ML19329G212
| ML19329G212 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 07/10/1980 |
| From: | Sylvia B VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton, Youngblood B Office of Nuclear Reactor Regulation |
| References | |
| 553, NUDOCS 8007140158 | |
| Download: ML19329G212 (16) | |
Text
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VrHOINrA ELucTurc Axu Pow n H Co >t eAxy R rcrrMo:vo,Vinoan rA unust July 10, 1980 Mr. Harold R. Deaton, Director Serial No. 553/061880 Office of Nuclear Reactor Regulation LQA/WRM:rab Attn:
Mr. B. Joe Youngblood, Chief Licensing Branch No. 1 Docket No. 50-339 Division of Licensing U. S. Nuclear Regulatory Cocalission License No. NPF-7 Washington, D.C.
20555
Dear Mr. Denton:
We have received the request for additional information for North Anna 2 from Mr. B. Joe Youngblood, dated June 18, 1980, pertaining to achieving cold shutdown using safety grade equipment.
Enclosed are our responses to the questions received. If you have any questions, please contact this office.
I Very truly yours, y fr _ ljlo. ~,
hP
!h E. R. ' Sylvia thnager-Nuclear Operations and Maintenance Attachment THIS DOCUMENT CONTAINS POOR QUALITY PAGES.3 800i140/$N g
4 QUESTION 9 Question 5 in the referenced submittal included a com-parative analysis of natursl circulation capability between North Anna 2 and Diablo Canyon. The analysis as presented does not address upper head effects.
The concern is that inadequate cooling of the upper head may produce thermal stresses with adverse consequences, particularly around the vessel flange.
Show that the Diablo Canyon tests would be applicable to North Anna in addressing this concern.
RESPONSE
Natural, or free, convection is observed as a result of the motion of the fluid due to density changes arising from the heating process in the reactor core and the cooling process in the steam generators. The move-ment of the fluid in free convection results from the buoyancy forces imposed-on the fluid when its density in the proximity of the heat transfer surface is decreased as a result of the heating process.
The buoyancy forces would not be present if the fluid were not acted upon by some external force (i.e., gravity in this situation). As the tempera-ture of the-fluid in the core is increased relative to the temperature of the fluid in the steam generators, the resultant buoyancy forces will give rise to natural convection.
In the physical layout of the reactor coolant system the reactor core is at a lower elevation with respect to the steam generators; consequently, the higher temperature heat source is below the heat sink. This situa-tion assures that heat will be transported from the reactor core to the steam generators via the free convection heat transfer mechanism.
Thermal effects in the upper head of the reactor vassel will not result in a thermal / hydraulic impedance to loop natural circulation flow. Loop natural circulation flow is dependent on reactor core decay heat which
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is a function of time based on core power operating history.
Under natural circulation flow conditions, flow into the upper head area will constitute only a small percentage of the total core natural circulation flow and therefore will not result in an unacceptable thermal / hydraulic impedance to the natural circulation flow required to cool the core.
For typical 3-loop and 4-loop plants (including North Anna & Diablo Canyon) there are two potential flow paths by which flow crosses the upper head region boundary in a reactor. These paths are the head cool-ing spray nozzles, and the guide tubes.
For 4-loop plants equipped with upper head injection (UHI), the support columns are also a path. The head cooling spray nozzle is a flow path between the downcomer region and the upper head region. The temperature of the flow which enters the head via this path corresponds t'o the cold leg value (i.e. Teald).
Fluid may also be exchanged between the upper plenum region (i.e., the portion of the reactor between the upper core plate and the upper sup-port plate) and the upper head region via the guide tubes.
Guide tubes are dispersed in the upper plenum region from the center to the peri-phery. Because of the nonuniform pressure distribution at the upper core plate c' vation and the flow distribution in the upper plenum region, the pressure in the guide tube varies from location to loca-tion. These guide tube pressure variations create th'e potential for flow to either enter or exit the upper head region via the guide tubes.
To ascertain any difference between the upper head cooling capabilities between Diablo Canyon and North Anna, a comparison of the hydraulic resistance of the upper head regions was made.
These flow paths were considered in parallel to obtain the following results.
i North Anna Unit 2 Diablo Canyon Unit 1 2
Flow Area (ft )
0.67 0.77 Loss Coefficient 1.52 1.51 Overall Hydraulic Resistance 3.38 2.57 (ft-4),
Relative Head Region Flowrate 0.87 1.00 (Based on Hydraulic Resistance)
Head Region Flow Rate Relative 1.16 1.00 to Loop Flow As indicated above the effective hydraulic resistance to flow in North Anna Unit 2 is 1.3 times greater than Diablo Canyon. Assuming that the same pressure. differential existed in both plants the North Anna head -
flow rate would be 87% of the Diablo Canyon flow.
North Anna is a 3-loop plant and Diablo Canyon is 4-loop; therefore, in terms of f
relative portions of loop flow communicating with the head region, the North Anna head flow as a fraction of loop flow is 16 percent greater than the corresponding Diablo Canyon fraction.
In addition the overall mass of metal associated with the North Anna upper head is significantly less than for Diablo Canyon due to the smaller physical size.
Thus the upper head cooling capability at North Anna #2 would be no worse and would likely be better than demonstrated by the Diablo Canyon
- 1 testing.
A finite element stress analysis has been performed to evaluate whether the temperature gradient in the reactor vessel during natural circula-tion _ operations would limit cooldown due to vessel stress limit viola-tions.
The vessel geometry used in the finite element stress analysis was.a/ typical.4-loop vessel.
The stress analysis addressed the situation in which the bulk tempera-ture of the water in the head is much higher than the bulk temperature of the water below the head.
The water in the head is assumed to be virtually stagnant, and a low convection coefficient (h = 50 Btu /hr/
2 f t /oF) is used for the stress analysis. The *. ater below the head is subjected to a low flow rate caused by the test transient, so that it also has a small convection coefficient. A value of h = 250 Btu /hr/
2 ft /oF was used for this part of the vessel.
Using these two con-vection coefficients, two cases were investigated in the stress analysis:
Case 1: Case 1 evaluated the envelope condition where the metal tem-perature of the head is maintained at 550 F (no-load tempera-ture) while the water below the head is cooled from 5500F to 0
350 F in four hours.
Case 2:
Case 2 evaluated the condition.where the temperature of the water in the closure head cools from 550 F to -450 F while 1
the water below the closure head is cooled from 550 F to 350 F.
In both cases, it was found that the critically stressed components are the reactor vessel. closure studs. For both cases, it was also, found that the stress ranges in the studs were well below the allowable limit of 2.75 -
m Since the 2.75 code limit is satisfied,- the reactor vessel integrity m
will not be impaired when subjected to the envelope transient.
Due to' similarity between a typical 4-loop vessel and typical 3-loop vessel, results of the stress analysis for the 4-Ioop vessel are consi-
"dered to be representative of the 3-loop vessel under the stated assump-tions described in Case 1 and Case 2.
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OUESTION #10 The response to Question 1 has discuaced the air supplias for the safety grade atmospheric dump valves.
Identify any other valves needed to achieve or cain-tain cold shutdown which require an air supply.
Diccuss air supply failurcs in these valves.
Identify operator actions which would be taken and estimate times required to perform the actions. Discuss any time or environmental re-straints imposed by reactor conditica which might dictate action time require-ments.
. RESPONSE The Residual IIcat Removal System has two air operated control valves, FCV-1605 and HCV-1158, located inside the containment. The loss of instructent air will cause FCV-1605 to f ail closed and prevent. flow from bypassing the RHR heat ex-changer. The loss of instrument air will cause HCV-1753 to fail open and allows full flow through the RHR heat exchangers. (Ref. FSAR FIG 5.5.4-1).
L Component cooling water is supplied to each RHR heat exchanger by independent p
piping each having an air operated valve, TV-CC203A or TV-CC2038, 'lecated out-side the,.ontainment. The air operated valves function as Containment Isolation Valves and the loss of: instrument air will cause the valves to fall closed.
Post-accident radiation levels will not prevent restoring an air supply' to these valves.
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4 6
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' QUESTION //il Discuss qualification of the RRR nystem.
Show that all active components (including necessary instrumentation) in the systen are qualified to operate in the worst environment for scenarios identified in the reference submittal (moderate temperature,- 100% humidity). For example, the reference submittal and subsequent discussions have suggested that RHR suction isolation valve motor operators have been qualified for 100% humidity and at least 2500F, an environment which envelopes that for any of the postulated scenarios.
RESPONSE
The Residual Heat Removal System (RHRS) is designed to recove residual decay heat from the reactor core.
It is placed in operation approximately four hours af ter reactor shutdown when the temperature and pressure of the Reactor Coolant System (RCS) are approximately 3500F and 450 psig, respectively. The RRRS does not serve as an Engineered Safety Feature (ESF) at-North Anna as it
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may at other facilities. The North Anna RHRS is not a safety-grade system e:
since it is not required to operate to bring the plant to a safe shutdcun condition and does not serve an Emergency Core Cooling (ECC) function. A separate Low Head Safety -Injection (LHSI) system provides this function at North Anna.
The RRRS is located completely inside the reactor contaimaent building and the components are not adversely affected by the radiation, temperature and humid-ity levels normally present in the containment.
The -Residual Heat Removal System has been designed to the requirements defined
' in-the FSAR which do not designate this system as Class IE safety-related.
For this reason components within this system were not required to temain functional L-in the containment post-accident environment.
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L The basis for discussion of qualification of the RHR System is an environment of 250 F and 100% humidity' The enclosed lists of RHR components (both West-inghouse and Stone & Webster supplied items) give qualification data, as available, for each componene. This qualification data has been derived from vendor catalog information, specification requirements and test data.
Acutal test reports which exist for some of the specific components are not on. file at Stone & Webster for verification. Components for which qualification infor-mation is not available at this time has been noted.
The RHR System for North Anna Unit was not designed-for operability during all postulated accidents or to the safe shutdown design basis. Therefore, the system was not designed to meet all requirements of the single failure criteria.
This is readily apparent upon examination of the flow diagram (En-39A) for the RHR System. The extremely remote single failure of either MOV-2700 or MOV-2701
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could prevent initiation of RHR cooling in the normai manner. In the event of Aj such a failure the plant.would rcrain in a safe hot standby condition with heat removal via the steam generators, and operator action would be required to initiate the RHR System.
Should the RHR system become inoperable during cooldown, there exists other methods to remove decay heat. The method to be utilized would be determined at the time of the RHR system failure based on availability of equipment, ra-diation levels in the area where the equipment is located, and adequacy of water-supplies.
- The alternate methods to cool the reactor coolant system below 3500F should the : RHR system become inoperabic are the steam dump system to the main con-
- denser, atmospheric steam dumps, feeding and bleeding the steam generators.
4
r The cooldown rate when utilizing these alternate systems would be substant-ially less than S00F/hr, and cooldoun to less than 200 F would be difficult with the steam dumps, but feeding and bleeding the steata generators with main or auxilisry feeduater pumps, and blowdown systen could be used to reduce the teinperature to less than 200 F.
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t STONE f. k'EBSTER EQUTPMENT INDEX FOR THE RHR SYSTEM
!;i s t rern t Vendor
+'
Locction (Model)
Func tion Qualificntions
~ ~,-: ? '.-;?J A Inside Not Available RHR Pumps 2-RH-P-1A Motor TE is part of Motor Specification which is unavailable Cor.ta inmen t Stator Temperature Clement at this time.
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Ins ida Not Available RHR Pump 2-R!l-P-1B TE is part of Motor Specifiction which is unavailable Containment Motor Stator Temperature at this timo Element T~-T-:-2 0 !.'.-1 Inside Leeds & Northrup RIIR Pump 2-R11-P-1 A Containne.
Motor Upper Bearing Temperature tiement IT-M12111.-2 Ins ida Leeds & Worthrup RitR Pump 2-R!!-P-1 A Con t tinment Motor Lower Bearing Temperaturo Element "E-Rii2]13-1 Inside Leeds & Northrup R!!R Pump 2-R!!-P-1B Containment Motor l'pper Bearing Temperature Element II-1:! 010-2 Inside Leeds &'Northrup RHR Pump 2-R!!-P-!B i
Motor Lower Bearing Containment Tenperature Element s
WESTINGil0USl; EQUIPMENT IN32X FOR Tile Ri!R SYSTEM 1N3TR":!ENT VEI' DOR -
TAG 30.
LOCAT10tl (MODEL)
FUNCTION QUALIFICATIONS TE-2604 Inside
!!E TE-260!. indicates the Ri!R pumps Vastinghouse specification sheat 6.12 external prenure Containmont Sostman outlet temperature at a control of 60 psig for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Norwal environ:.nt'1C psis.
(Nic kel) room recorder and at the computer 120* F maximum, 60*F minimen. Post Accident initially pressure 42.5 psig, temperature 280*F TE-2606 Inside HE TE-2606 indicates the RH!t Heat Same Qualifications as for TE-2604 Con tainment Sostman Exchangers outlet temparature at (Nickel)
Control Room Recordar and at the computer PIC-2602 Inside Foxboro PIC-2602 alar.as in the control Information of Qualifications not known at p caint Containment (43-E)
Room 111 discharge pressure from the c irca Rt!R pumps PT-2403 Outside Rosemount PT-2403 indicates in the Control Westinghouse specification shact 4.9. Ear.a Containment 1152CP9A Roou the raaetor coolant pressurc Qualifications as for TE-2604. llydrostatic tested on the hot-leg piping of Icop I to 3750 psig also provides tha permissive signal for the RitR loop isolation valves interlock circuit (MOV-1701
?I-2600 Incide Ashcroft PI-2600, Local indication of Non-cicc e r ical. Non-instrurr.ent air, la fo r.:at ica containment
-(1079)
RI:R puup (1-nit-P-18) diceharge of Qualification not kura;. -:t p re r.en t t iu.
pres su ro TI-2601 Inside Ashc rof t PI-2601, local indication of Non-ele:: t rical. Non-instrument air. Inferuttaa c f Containment (1079)
Rt:R pump (1-R11-P-I A) discharge Qualification no known at present time, pressure F::-2605 Inside (Orifice)
R!!R discharge haader flow Man-olec trical. None-instrument air. Ir.fo rr.at ion Containment indication, control alarm and of Qualification not known at present timo, computar input k
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k'ESTINCHOUSE EQUIPMENT INDEX FOR TIC RHR $YSTEM INGTRUMEL'T.
VENDOR TAC NO.
LOCATION (MODEL)
EWCTION QUALIF2 CATION FT-2605 Inside Foxboro RitR discharge header flow WestingLcusa specification sheet 6.6.
Sace qu:lification Containment (E13DM) indication, control alarm and as TE-2604 computer input
,2605 Inside Continental Instrucent air required for Information of Qualification not known at preseat Conta incent (7613)
. valve operation time.
Actuatar Inside Fisher RitR discharge header flow Information of Qualification not known at present for FCV-2605
. Containment (472) indication, control alam and time.
(Teg-123A54R) computer input E/P-2505 Inside Fisher R!!R discharge haader flow FIS!!ER Cat' log 400 section 13.
Arbient a
Containment (546) indication, control alarm and temperature operating limits -40*F to +150*F computer input ECV-2758 Inside Continental RHR Heat Exe bange r outlet flow Westinghousa specification C76368 sheet 005 Containment (TY 7613) control. Instrument air required No Qualification stated.
for valve operation A:toctor Inside Fisher Rmt Ifeat E.cchanger outlet flow Information of Qualification not known at prasent for IICV-2758 Containment (472) control. Instrurent air required
- time, for valve operation Poeitior.or Inside Ffsher RHR HJat Exchanger outlet flow Information of Qualificatio not known at presa.n
~ f ar !!C"-2753 Containment (3570) control. Inctru:aant air required time.
for valve operation 4
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WESTINC1100SE EQUIPMENT IisDCX FOR Tl!E RhR SYSTEM 4
IW5TR'JME3T UENDOR TA0 NO.
LOCATION (MODEL)
FUNCTION OUALIFICATION E/P-2753 Inside Fisher R'lR lleat Exchanger outlet flow FISilER Catalog 400 section 13 Containment (546).
control. Instrument air required Ambient temperature operating limits for valve operation
-40*F to +150*F hCV-2720A Inside Limitorque Corp.
MOV_2720A isolates 2 x 107 R Containment Reliance Motor the RI!R system from the Temp 250*F with Class B Reactor Cooling System
!!umidity ICOI Insulation Test duration: 15 days Data Sheet dated 11-1-79 MOV-2720B Inside Limitorques Corp.
MOV-27203 isolates 2 x 107 Rads Containmcat Balance Class B the RhR system from Temp: 250*F Insulation ths Reactor Cooling I!umid it y: 100%
System Test duration: 15 days Limitorque Nuclear Qualification Deira Sheat dated 11/1/79 M07-2700 Inside Limitorque Corp.
MOV-2700 isolates Same Qualification as for MOV-2720B.
Containment Reliance Class B the Ri!R system from Isulation tho Reactor Cooling Systes MCV-2701 Insida Limitorque Corp.
MOV-2701 isalates Same Qualification as for MOV-27208.
Containment Relia nc 2 Class B the RIIR systeu frum the Insula. tion Reactor Cooling System 2-RH-P-1 A Insida Westinghouso *Clasa Long Term Core Cooling 2 x 103 Rads Camma Containment H Insulation Full speed in 75 sec.
Temp:
40*C Pressure: Atmosphcro Ilumit idy: naae Qualification Reference VCAP-8754*
2-R11-P-15 Inside Westinghouse Long Term Core Coaling 2 x 108 Rads Camma Containment
- Class 11 Full speed in 75 scc Insulation Temp: 40*C Preasure: Atmosphere ilumidity: none Qualification Raference WCAP-8754 4 0-Raference Westiaghouse generic testing on Class II. motor insulation 4
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W5 WCST1!;CI:00SE CQUIPME ff INDEX FOR I!!C RIIR SYSTdM
, (Bue)
IN3 fC:O.T VENDOR TAC No.
LOCATION (M002L)
FUNCf10N QUALIFICAfION IKV Power Cable Ins ide Ceneral Supply power to RHR pump motors Sco NUREC-05S8 Containment Cable Review Pago 10 Electric Ins ida Conax Supply power to Safety Systmes See NUREC-05SS Penetration Conta ir. mon t Corporation Revi.:w pages 24 thru 27 600V Cu Powar Inside Okonite Cabic Supply power to Safety Systems See NUREC 0533 Review Pa3es 6 C-ble Con tainment 3)CV 2nstrument Inside Boston Insulated Supply Power to Safety Systems See NUREC 0588 Review pages 9 Crala Containment Wired & Cable 0
4
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' QUESTION 12:
Identify the control room alarm (s) which is (are) provided to alert the operator to a losu of flow in the RHR System.
Discuss the scenarios of a spu rious ' closing of either RHR suction isolation valve or the RHR flow control valve from event initiation to RHR flow restoration.
RESPONSE
There is a low flow alarm on the RHR System. - This alarm is initiated by a flow element in the conbined return upstream of MOV-2720A and MOV-2720B.
When flow is reduced to less than 3200 gpm, an audible and visual alarm is initiated in the control room to alert the operator of this abnormal condition.
Action which the operator must take is outlined in an approved procedure.
In the event the RHR suction valve shot?.d spuriously close, a low flow alarm would be initiated to alert the operator of this condition. The operator will take the appropriate action to either reopen the suction valve manually or shut down the RRR pump until the suction valve can be reopened.
When in MODE 4 or MODE 5 the North Anna 2 Technical Specifications allow the reactor coolant pumps and the residual heat removal pumps to be deenergized for up to one hour provided no operations are permitted which could cause dilution of the reactor coolant system boron concentration.
In the event the RHR flow control valve should spuriously close, a low flow alarm would be initiated to alert the operator of the condition.
Operator
. action would be tak'en to open the flou control valve bypass (HCV-1758). This
- action would -increase the reactor coolant system cooldown rate; however, the cooldown rate,could be controlled by manually throttling component cooling to the RHR. heat exchangers until' repairs could be made. to the flow control valve.