference 3) thereby superceding l
Supplements No. 1 (Reference 4) and No. 2 (Reference 5). PJOPCSED CHANGE: Reference is made to Section 411 of the FHSR of the referenced license. We propose to mcdify Section 411 as described below. REASON FOR CHANGE: In order to conply with your order for Modification of License, a revised ECCS analysin has been completed and modifications to previous results (Reference 3) atre required. 8012040ffy ^
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( United States Nuclear Regulatory Commission July 8, 1975 Attn: Robert A. Purple, Chief Page Two o DESCRIPTION OF CHANGE: Consistent with the ECCS perfommance re-analysis the following revisions and addition are to be made to Section 411 of the referenced Proposed Changer Remove pages 411:05, 411:06, 411:08, 411:26, 411:30, 411:31, 411:33, 411:35, 411:39, 411:43, 411:44, 411:46, 411:48, 411:52, 411:56, 411:57, 411:59, 411:61, 411:65, 411:69, 411:70, 411:72, 411:74, and replace them with the corresponding attached pages. Add page 411:07a SAFETY CONSIDERATIONS: The safety considerations of the revised analysis are presented belcw in the order of requested information identified in Reference 6. Small Breaks have not been re-analyzed for the following reasons: (1) The Yankee Ro'e core is sufficiently small that power cannot be made to peak towards the core top. This reduces the consequences of small breaks where portions of the upper core may be uncovered for brief intervals during blowdown; (2) The ECCS system is designed with high pressure injection capability to minimize those periods when portions of the upper core nay be uncovered for small breaks; (3) The previously submitted ECCS performance analysis (Reference 3) included small breaks and showed that they were not limiting. Changes made in the evaluation model since the previous anaylsis, and changes made in the assumed safety injection capability (see References 4 and 5? at low pressures, wculd not change the small break results. Indeed, the revised analysis. shows that the most liniting breaks did not change in size and a sufficient nurber of breaks have been analyzed to show a trend of decreasing peak clad tenperature with decreasing break size. Breaks in the hot leg or pump suction lines have not been analyzed for the following reasons: (1) The check valves in the discharge of the main coolant pumps prevent significan: backflow thrcugh the broken loop. This effectively reduces the size of the break and prevents the inducement of the flos stagnaticn in the core during blowdown; (2) A hot leg or pump suction line break would induce strong forward coolant flow through the core during blowdown, thus providing good core cooling. In contrast, a cold leg break would tend to induce flow reversals and stagnation during blowdown, thus yielding poor core cooling; (3) The steam vent path during the reflood phase would have a lower resistence for a hot leg or pump suction line break;
. . - - . - . . .=
United States Nuclear Regulatory Commies. ion July 8, 1975 Attn: _ Robert A.' Purple, Chief Page Threa o (4) 'A hot leg break was previously analyzed' (Reference 7) . This ) analysis unequivocally showed that a cold leg break is much l more limiting. ! 1 l
]. The four large breaks previously analyzed (Reference 3) have been i re-analyzed and results show an increased margin to 10 CFR 50.46 limits.
l+ . 1
- i. Until a 3-loop ECCS analysis is available, the Yankee Nuclear Power
; Station will not be operated under 3-loop conditions.
l
'The potential for boron precipitation in the. core during long term
- recirculation following a Loss of Coolant Accident is addressed in '!
Enclosure A. In summary, it was_shown that a minimum of 25.2 gpm injected
; into the hot leg piping with initiation between 20 and 24 hours would .
preclude the possibility of boron precipitation. Component qdalification ; in the LOCA environment is addressed in Enclosure B. The re-analysis is r required to incorporate the electrical and operator failures addressed by Branch Technical Position EICSB-18 and to address _the potential for a l'
.pansive fluid system failure. We. conclude that the ECCS can accept a single active failure, operator error, or passive fluid system failure :
following a LOCA and still' perform its required safety function. ! The specific equipment arrangement within the Yankee Plant has been ! reviewed- (cre Enclosure C) and no valve motors within containment required i for ECCS operation or long term cooling (either cold leg or hot leg) will ; becon.e submerged following a LOCA. i The conteinment pressure used to evaluate the performance capability ! of the ICCS has been calculated in accordance with the provisions of ; Branch Technical Position CSB 6-1. Input data to the results of the ! contair.r.ent analysis have been previously transmitted to the commission j
. (Reference 8) . '
The ECCS analysis has been performed with Westinghouse's 12-25-74 ! evaluation model which was determined to be wholly in conformance with ! 10 CFR 50.46 and Appendix K thereto as stated on page 15 of the Commission's l Safety Evaluation Report (Reference 2). Thus, it is our position that j this sthmittal is in conformance with the Commission's December 27, 1974 l Order f or Modification of License. However, your letter of 6/13/75 l requested the use of a more recent Westinghouse model for specific j
. situations. This model would be applicable to Yankee Rowe. Due to the '
lateness of your request (June 13, 1975) , this reanalysis could not be i conducted by the July 9th deadline established in your " Order for ! Modification of License" which was dated December 27, 1974. Core XI is currently approaching the end of its full power lifetime at which l time the reactor power will be gradually decreased until the reactor is l shutdown for refueling in October. The analysis provided for reload Core XII will conform to your requirements which have been established i ! to date. l i i l, ace r - __ _
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(/ (- United States Nuclear Regulatory Commission July 8, 1975 Attn: Robert A.'Purpl6, Chief Page Four l This supplement to Proposed Change No. 117 has been reviewed by the Nuclear Safety Audit and Review Conunittee. ! SCHEDULE OF CHANGE: We respectfully request review of this Proposed Change by the Commission as soon as practical. Any questions regarding this Proposed Change should be directed to Mr. Alan E. Ladieu at our Engineering Office, 20 Turnpike Road, Westboro, Massachuse.tts 01581, (617) 366-9011, Extension 2845. Respectfully submitted, YANKEE ATOMIC ELECTRIC COMPA,NY
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D. E. Vandenburgh DEV/kg Vice Presider t Enclosures COMMONWEALTH OF MASSACHUSETTS)
)ss.
COUNTY OF WORCESTER ) Then personally appeared before me, E. E. Vandenburgh, who, being duly sworn, did state that he is Vice President of Yankee Atomic Electric Company, that he is duly authorized to execute and file the foregoing request in the name and on the behalf of Yankee Atomic Electric Company, and that the statements therein are true to the best of his knowledge and belief. I, i A t e t #W , N [ Armand R. Soucy ' Notary Publtc ~ , My Comnission Expires September 9, 1977 ' N 1 d h h s In l
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, 4 inch break. The peak clad temperattres for the 3 inch break is given in Figure 411-11. The peak clad temperature for the 2 inch break ic not shown since the core is not uncovered during the tr.2nsient.
The analyses presented demonstrate that the Emergency Ccre Cooling System provides sufficient core floeding to keep the calculated peak clad temperatures below the required limits of 10CFR50. 46. Hence, adequate protection is afforded by the 1:nergency Cort Cooling System in the event of a small break loss of ccolant acciden'. Large Ercak Analysis At, the beginning of the blowdown phase of a major break, the entire main coolant system contains subcooled liquid which transfers heat from the core by forced convection with some fully developed nucleate boiling. After t'he break develops, the time to departure from nucleate boiling is calculated, consistent with Appendix K of 10CFR50(5). Thereafter, the , core heat transfer is unstable, with both r ucleate boiling and film boiling occurring. As the core becomes ur1 covered both turbulent and laminar forced convection and radiation are considered as core heat transfer mechanisms. When the main coolant system pressure falls below G20 psia the q accumula. tor begins to inject borated water. The conservative assumption h is made that accumulator water injected bypasses the core and goes out ( through the break until the termination of bypass. This conservatism is again ccnsistent with Appendix K of 10CFR50. i The Emergency Core Cooling System, even when operating during the injection mode with the most severe singic active failure meets the Acceptance Criteria (5), The description of the various aspects of the LOCA analysis is given in WCAP-3339(6). This document describes the tiajor phenomena nodeled, the inte:: faces among t.he computer codes and features of the codes which maintain compliance with the Acceptance Criteria. The Individual codes are described in detail in separate reports (7-J0). The containment parameteis used in the containment pressure code (10) for determination of ECCS backpressure are given in Table 411.2. The analysis presented here was performed with the version of the Nestinghouse evalut. tion model with modifications as specified by the staff in Reference 12 and complies with Appendix K to 10CFR50.46. Tabic 411-1 presents the peak clad temperatures and hot spot metal reaction for a range of break sizes. The analysis of the loss of coolant accident is performed at 103 percent of rated power. The peak linear power, and cere power used in the analyses are also given in Table 411-1, along with the equivalent core peaking factor at the license application power level. Since there is a margin between the value of t he peak linear power density used in this analysis and the value expected in operation, a lower peak clad temperature vould be i obtained by using the peak linear power density expected during operation. J l j
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f .,:, [; ' U' j -411t06 1 L l' h Revised 7/9/75 ! h d DRAFT l j i Fitures 411-12 through 411-24'preacnt the t:ransient values of the l [,;1 , princip1h parameters for the break siaes analyzed. These figures l l !! /> pres 2nt the following informatinnt j
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. 'b 411132- Value of fluid qu$lity at the clad burst location and at iI the hot spot. !!
l 411-13 tiass velocity at the clad Lurst location and at'the 'i
/!
> hot spot.
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411-14 Filn. heat transf er coef ficient ocalculated by !LO(ITA IV l at the clad burst location and :at thci hot s po t . I i
- j , '4.11115 Presnure calculates in the core. *
.J.i .\ {< 411-16 Break flowrate. F:r guillotine breaka ' this is the -
.sua of the flows out both ends of the break. i
] 1 411-17 Coxe pressure drop from the lower plenum, near t he y co::e, to the upper plenum at' the core outist. l i 411-18 Clad temperature transient at the hot. spot and s,t the ( burnt location. 't- i 411-19 Fluid tecperature at the hot spot and burst location. { ')' i
, 411-20 Core flowrate (top and bottom) . , j f
411-21 Core reflood transient, p i 411-22 -Accumulator flow. Established in the refill-reflood f calculations and is the sum 'of that injected in the [ intact cold legs. ; 411-23 Pumped ECCS flow. [ I 411-24 Containment pressure transient. ! The clad temperature analysis is based on a total peaking factor ! of 2.96. The hot spot cetal water reaction reached is :2.0 percent, which is well below the embrittlement limit of 17 percent:, as required by , 10CFR50.46. In addition, the total core metal-water reaction is less than ! 0.3 percent for all breaks as compared with the 1 percent criterion of l 10CFR50.46. l
. e I The results of several sensitivity studies are reported in Reference !'
These results are for conditions which are not limiting. in nature
~
11. and hence are reported on a generic basis. j t CONCLUSIONS l For breaks up to and including the double ended severance of a l reactor coolant pipe: l
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h h h b N b h d d @ d E d M i h it m r - - ~" ., , 3 i i. ()Revised 7/9/75 411:07a
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rederal Register, " Supplement 'to the Status Report by the
' '(12) Directorate i
of licensing in the Matter of Westinghouse Electric Company ECCS Evaluation Model Conformance to 10CFR50, Appendix K", . Novceber 1974. !
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_n TABLE 411-1 - LARGE BREAK ANALYSIS
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SGGIARY OF RESULTS* -
-~~
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_ Break Type DECLG** . DECLG _ 0.6 x DECLS*** l.0 ft2 3 * (CD = 1.0) (CD = 0.6)
~
Peak Clad Temperature, OF 1914 1978 1921 1860 Location of Peak, ft 3.8 3.8 '3.8 4.1 Local Maximum Zr-H 2O Reaction, % 1.4 2.0 1.6 ' 1. 3
^
Location of Max. Zr-H 2O Reaction, ft 4.1 4.1 4.1- 4.1 Total Zr-H 2 O Reaction, % <0.3 <0.3 <0.3 <0.3
.O Hot Rod Burst Time, sec No No No . No Burst Burst Burst Burst Hot Rod Burst Location, ft
~
- Calculations are based on Reactor Power, Mwt 618 Peak Linear Pcwer, kw/ft 13.06 -
1 Peaking Factor at Rated - Power 2.96' , a Accur::ulator Water Volume, ft3 700 O 1 U Fuel Cycle 11 ';l l, - g O'
- Fuel Region B
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**DECLC - Double-ended cold leg guillotine break. '
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***DECLS - Double-ended cold leg slot break. '** '
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P00R ORIGINAL YA::KEE I;UCLEAR LOSS OF COOLA5I ACCIDEhT FIGURE PO'n'ER STATION HOT SPOT FILM COEFFICIEliT VS. TIME 411-14A DECLG (CD=1.0)
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e O p l l 1 YANKEE NUCLEAR LOSS OF COOLANT ACCIDENT FIGUllE POWElt STATION PUMPED ECCS FLON (!!EFLOOD) 411-23D 1.0 ft2 S10T .
~
f'., , O G ENCLOSURE A SUPPORTING ANALYSIS CONCERNING BORON CONCENTRATION FOLLOWING A LOSS OF COOLANT ACCIDENT
- During the long term cooling mode following a Loss of Coolant Accident, (LOCA) the potential exists for the reactor core boron concentration to reach the precipitation level. Specifically, this phenomena could occur following a cold leg break with cold leg safety injection.
- A large portion of the flow injected through the cold leg in this case would only be effective in ma.intaining reactor water level. Therefore, water injected through the cold leg that would reach the vessel would be that amount which would replace the water that boils away as a result of the generation of decay heat ener gy. The steam that is leaving the vessel under these conditions carries little boron, resulting in a steady increase in the core boron concentra-tion until the precipitation limit is reached.
The solubility of natural boron in water as a function of temperature is shown in Figure 1 as taken from Refercnce 1. At 2120F, Figure 1 depicts a l solubility limit of 47,500 ppm. j 1 One of the requirements of Appendix K to 10CFR50 is that the fuel be maintained in a coolable geometry following a LOCA. Maintaining a coolable geometry would be consistent with avoiding channel blockage due to boron prceipitation. T'o prevent' the core boron concentration from reaching solubility limits fc 11owing a LOCA, part of the flow f rom the long term recirculation system l will be diverted so as to deliver cooling water to the hot leg piping. The l remaining recirculation flow will continue to be injected into the cold leg i piping. This scheme will be employed independent of break location. An analysis was performed to determine when to initiate hot leg injection and to determine the minimum required flow to prevent boron precipitation. To determine the core boron concentration as a fun'etion of time following . a LOCA, the analysis was divided into two phases. These phases were: 1) the boron concentration build-up prior to hot leg injection, and 2) the boron concentration response following injection into the hot leg. The assumptions made for the phase 1 analysis were:
- 1) Core cooling is provided by cold leg injection during this phase; i
- 2) It it assumed that the core boron concentration begins to increase '
at the initiation of core reflood;
- 3) The boron concentration of the injection water is assumed to be equal to the time dependent containment sump concentration;
- 4) The initial sump and core concen'tration is 2314 ppm. It is a volume weighted average calculated from the volumes and maximum initial concentrations of the fluids stored in the main . coolant system, the refueling water storage tank (RWST), and accumulator before the event. These volumes and concentrations are as follows:
Volgme Initial Concentration (ft ) (ppm) Main Coolant System 2,920 1800 RWST 16,709 2400 Accumulator 752 2400
I's - [
- 5) The rate of cold leg injection. flow actually reaching the core is a :
function of time' and is based on the amount of fluid boiling off; s
- 6) The steam leaving, as a result of boil-off carries no boron with. it; ;
- 7) System pressure is conservatively assumed to be atmospheric; '
- 8) No credit is assumed for subcooling of the injection water. .This maximizes the boil-off rate which, in turn, maximizes the rate of increase ,
of the core boron concentration; . ,
- 9) Boron is assumed to concentrate in the core and*the region above ,
the core up to the bottom of the hot leg nozzles. This volume is approximately 278 ft3; 3
- 10) Decay heat is represented by the equation Q = Qo At-a, where l A = 0.130 and a = 0.283 (see Reference 2); ;
- 11) Prior to the LOCA, the reactor is assumed to be operating at full power {
(600 Mwt). 1 The resulting core boron concentration versus time during phase 1 is shown ,
- in Figure 2. It'was conservatively calculated that boron solubility limits l could be attained in 26 hours without corrective action. l The assumptions made for the phase 2 analysis were: ;
- 1) . Hot leg injection is initiated between 20 and :24 hours following the }
LOCA; ;
- 2) The boron concentration of the11njection water is assumed to be equal i
. to the time dependent containment sump concentration; ,
- 3) The initial core and sump concentrations at the start of hot leg injection are provided by the results of the phase 1 analysis; l
- 4) The difference between the hot leg injection flow rate and the core i steaming rate is the liquid flow that is effective in removing boron j from the core;- .
t
- 5) The hot leg injection flow rate is assumed to be constant following initiation; ;
- 6) The steam that leaves the core removes no boron-
- 7) The flow rate leaving the core as liquid has a concentration equal ;
to the core concentration;
'8) No credit is assumed for subcooling of the injection water. This j maximizes the boil-off rate which, in turn, minimizes the hot leg ,
injection effectiveness;
- 9) The boron concentration is assumed to be highest in the core and the j region above it up to the bottom of the vessel nozzles; ;
- 10) Decay heat is represented as in the phase 1 analysis; 11)- Prior to the LOCA, the reactor is assumed to be operating at full j power (600 FMt);
- 12) System pressure is conservatively assumed to be atmospheric.
An equation for the time ierivative of the core boron concentration following ! hot leg injection was developed. From this equation the minimum flow rate ' l required to prevent further increases in concentration was calculated. In conclusiona minimum hot leg injection flow ' rate of 25.2 gpm initiated between 20 and 24 hours following a LOCA will keep the core boron concentration well : below solubility limits and will provide more than enough water to replace { boil-off. I i l
Q Yankee proposes to utilize a charging pump for hot leg recirculation. There l are three charging pumps each rated at 33 gpm and capable of being powered i from an emergency diesel. The charging pumps discharge through the normal charging path to the hot leg of loop 4. The charging line will be protected where necessary to insure its integrity follouing a cold leg rupture. The
- charging connection is on the reactor side of the loop 4 Th' loop-isolation valve. I
. .- - l Yankee proposes to modify the existing recirculation system to provide a di-rect flow path from the discharge of the purification pumps during post acci--
~
dent recirculation to the suction of the charging pumps. The crossover pro-posed is shown on Figures 6.3-2 and 9.3-4. It consists of a line with a . locked lopen globe valve in series with a check valve which joins the recirc- , . ulation header to the charging pump suction line. -During post accident re-circulation, injection to the hot leg is established by starting a charging pump. t t f I 9 _ _ _ _ _ _ _ _ _ _ _ . _ __ m _ - , - - - . _c , . . , . . . -_ _ _, y , ,_
. REFERENCE $' .
- 1. D. E. Byrnes and W. E. Foster, " Literature' Values for Selected Chemical /
- Physical Properties of Aqueous Boric Acid Solutions", WCAP-1570, (January 1961).
- 2. ANS Standard Decay Heat Release Rates Following Shutdown of Uranium-Fueled Thermal Reactors, October 1971. , ,
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ENCLOSUitE B SINGLE FAILURE ANALYS'IS 4 j1 Yankee has reanalyzed the ECCS for single failure. The reanalysis is I required to incorporate the electrical and operator failures addressed j, by Branch Technical Position EICSB-18 and to address the potential for a passive failure. i ,
- 1)
! Valve operation failures which result in adve'rse ECCS capability are listed below: - , SI-MOV This valve is closed when the accumulator is exhausted. This I op open. aation provides a back-up to check valve SI-V-2 in case it hangs Premature closure of this valve would'stop all flow from the ; ( accumulator and delay recovering of the core. ! i l j
; Yankee proposes to prevent the possibility of the operator closing the i o valve prematurely by modifying the control circuitry to include permissive i l contacts which require a two out of four accumulator low level signal logic to complete a manually initiated close signal. The position of the j valve is indicated by position lights on the control board.
CS-MOV-535 - This valve is normally open and must remain open during j the initial injection phase. The valve provides the capability to l' isolate the low pressure safety injection header in the event that 1 CS-MOV-533 or CS-MOV-534 fail to function for cut-back mode injection } flow which is initiated 15 minutes after the LOCA. In this case,
,CS-MOV-535 can be closed, securing all flow in the low pressure safety
- injection header and injection flow regulated by SI-MOV-46 in the high pressure injection header. $
Yankee proposes to prevent single operator error by removing and placing the control switch handle under administrative control, h 1 SI-Mov This valve is normally open and is required to remain open during the initial injection phase to provide boosting to tha HPSI pumpe for small break capability. During the injeccion'eut-back phase, for a large break, this valve is required to be operable to provide back-up capability to isolate flow to the HPSI header in case SI-MOV-46 fails to ;
- limit flow during the injection cut-back mode. Valvo position is i
- indicated on the control board. The breaker cubicle for this valve is
- located large in the break PAB and is not accessible to the operator following a LOCA.
3 Yankee proposes to prevent inadvertent operation of this valve due to a single operator error by removing and placing the control switch handle under administrative control. [ CS-Mov-533 a LOCA.
- This valve remains open for the first 15 minutes following After 15 minutes, the valve is closed and flow regulated with '";
CS-MOV-534 or SI-MOV-46. Premature closure can reduce injection to less l than required flow rates. Valve position is indicated on the control board. , O t i
l O _2_ o B 4 Yankee proposes to prevent premature closure of this valve due to a single operator arror by ren3ving and placing the centrol switch handle i 1 under administrative control. SI-MOV This valve remains open during the initial injection phase following a LOCA,'i.e., 15 minutes. The valve is required to be operable after 15 minutes for flow control during the. cut-back phase. Valve , position is indicated on the control board. The bre,aker cubicle for this valve is located in the PAB and is net accessible' to the operator following a large break LOCA. Yankee proposes to prevent premature closure of the valve due to a single operator error by removing and placing the control switch handles under administrative control. SV-1 and SV These are solenoid operated pilot relief valves which open
. to depressurize the accumulator and minimize Up injection. They are automatically operated by two out of four accumulator low level signals which meets single failure criteria. Premature opening would result in decreasing the accumulator driving force.
The control circuit has a permissive which is satisfied by one of two low lavel signals. Yankee proposes to meet single error by removing and placing the control switch handles under administrative control. CS-Mov-532 - This valve provides recirculation back to the safety injection tank (Tk-28) during pump tests. Otherwise, the valve remains closed during normal operation and during the accident. If the valve was opened during a LOCA injection, flow from the LPSI pumps may be, reduced. The valve is prevented from opening during a LOCA and it will close automatically (if open) by a safety injection signal. Yankee preposes to prevent singic operator error by removing and placing the control switch handle under administrative control. CS-MO7-536, 537, 538, 539 and SI-MOV-22, 23, 24, 25 - These valves provide two valve injection shut-off capability in the safety injection header branch lines to each of the reactor coolant loops. The valves are open during normal operation and will remain open during a LOCA unless the LOCA is a rupture of one of the SI header branch lines downstream of the check valve. In this case, the rupture would be small and hold up reactor coolant system pressure. Ihe SI flow would tend to go out the ruptured branch line instead of the intact branch lines to the. reactor coolant system. The operator identifies the ruptured branch line using the flow indicators in each branch line. Since two valves are in series in each branch line, I shutoff to the ruptured branch is assured. Flow in the intact branch lines is assured as soon as flow in the ruptured branch line is secured. The ECCS analysis did not assume isolation of the flow to the break, therefore, it is not essential that the valves remain operabic from the control room; however, retaining operability makes the ECCS more effective and is considered a desirabic feature from a safety view. If, hot.cver, ' the wrong valve closes during a LOCA, it has an adverse effect on ECCS.
1 j Yankee: proposes to analyze the effects.to'ECCS considering the erroneous ' I closure of flow in one of,the branch lines. Yankee proposes to prevent r single operator error by removing and placing the control switch handles l under administrative control. ' t MC-MOV-301, 302, 309, 310, 318, 319, 325, 326 - These are loop isolation ( valves and have several safety functions. They grovide the capability to:
.1. Isolate a ' steam generator tube rupture and re-establish containment. I
- 2. Terminate an1 uncontrolled cooldown due to a safety valve' sticking open. , , !
- 3. In the long term loop isolation are used to terminate a LOCA that l occurs from a failure of the loop piping.
o Erroneous operation of these valves in the early stages of a LOCA can , change the blowdown flow path to one that was not considered in the ECCS ! analysis. ! Yankee proposes to prevent erroneous valve operation by removing and ! placing the control switch handles under administrative ' control. CH-MCV-523, 524 - These valves isolate the flow path from the charging {
, purp discharge header to the hotleg of loop 4. These valves are open ;
during normal operation. l i The closure of either valve during the hotleg injection phase of a LOCA l Sd11 terminate hotleg injection. However, it is desirable to maintain l the ability to close these valves to divert the high pressure charging l flow through valve CH-MCV-522 for additional coldi,eg injection capacity. Yankea proposes to meet single operator error by removing and placing l the control switch handles under administrative control. ! L CH "C7-522 - This valve isolates flow from the charging pump discharge j header to the low pressure injection headur. It is closed during normal 1 operation. ' It is desirable to maintain the ability to open this valve to provide j additional celdleg injection capacity. Yankee proposes to meet singl'e operator error by removing and placing the control switch handle under administrative control. < I ACTIVE A:D PASSIVE FLUID SYSTEM FAILURES l
- 2) In addition to the electrical and operator failure analysis described above, Yankee has analyzed the ECCS for potential passive failures of fluid systems during long term cooling as well as singic failures of I
4
4 e of active components. This analysis considered the capability to initiate injection to the hotleg to prevent boron precipitation as an integral part of long term cooling. Except for those electrica) or operator failures which respond to , BTP EICSB-18, mentioned above, the ECCS system as it presently exists can perform its required safety function and accept a single active failure. ' The potential for passive failure of the ECCS ir the long term exists at valve packing glands, pump seals, gasketed pipe joints. . .etc. , j where pressure boundary integrity is maintained by seals of soft pliable i materials or consumable materials such as would be found in the mechanical seals of pu=ps. Figure 6.3-2 attached, shows the recirculation system with the valves positioned for the long term recirculation mode. , The potential for passive f ailure at each of ' he points identified above is examined as follows: 2
- 1. Containment sump flange - The internal standpipe is seal welded to I
a diaphragm located inside the cylindr,ical sump. The diaphragm is, in turn, welded to the walls of the sump. This diaphragm is fitted with a pipe plug so that the cavity between the gasketed flanged joint and the diaphragm can be pressurized for the purpose of testing the gasketed joint. In addition, the diaphragm provides an inner barrier during a LOCA which prevents the gasketed joint from being pressurized. Yankee concludes that there is no potential for passive failure in this area.
- 2. Punp Seals - The pumps are single stage back pull overhung pumps and as such have ene shaft sealing box which sees only suction pressure. The seal assembly consists of two mechanical seals in series with cican seal water injected in the cavity between the seals. Each seal is designed to functicn independent of the other.
The condition of each seal can be determined during normal operation by the make up frequency to the seal injection flask. Yankee concludes that a single seal failure will not result in a passive failure of the pump seal.
- 3. Valves Sten Packing Glands - All of the valves shown in the open position are designed with backscats which effectively isolate the packing glands when the valve is open. PU-MOV-545 is a globe valve which shuts off against recirculation flow and thus isolates its packing gland also. PU-MOV-542 on the other hand, shuts off against the low pressure surge tank and thus expcscs its gland to the recirculation section pressure. Failure of this packing gland results in a maximum estimated flow'of two gallons per minute. This flow is easily acco=nodated by the floor drain system.
- 4. Gasketed Pipe Joints - This includes valve bonnet to body joints as well. All such joints are located in the PAB or in the pipe 1
l
( -s- k-tuhnel which drains to the PAB. In each case, spiral wound (Flexitallic) gaskets are used which can be postulated to leak but not " blow out" due to their high design strength and the low pressures involved. This leakage is estimated not to exceed two gallons per minute. Yankee concludes that the Yankee Rowe ECCE can accept a passive . failure in the long term and perform. its required safety function. l + l 1
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E C I 4 ENCLOSURE C SUBMERGED VALVES ? i i Yankee has reviewed the equipment arrangement with respect to water
' level following a LOCA.
No motor operated va1ves inside the containment will be submerged following the LOCA. This review included all motor operated valves
' regardless of whether they are part cf the ECCS i
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.b NRC DISTRIBUTION FOR PART S0 DOCKET MATERI AL 4 (T,EMPOR ARY FORM) ;
. - CONTROL NO: 7307 4
FILE: DATE OF DOC DATE REC'D LTR TWX RPT OTHER FROM. Westboro,
- Yankee Ato:nicMass Electric 7-8-75 7-9-75 D E Vandenburgh XX TO ORIG CC OTHER SENTl;Rc PDR XXX ___
3 signed . SENT LOCAL PDR vvv CLASS UNCLASS PROPINFO INPUT NO CYS REC'D DOCKET NO: go 50-29 xxx DESCRIPTION: ENCLOSURES: j Ltr notarized 7-8-75 trans the Proposed change no. 117, following: supplement no. 3, concerns modification ^of [',/.3[' IIO II[COC. section 411, a revised ECCS analysis.... , 40 copies encl rec'd D. " , - . . - , , . _ _ PLANT NAME: Yankee Rove FOR ACTION /INFORMATION vtm 7-9-75 BUTLER (L) SCHWENCER (L) ZIEMANN (L) R EG AN (E) W/ Copies W/ Copies W/ Copies W/ Copics CLARK (L) STOLZ (L) DICKER (E) LE AR (L) W/ Copies W/ Copies W/ Copies W/ Copics PARR (L) VASSALLO (L) KNIGHTON (E) spies ' W/ Copies W/ Copics W/ Copics W/ Copies KNIEL (L) / PURPLE (L) YOUNGBLOOD (E) LPM W/ Copies W/oCopics W/ Copics W/ Copies
~
INTERNAL DISTRIBUTION l G FILE TECH REVIEW DENTON LIC ASST A/T IND . SCHROEDER GRIMES R. DIGGS (L) BRAITMAN gN t SALTZMAN 1 OGC, ROOM P 606A MACCARY GAMMILL H. GE ARIN (L)
,/GOSSICK/STAF F KNIGHT K ASTNE R E GOULBOURNE (L) MELTZ
- y. CASE PAWLICl$1 BALLARD P. KREUTZER (E)
GIAMBUSSO SHAO SPANG LER J. LEE (L) PLANS _ BOYD /STELLO M. RU3 inn 00K(L) MCDONALD MOORE (L) ENVIRO. S. REED (E) CHAPMAN UBE (Ltr DEYOUNG (L) / HOUSTON) MULLEk NOVAK(3 M. SERVICE (L) [E' C,"blMd RSON SKOVHOLT (L) [f0SS DICKER S. SHEPPARD (L)
/ PET GOLLER (L) (Ltr) / IPPOLITO KNIGHTON f /M SLATER (E)
P. CO L LINS jTEDESCO YOUNGBLOOD H. SMITH (L) HARTFIELD (2) DENISE /'J.COLI.IliS REGAN S. TEETS (L) KLECKER REG OPR LAIN AS G. WI LLI AMS (E) EISENHUT WIGG TON FILE & REGION (2) BENAROYA [_PJOJECT _%_E\/M LDRV. WILSON (L) ,, l' MIPC VOLLMER H AR LESS R. INGRAM (L) j/b. If. DUNCAN (E) d4b.gl ~_--. EXTERNAL DISTRIBUTION / V 1 - LOCAL PDR.G.R64M 86 O) kb 1 - PDR SAN /LA/NY 1 - TIC (ABERNATHY) (1)(2)(10) - N ATION AL LABS 1 - BROOKHAVEN NAT LAB 1 1 - W. PENNINGTON, Rm E.201 GT
/ .-1NSIC- ASLB (BUCHANAN) 1 - CONSULTANTS 1 - G. ULRIKSON ORNL 1 - Newton Anderson NEWM ARK /BLUM E/AG BABI AN q/,- ACRS H6hetQWSENT i
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