ML20213C822

Forwards Request for Addl Info Re FSAR in Areas of Missile Protection,Protection Against Postulated Piping Rupture,Fuel Handling Facility,Ultimate Heat Sink Equipment & Floor Drainage Sys

ML20213C822
Person / Time
Site:	Columbia
Issue date:	12/04/1978
From:	Matthews P Office of Nuclear Reactor Regulation
To:	Varga S Office of Nuclear Reactor Regulation
References
CON-WNP-0242, CON-WNP-242 NUDOCS 7812130325
Download: ML20213C822 (20)

Text

{{#Wiki_filter:* y I DISTRIBUTI0ft: DOCKET FILE [ g 3g /' IIRR READIrlG ASB READIrlG + f*0 _77 Y Docket !!ol 50-496* (%. / N j liEMORAilDUM FOR: S. A. Varga, Chief, Light Water Reactors Branch 4, DPM FROM: P. Matthews, Acting Chief, Auxiliary Systems Branch, DSS l

SUBJECT:

WASHINGTON PUBLIC POWER SUPPLY SYSTEM, UflIT 2 i Plant Name: WPPSS, Unit 2 Docket Number: 5'UNQ6 50g ~ Licensing Stage: OL i fillestone liumber: 05-02 Responsible Branch: LWR-4 l Project Manager: D. Lynch . Requested Completion Date: December 1, 1978 Review Status: Complete The enclosed first round frequest for additional infomation covers .those sections of the WPPSS-2 FSAR for which the Auxiliary Systems Branch has primary responsibility. Our review is based on the HPPSS-2 i FSAR up to and including Amendment 1. The enclosure identifies areas where we need additional infomation. These areas include missile protection, protection against postulated rupture of piping, fuel handling facilities, ultimate heat sink, equipment and floor drainage system, the main steam isolation valve leakage control system and the heating ventilation and air condition l systems. l The applicant's submittal of their fire protection program is currently j under review. We will provide our request for additional information in this area at a later date. i i i l P. Matthews, Acting Branch Chief Auxiliary Systems Branch Division of Systems Safety

Enclosure:

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~ 6 ~, /Ipn *Ec,*% u UNITED STATES 8,",,.,,, NUCLEAR REGULATORY COMMl3SION .E W ASHINGTON. D. C. 20555 g. -[,/ Docket No: 50-396 MEMORANDUM FOR: S. A. Varga, Chief, Light Water Reactors Branch 4, DPM FROM: P. Matthews, Acting Chief, Auxiliary Systems Branch, DSS

SUBJECT:

WASHINGTON PUBLIC POWER SUPPLY SYSTEM, UNIT 2 Plant Name: WPPSS, Unit 2 Docket Number: 50-396 Licensing Stage: 0L Milestone Number: 05-02 Responsible Branch: LWR-4 Project Manager: D. Lynch Requested Completion Date: December 1, 1978 Review Status: Complete The enclosed first round request for additional information covers those sections of the WPPSS-2 FSAR for which the Auxiliary Systems Branch has primary responsibility. Our review is based on the WPPSS-2 FSAR up to and including Amendment 1. The enclosure identifies areas where we need additional information. These areas include missile protection, prctection against postulated rupture of piping, fuel handling facilities, ultimate heat sink, equipment and floor drainage system, the main steam isolation valve leakage control system and the heating ventilation and air condition systems. The applicant's submittal of their fire protection program is currently under review. We will prcvide our request for additional information in this area at a later date. /k-..,: l y l' b ff IW P. Matthews, Acting Branch Chief Auxiliary Systems Brcnch Division of Systems Safety

Enclosure:

As stated cc: See next page

Contact:

P. Hearn x27763

S. A. Varga DEC 4 'i cc: S. Hanauer R. Mattson R. Boyd V. Benaroya D. Vassallo W. Pike D. Lynch C. Long D. Fischer P. Matthews P. Hearn i 1 l l l i l l l l l

Auxiliary Systems Branch First Round Questions Washington Public Power System Unit 2 Docket No. 50-396 010.10 Verify that all piping and electrical penetrations in safety related (3.4.1) structures that are below the PMF level are water tight. 010.11 Provide the details of your evaluation for the jet impingement and (3.6) environmental effects resulting from postulate piping failure of the main steam and main feedwater systems. We require that compartments which house the main steam lines and feedwater lines and the isolation valves for those lines, be designed to consider the environmental effects (pressure, tempara-ture, humidity) and potential flooding consequences from an assumed crack, equivalent to the flow area of a single ended pipe rupture in 'hese lines. We require that essential equipment located within t j the compartment, including the main steam isolation and feedwater valves and their operators be capable of operating in the environ-I ment resulting from the above crack. We also require that if this ) assumed crack could cause the structural failure of this compart-9

2 ment, then the compartment failure should not jeopardize the safe shutdcwn of the plant. In addition, we require that the remaining portion of the pipe in the tunnel between the safety valve house and the turbine building meet the guidelines of Branch Technical Position ASB 3-1. We require that you submit a subcompartment pressure analysis to confim that the design of the pipe tunnel confoms to our position as outlined above. W request that you evaluate the design against this staff position, and advise us as to the outcome of your review, including any design changes which may be required. The evaluation should include a verification that the methods used to calculate the pressure build-up in the subccmpartments outside of the containment for postulated breaks are the same as those used for subcompartments inside the containment. Also, the allowance for structural design margins j (pressure) should be the same. If different methods are used, ~ justify that they provide) edequate design margins and identify the margins that are available. When you submit the results of your evaluation, identify the computer codes used, the assunpttions used for mass and energy release rates, and sufficient design data so that we may perform independent calculations. l l

s The peak pressures and temperatures resulting from the postulated break of a high energy pipe located in compartments or buildings is dependent on the mass and energy flows during the time of the break. You have not provided the information necessary to deter-mine what terminates the blowdown or to detennine the length of time blowdown exists. For each pipe break or leakage crack analyzed, provide the total blowdown time and the mechanism used to terminate or limit the blowdown time of flow so that the environ-mental effects will not affect safe shutdown of the facility. 010.12 According to FSAR Section 3.6.1.1.1, fluid piping systems whicn are (3.6) classified as high energy, whose temperatures are below 200*F and whose pressure is derived by a centrifugal pump instead of a fluid reservoir are considered moderate energy systems. Justify that these systems do not contain enough energy to generate pipa whips and justify performing a flooding analysis based on the moderate energy crack criteria instead of the full break of the high-energy break criteria. 010.'3 It is our position that the flain Steam Isolatiun Valve Leekage i l (6.7) (RSP) Control System meet R.G.1.96. l l l N

In particular: a) the MSIV-LCS is designed to permit actuation within 20 minut'es. b) the MSIV stem leakage system is designed to the same standards as the MSIV-LCS. c) operation of the MSIV-LCS during normal plant operation is prevented by interlocks capable of functioning after a single failure in the interlocking system. Modify your plant to meet these positions. 010.14 Identify safety related equipment that could be exposed to dust storms. (9.0) Describe the means of assuring the functioning of this equipment during dust storms. Include in your response the description of the methcds used to prevent the blockage of vital air supplies to safety related equipment. 010.15 You provided your spent fuel rack design which in-(9.1.2) cludes a neutron absorber material sealed between stainless steel plates. Recent experience at operating reactors has shown that auckling of tne stainless steel may occur duie i.o gasses given off by the irradiation of the neutron absorber material, thus resulting in stuck fuel assemblies. Describe the methods used to prevent this from occurring such as venting the plates to release gases. 010.16 In Section 9.1.2 of your FSAR you list the results of radiation, (9.1.2) thermal, seismic and baron testing of the B C plates. Describe i 4 the procedures used for these tests or reference where these pro-cedures have been accepted by NRC on a previous docket.

% 010.17 According to Section 9.1.2.3.3 of your FSAR the interlocks that (9.1.2) (RSP) prevent the 125 ton Reactor Butiding Crane from traversing the spent fuel pool are occasionally bypassed. Traversing the spent fuel pool with the Reactor Building Crane would result in damage to the stored fuel in the event a load is accidently dropped. It is our position that the Reactor Building Crane interlock.s remain in place when there is spent fuel in the fuel pool. Modify your pro-cedures to meet this position. 010.18 According to Secticn 9.1.2 of your FSAR a portion of the Fuel Handl-(9.1.2) (RSP) ing Building above the refueling floor is constructed of sheet metal. It is our position that the spent fuel pool be housed in a struc-ture that is Seismic Cate' gory I and tornado missile protected. De-monstrate that your plant meets this position. 010.19 Provide a Seismic Category I spent fuel cooling system and a Seis-(9.1.3) mic Category I makeup water source to the spent fuel pool per the guidelines of Regulatory Guide 1.13.

. 010.2 0 Identify the valves used to isolate the non-seismic portion of the (9.2.1) plant service water system from the seismic portion. Provide a failure modes and effects analysis for the system assuming a seis-mic event has occurred. 010.21 In order to permit an assessment of the Ultimate Heat Sink, provide (9.2.5) (RSP) the'results of an analysis of the thirty-day period fcllowing a design basis accident that determines the total heat rejected, the sensible heat rejected, the station auxiliary system heat rejected, and the decay heat release from the reactor. In submitting the results of the analysis requested, include the following information in both tabular and graphical presentations: (1) The total integrated decay heat. (2) The heat rejection rate and integrated heat rejected by the station auxiliary systems, including all operating pumps, ventilation equipment, diesels and other heat sources. (3) The heat rejection rate and integrated heat rejected due to sensible heat removed from containment and the primary system. (4) The total integrated heat rejected due to the above.

~ 1 i. (5) The maximum allowable inlet water temperature taking into account the rate at which the heat energy must be removed, cooling water flow rate, and the capabilities of the respective heat i <. changers. (6) The required and available NPSH to the service water pumps at the minimum Ultimate Heat Sink water level vs. the required NPSH. The above analysis, including pertinent backup information, should demonstrate the capability to provide adequate water inventory and provide sufficient heat dissipation to limit essential cooling water operating temperatures within the design ranges of system components. Use the methods set forth in the enclosed Branch Technical Position ASB 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling," (enclosure 2) to establish the input due to fission produce decay and heavy e16 ment decay. Assume an initial cooling water temperature based on the most adverse conditions for normal operations. Assume the meteoroological conditions set forth in Position 1 of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Power Plants," Revision 1, dated March 1974

% 010.22 Because the sprays in the Ultimate Heat Sink are required for safely (9.2.5) (RSP) shutting down the plant, it is our position that the sprays be pro-tected from the effects of tornadoes and tornadoe missiles. Modify your plant to meet this positior.. 010.23 If a tornado syphons water from the Ultimate Heat Sink (UHS), the make (9.2.5) up water pumps will replenish tre UHS. Transformers that are required to operate the makeup water pumps are located in the Turbine Building. Verify that these transformers and the cables necessary to operate the make up pumps are protected from tcrnadoes and tornado missiles. 010.24 According to Section 9.2.5 the two ponds of the Ultimate Heat Sink (9.2.5) are connected by a syphon that allows water to flow from one pond to the other. Demonstrate that a failure in the syphon line or one of the ponds will not result in the draining of both ponds. 010.25 Because the standby service water system is required for the safe (9.2.7) (RSP) shutdown of the plant, it is our position that the standby service water system be protected from tornado missiles. 010.25 Describe the means of preventing flooding of safety related equipment (9.3.4) due to back flooding through the equipment and floor drainage system. Provide information that demonstrates that portions of the system necessary to prevent backflooding such as check valves must be seismic Category I and system function should be maintained assuming a single active failure.

u m .g. 010.27 Verify that all engineered safety features heating ventilation and (9.4.0) air conditioning systems are protected from tornado missiles. 010.28 According to your FSAR the outdoor design temperature range for the (9.4.0) HVAC systens is 0*F to 105'F while the extreme outdoor temperature range is -27'F to ll5'F. Provide the results of analysis that verifies that the functional capability of safety related equipment will not be jeopardized due to the indoor temperatures resulting from these extreme outdoor conditions. 010. 29 The non-seismic Radwaste Building Chilled Water System is connected (9.4.4) to the Control Room HVAC System and the standby service water sys-tem. Provide an analysis that demonstrates that failure of the non-seismic Radwaste Building Chilled Water System during an earth-quake will not cause unacceptable degradation of the Control Room HVAC system and the standby service water system. 010. 30 Verify that the Diesel Generator Fuel Oil Pump Room ventilation (9.4.7) system is Seismic Category I and powered from the emergency power busses. l

10-010.31 Provide an analysis that demonstrates the failure of the non-seismic (9.4.7) heaters in the Diesel Generator HVAC system will not have an adverse effect on the functional capability of the Diesel Generator's or the Diesel Generator HVAC System. 010.32 Your response to our question 010.9 is unacceptable. Your analysis (10.4.5) of flooding due to failure of the circulating water system is based on a crack area of 1/2 pipe thickness of 1/2 pipe diameter. Provide an analysis of flooding due to failure of the Circulating Water Sys-tem that is based on the failure of the expansion joint, whereby you assumed that this results in a guillotine type break in the circulat-ing water system.

~ E*A*.Ori Tice:;; CAL F05:T 0?; '5B 9 2 l RESII'tL DECAY E!.EE3Y FOR LISHI WATER ~ REACTORS FOR L7GTEP3 COOLING A. Ev7U';0 The 'exiliary Syste s Ertr:r. *as :e.el:;ec a :.e:tacle essn-iv s anc fc Naticms l [ that ay be used to calculate the resid 41 decay energy release rate fer ligt.: water ccoled reacters for Icng-term ccoling of the reactor #acility. Esperi. ental data (Refs. I and 2) cn tetal beta anc garra er.ergy releases fer long half, life (> 60 seconds) fissien ;ronets from ther al neutren fission of !; 235 have teen 3 9 considered reliable for decay tiets of 10 to 10' seconds. Over tnis decay tire, even with the exclusion of short-lived fissien products, the decay heat rate can te predicted to within 10 percent of excerimental data (Refs. 3. 7. and 8). The short-lived fission prcducts contribute a;;reciably to the decay energy for decay 3 times less than 10 seconds. Although consistent excerirental data are not as nu ercus (Refs. 4 and 5) and the results of various calcula:1cns differ, t*e effect of all uncer- ~ 4 3 tainties can be treated in the zero to 10 second time range by a suitably conservative multiplying factor. B. BRANCH TECHNICAL POSITION 1. Fission Product Cecay For finite reactor operating time (t,) the fraction of operating power, h, (t,. t,). to be used for the fission product decay power at a tire t, after shutdown ray be calculated as follows: n=11 ht A exp(-a t,) (1) h,(=.t,) = n n nel . =e ,h,(t,t,) (1 + K) h,(=t,)-h,(=.t,+t,) (2) = 1 where: P To fraction of operating power = t, cumulative reactor operating titre, seconds a e, time after shutdown seconds = 3 I 3 t, i uncertainty factor; 0.2 for o i t, i 10 and 0.1 for 10 10 K = A.a fit coefficients having the following values: a n n

9. 2. 5 -3a Rev. 1

..s n 'W 5 a T., 3 1 0.i733 !.77~ > 10 f.774 x 10'I 2 1.6500 3 3.1000 6.?"3 a 10-2 6.2!4 x 10'3 4 3 E700 .739 x 10*# 5 2.33 0 *- 4.810 x 10 5 6 1.2903 5.I"4 x 10-6 7 0.*620 5.716 x 10*I E 0.3'30 1.036 x 10"I 9 0.1700 2.959 x 10'8 10 0.C?55 7.5SS x 10-10 11 0.1140 The expressions for finite reactor cperation esy be used to esteu! ate the decay energy from a complex operating history; however, in accident analysis a suitably conservative history should be used. For example, end of first cycle calculations should assume continuous operation at full power for a full cycle time period, and end of equilibrium cycle calculations should assume appropriate fractions of the core to have operated continuously for multiple cycle times. An operating nistory of 16,000 hours is considered to te representative of many end of-first or equilibrium cycle conditions and is, therefore, acceptable. In calculating the fission produce decay energy, a 20 percent uncertain,ty factor (K) should be added for any cooling time less than 10 seconds, and a factor of 10 percent should be added for cooling 3 7 times greater than 10 but less than 10 seconds. 2. Heavy Element Cecay Heat The decay heat generation due to the heavy ele-ents U-239 and N -239 may te calculated p according to the following empressions (Ref 6): -(1-exp(4.91x10'"t))[exp(-4.91x10~4 tIl I3I 25 P (U 239_) = 2.28 x 10'3 C o s Po f25 M N 230, ;,37, jg.3 g '25 [0.007 (1 - esp (-4.91 x 10'4 t.)] Po 'f25 l . (exp(-3.41 x 10~0 t,) - exp(-4.91 x 10~4 t,)] 'I +(1-exp(-3.41x10-6 t))(exp(-3.41x10'0 t,)]F (4) o s u-9.2.5 9 11/24/75

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P C. t#Ti:E*.;E! 1 J. F. "erkins and P. W. v.ing. " Energy Release Frc., the "e:ay of Fiss'cn "ro:a:ts. N a:! ear Scier.:e and Enc,inserir.;." Vol. 3, 7~5 (1355). 2. A. M. "erry F. C. Matenschein, anc D. R. Vondy. "Fissi:n-Prods:: Aft:- est: A *e.-iew cf ti;erizants Fertinent to the Ther s!

• eutr:n Fission of ;I!." C3'a *:*- 197. Oak U
• f d;e ?.sti:nal Lat:-story. Oct:ter 1973.

3. A. T:of as. ",*t e Enarg/ Release From Pissicn Products " Jo.rral of *:sciese Energy, Vol.27,725(1973). 4 J. 5:stie. R. 3. Scott, and H. W. Wilsen. "Esta Ertegy Release 'Olit ing 19e it.er al

• aute:n':r.duced Fissi:n of !??v' and ;!!." Joar al cf Nuc* ear Er.er;y, Vol. 25,1 U

(1971). 5. L. Costa and R. de fourreil. "Activite e et 2 res Prodscts d'une Fission de "!!U et !!'Pu.* Jcurnal of Nuclear Energy. Vol. 25, 255 (1971). 6. Freecsed ANS Stardard. "0ecay Ener;y Release Rates Fc11cwing !butd:wn of Uranium - Fueled *tereal Reactors." American Nuclear Society. October 1973. 7. J.,5cebte and R. D. Scott " Calculation of Seta Energy Release Rates Fo11cwing Thermal Neutron Induced Fission of 231g,135 21'Pu and "'*IPu.", Journal of Nuclear U Energy, Vol. 25,339(1971). 8. K. Shure. " Fission Product Cecay Energy." WAPD Bi-24. 'a'estinghouse Electric Carperation. Cecember 1961. 9.2.5 14 11/24/75 .}}