ML19326D101
| ML19326D101 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 02/03/1976 |
| From: | Bauman R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Boyd R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8006060625 | |
| Download: ML19326D101 (29) | |
Text
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(TEMPOf?ARY FORM) 1 g , , CONTROL NO: __ FILE: FROM: Consumers Pwr Co DATE OF DOC DATE HEC'D LTR TWX RPT OTHER Jackson, Micnigan K C Bauman 2-3-7e 2-5-7e XXX ~ TO: ORIG CC OTHER SENT URC PDR- XX uL one signed SENT LOCAL PDR XX ,_ CLASS UNCLASS PROPINTO INPUT NO CYS REC'D dor SET NO: XXLXX 1 c o-32, 30 DESCRIPTION: ENCLOSURES: Lor crans t.ne following: Auul into concerning Reg Gulues in respons to ouc 1-1.$- to ltr. . . . .provic Ang Anfo on ECCS THIS DOCUMENT CONTAINS POOR QUAUTY PAGES
. PLANT N AME: Midland 1 & 2 SAFETY FOR ACTION /INFC RMATION v emo 2-10-76 ehf Ass)CNED AD ASSICXCD BRA'.:CII ClllEF
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O 8"- -.I e ( , P05'IST e Com,p,,u,,,,In. s2 y o.m.,.i om..: M West Michigan Avenue. Jackson, Michigan @01. Area Code 517788 0550 February 3,1976
' lth.,
- r. . g Director of Nuclear Reactor Regulation ")
Attenticn: Mr Ro6er Boyd, Director Division of Project Management FEB9 1976 -1 *k,)) U. S. Nuclear Regulatory Cecnission I .
.. D Washington, DC 20555 A Ey^ y$gmums
?h- l MIDLAllD PROJECT
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DOCEET NUIGERS 50-329, 50-330 'M. REGULATORY GUIDE IMPLEMENTATION FILE: 0505 SERIAL: 2123 The enclosed information responda to Mr A. Schwencer's January 13, 1976 letter requesting additional information on implementation of 10 miscellaneous Re6-ulatory Guides and the emer6ency cooling system for the Midland Plant. The Regulatory Guides were discussed with your staff on December 19, 1975 1 1 ~ R. C. Bauman / Project Engineer ;. RCB/lc 3, C,
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000.0 GENERAL 000.1 Based on our discussions of December 19, 1975, we understand that your design will conform with regulatory guides 1.1 (NPSH for ECCS pumps - 11/2/70), 1.7 (Control of Combustible Gas In containment 3/10/71) and 1.49 (Power Levels).
Response
Regulatory Guide 1.1: During the injection mode of ECCS operation the Regulatory Guide is met. During recirculation it is assumed that the vapor. pressure at the pump is equal to the containment pressure; thus, the available NPSH equals the elevation head minus the head associated with friction losses and is not a function of pressure or temperature. The above description.provides the calculational assumptions which are used for Midland. The applicant believes this approach meets the intent of Regulatory Guide 1.1 although not the specific letter of this guide. This position was accepted by the AEC staff in its DRL safety evaluation of the Midland Plant and in the Staff's testimony, page 2454, Staff exhibit No. 4 of the Construction Permit ASLB Procedings.
. Regulatory Guide 1.7: The project will comply with the Regulatory Guide and will use NRC Branch Technical Position Paper CSB6-2 in determining hydrogen gas generation.
Regulatory Guide 1.49: The project will comply with Sections C.1, and C.3 of Regulatory Guide 1.49. For Section C.2 B&W will use an appropriate alternative which will be described in the FSAR. 4 k 1
p q 000.3 Regulatory Guide 1.70 Rev. 2 (Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants) Your letter of November 7, 1975 indicates that the degree of conformance . to Regulatory Guide 1.70 remains u '.efined with regard to analyses for the Safety Analysis Report. Please define the alternatives you wish to follow and provide the basis for such alternatives.
Response
The subject of this question was the topic of discussion with the NRC management during the January 13, 1976 Midland FSAR meeting. The applicant's position and specific alternatives will be submitted under separate cover letter. f l-2
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010.0' EFFLUENT TREATMENT SYSTEMS BRANCH 010.1 Regulatory Guide 1.52 (Design, Testing, and Maintenance Criteria (6.5) for Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants 6/73) Regulatory Guide 1.52 provides guidelines for the design of ESF air filtration systems. Identify your ESF air filtration systems and provide the volumetric flow rate and adsorption bed depth for each system. You indicate that your design will not meet several positions of Regulatory Guide 1.52. Justify your design with regard to the following recommendations of Regulatory Guide 1.52: C.2.a You should state that you will conform to the Guide. C.2.c The fuel storage facility ventilation and filtration systems should be designated as seismic Category I. C.2.j Filtration systems should be totally enclosed and installed in a manner which permits replacement of the train as a minimum number of segmented sections without removal.of individual components. C.3.b You should state you will conform to the Guide. C. 3.j The design of the adsorber section should consider possible iodine desorption and adsorbent autoignition that may result from radioactivity induced heat in the adsorbent and concomitant temperature rise. Acceptable designs include a low flow air bleed system, cooling coils for the adsorber section, or other cooling mechanisms. The system design should provide for fire protection to inhibit adsorber fires. Combustible gas that may be generated by an adsorbent fire should be considered in the design. C.4.d Replaceable components should be spaced five linear feet from mounting frame to mounting frame.
Response
The engineering safeguard air filtration system for the Midland Plant consists of the following:
- 1. Main Control Room Makeup Air Filtration Units Volumetric Flow Rate = 2,000 cfm Adsorption Bed Depth = 4" 3
010.1 c:nt'd I; )
- 2. Main Control Room Recirculation Air Filtration Units Volumetric Flow Rate = 4,000 cfm Adsorption Bed Depth = 4"
- 3. Emergency Fuel Handling Area Exhaust Air Filtration Units Volumetric Flow Rate = 18,000 cfm Adsorption Bed Depth = 4"-
C.2.a Atmosphere cleanup systems designed and installed for the purpose of mitigating accident doses are redundant and comply with the requirements of Regulatory Position C.2.a, except that-demisters are not provided because water droplets are not expected to be entrained in the air stream. C.2.c The fuel storage facility ventilation and filtration systems will be designed as seismic Category I systems. C.2.j It is envisioned that workers will not handle filter units immediately after a design basis accident and will thereby avoid exposures associated with immediate post accident filter handling. All filter components exospt fan are enclosed in a unitized housing. In addition to this, filter housings and associated fana are located in proximity of outside walls or roofs. No other specific provisions have
, been made for removal of the train following an accident.
C.3.b We intend to comply with the requirements of this section. C.3.j Compliance with this section of the Regulatory Guide is provided by designing the adsorber for an iodine loading not to exceed 2.5 mg of iodine per gram of charcoal. The anticipated temperature rise due to radioactively-induced heating of the adsorber is insignificant; therefore, iodine desorption or adsorbent autoignition will not occur. Consequently, provisions for a charcoal dousing system are not considered in the design. However, the following two protective measures are included in the design. 4
f.; :~ 010.1 ctnt'd , , 1. Two seismically qualified thermostatic devices are included in the design: one installed on the upstream side of the prefilter to measure the air temperature entering the filter unit and the second located in the adsorber section, to warn of a temperature buildup above ambient with a setpoint appreciably below the charcoal ignition temperature. Upon a high temperature alarm, the appropriate action to institute full air flow through adsorber section, can be initiated from the control room.
- 2. In addition to the design provision described
-above, non seismic local fire hoses are provided in a separate room adjacent to the air filtration units.
C.4.d We intend to comply with this requirement. 4 + 5
3 020.0 AUXILIARY AND POWER CONVERSION SYSTEMS BRANCH 020.1 Regulatory Guide 1.27 (Ultimate Heat Sink for Nuclear Power (9.2.5) Plants) In order to permit an evaluation of the ultimate heat sink and other heat removal systems, provide an analysis of the thirty-day period following a design basis accident listing 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.
- 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 exchangers.
- 6. The 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 infe mation, should demonstrate the capability of the ultimate heat sink to provide adequate water inventory and provide sufficient heat dissipation for the safe shutdown and cooldown of both units following a LOCA in one unit. Use the methods set forth in the enclosed Branch Technical Position APCSB 9-2, " Residual Decay Energy for Light Water Reactors for Long Term Cooling," to establish the input due to fission product decay and' heavy element decay. Assume an initial service water temperature based on the most adverse conditions for normal operation. 6
020.1 ccnt'd ' '
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Response
The decay heat generation rate for Midland is presented in Figure 020.1-1 and-Table 020.1-1. The integrated decay heat is also tabulated in Table 020.1-1. Due to the unknowns in the station auxiliary loads at the time the heat loads were compiled in 1973, the decay heat generation was very conservatively calculated based on infinite operation at 106% of the ultimate licensable core power level of 2,552 MWe ; i.e., at 2,705 MWt . This value represents 110.3% of the proposed plant licensed power level of 2,452 MWe or 8.3% higher than that recommended by Regulatory Guide 1.49. In addition to the fission product decay energy, heavy element decay energy and residual heat due to delayed fissions are also included. Fission of Shurep(roduct decay energy l) as acknowledged is calculated in the proposed using the equations ANS standard (2), In addition, twenty percent and ten percent margins are con-servatively added to fission product decay energy and heavy element decay, respectively, for the full 30 day range of the calculation. This method is more conservative than that set forth~in Branch Technical Position APCSB 9.2 and has been used to account for loads unknown at the time of the analysis. The final analysis done for the FSAR will include the station auxiliaries and the method set forth in Branch Technical Position APCSB 9.2. The heat rejection rate from the containment and primary system of the unit with the postulated design basis accident is shown in Figure 020.1-2 and Table 020.1-2. The corresponding integrated heat rejection is shown in Figure 020.1-5 and Table 020.1-2. Both redundant trains of safety related equipment are conservatively assumed to be operating as designed to yield the highest heat rejection rate to the pond. For the safe shutdown unit, normal cooldown is commenced with initiation of the auxiliary feedwater system. From the time when cooldown is initiated until a reactor coolant system temperature is. reduced to 2800F, the entire primary system i energy and decay heat from the safe shutdown unit is rejected by dumping steam from the steam generators to'the atmosphere. After 6 hours when the reactor coolant system reaches 2800F, both trains of the decay heat removal system are started to complete the cooldown. This final cooldown phase includes the decay heat and the sensible heat remaining in the primary system.
'The heat rejection rate from the safe shutdown unit is shown in Figure 020.1-3 and Table 020.1-3. The corresponding inte- ,
grated heat rejected is presented in Figure 020.1-5 and Table ) 020.1-3. The total heat rejection rate from both units is l 1 t . - .
02L.1 cont'd . shown in Figure 020.1-4 and Table 020.1-4. The stepwise approximation of the total heat rejection rate used in the ultimate heat sink analysis is also shown in Figure 020.1-4. The total heat removed from both units is shown in Figure 020.1-6 and Table 020.1-4. The total heat rejected to the cooling pond in the 30 day period is 4.95 x 1010 BTU. The maximum calculated water temperature of the ultimate heat sink is 108.5 F based on initial pond temperature of 96 F and heat loads described previously. The design ratings of the major heat removal equipment are listed below,
- a. Decay Heat Removal Heat Exchanger:
Heat transfer capacity = 90 x 10 6BTU /hr. Tube side Flow = 1.5 x 106 lb/hr. Temperature in = 230 F Shell side Flow = 1.5 x 106 lb/hr. Temperature in = 105 F
- b. Component Cooling Heat Exchanger 6
Heattransfercapacity=lg2.5x10 BTU /hr. Tube side Flow = 3 x 10 lb/hr. Temperature in = 105 F 6 Shell side Flow = 1.535 x 10 lb/hr. Temperature in = 187 F
- c. Reactor Building Air Cooler:
Heat transfer capacity: 50 x 106 BTU /hr. Water side Flow = 0.36 x 106 lb/hr. Temperature in = 100 F Air side Flow = 36,500 CFM Temperature in = 300 F The design heat transfer capacity for component cooling water heat exchangers and decay heat removal heat exchangers is based on other than DBA heat loads. Therefore an increased inlet temperature will not negate the capability of the heat exchangers to perform a plant safe shutdown through removal of decay heat. Reactor building air cooler design heat transfer capacity is based on DBA conditions. The air cooler rating is i based on controlling, in concert with the containment sprays, the short term containment pressure temperature response. However, the short term response is not affected by the later peak temperature obtained in the ultimate heat sink. Service water pumps are mixed flow type, single stage wet pit pumps. The available NPSH for the service water pumps at the f beginning of the 30 day period, assuming loss of the dike surrounding the main pond is 38'. The NPSH available for the 8
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service water pumps at-the end of the 30 day period considering pond evaporation and seepage is 36'-7". The required NPSH - for the flowrates associated with a safe shutdown of one unit and a DBA in the other unit is 32 feet. Thus a minimum NPSH margin of 4'-7" exists. 4 i I t 4 C ( 1 l L' l 9
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References
.1. K. Shure, " Fission Product Decay Energy," WAPD-BT-24, Westinghouse Electric Corporation, December 1961.
- 2. Proposed ANS. Standard, " Decay Energy Release Rates Following Shutdown of Uranium Fuelad, Thermal Reactors," American Nuclear-Society, October 1973.
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TABLE 020.1-1: Decay Heat Generation Rate and Integrated Decay Heat for Each Unit, Assuming Infinite Operation at 106% of 2552 Wt Time Decay heat generation Integrated decay heat (Sec) rate for each unit for each unit (BTU /hr.) (BTU) 0.1 2.44 x 109 1 2.24 x 109 6.50 x 105 10 1.45 x 109 5.78 x 106 30 7.18 x 108 1.18 x 107 60 4.68 x 108 1.68 x 107 70 4.44 x 108 1.80 x 107 80 4.25 x 108 1.92 x 107 90 4.12 x 108 2.04 x 107 100 4.03 x 108 2.15 x 107 200 3.33 x 108 3.17 x 107 300 3.18 x 108 4.08 x 107 400 2.95 x 108 4.93 x 107 500 2.79 x 108 5.72 x 107 600- 2.66 x 108 6.48 x 107 700 2.56 x 108 7.21 x 107 800 2.47 x 108 7.91 x 107 900 2.40 x 108 8.58 x 107 1,000 2.34 x 108 9.24 x 107 1,250 2.26 x 108 1.08 x 108 1,500 2.15 x 108 1.24 x 108 1,750 2.06 x 108 1.38 x 108 2,000 1.96 x 108 1.52 x 108 3,000 1.77 x 108 2.04 x 108 5,000 1.56 x 108 2.96 x 108 7,000 1.44 x 108 3.8t) x 108 10,000 1.33 x 108 4.95 x 108 15,000 1.22 x 108 6.72 x 108 20,000 1.12 x 108 8.34 x 108 30,000 1.02 x 108 1.13 x 109 50,000 9.03 x 107 1.67 x 109 70,000 8.27 x 107 2.14 x 109 100,000 7.47 x 107 2.80 x 109
'150,000 6.69 x'107 3.78 x 109 200,000 5.92 x 107 4.66 x 109 300,000 5.03 x 107 6.18 x 109 400,000 4.43 x 107 7.50 x 109 500,000 3.99 x 107 8.66 x 109 700,000 3.43 x 107 1.07 x 1010 1,000,000 2.96 x 107 1.34 x 1010 1,100,000 2.84 x 107 1.42 x 1010 1,300,000 2.72 x'107 1.57 x 1010 1,500,000 2.61 x 107 1.72 x 1010 .
1,700,000 2.48 x 107 1.86 x 1010 l 2,000,000 2.36 x 107 2.06 x 1010 2,200,000 2.31 x 107 2.19 x 1010 2,400,000 2.26 x 107 2.32 x 1010 2,600,000 (=30.09 Days) 2.21 x 107 2.44 x 1010 l l 5UL J
D- O TABLE 020.1-2: Heat Rejection Rate and Integrated Heat Rejected from Containment and the Primary System for the Unit with the Design Basis Accident. Total integrated heat Time Heat rejection rate to rejected into (sec) ultimate heat sink ultimate heat sink (BTU /hr.) (BTU) 0 0 0 60 0 0 60 1.81 x 108 0 70 1.81 x 108 5.52 x 105 80 1.82 x 108 1.06 x 106 90 1.83 x 108 1.56 x 106 100 1.84 x 108 2.07 x 106 200 1.81 x 108 7.18 x 106 300 1.78 x 108 1.22 x 10 7 400 1.73 x 108 1,70 x 107 500 1.67 x 108 2.18 x 107 600 1.61 x 108 2.63 x 107 700 1.53 x 108 3.07 x 107 800 1.45 x 108 3.48 x 107 900 1.38 x 108 3.87 x 107 1,000 1.30 x 108 4.24 x 107 1,250 1.14 x 10 8 5.08 x 107
-1,500 9.66 x 10' 5.80 x 107 1,750 2.63 x 108 7.16 x 107 2,000 2.54 x 108 8.95 x 107 3,000 2.34 x 108 1.57 x 108 5,000 2.07 x 108 2.78 x 108 7,000 1.90 x 108 3.89 x 108 10,000 1.67 x 108 5.37 x 108 15,000 1,44 x 108 7.50 x 108 20,000 1.29 x 108 9.38 x 108 30,000 1.12 x 108 1.27 x 109 50,000 9.72 x 10 7 1.85 x 109 70,000 8.94 x 107 2.36 x 109 100,000 7.79 x 107 3.06 x 109 150,000 6.94 x 107 4.07 x 109 200,000 6.32 x 107 5.00 x 109 300,000 5.26 x 107 6.59 x 109 400,000 4.90 x 107 7.97 x 109 500,000 4.52 x 107 9.26 x 109 700,000 3.97 x 107 1.16 x 1010 1,000,000 3.02 x 107 1.45 x 1010 1,100,000 2.92 x 10 7 1.53 x 1010 1,-300,000 2.81 x 107 1.74 x 1010 1,500,000 2.76 x 107 1.85 x 1010 1,700,000 2.66 x 107 2.04 x 1010 2,000,000 2.60 x 107 2.22 x 1010 2,200,000 2.50 x 107 2.36 x 1010 2,400,000 2.45 x 107 2.49 x 1010 2,600,000 (=30.09 Days) 2.36 x 107 2.63 x 1010 e
I 12 ._.
E 1 TABLE 020.1-3i Heat Rejection Rate and Integrated Heat Rejected from Containment and the Primary System for the Safe-Shutdown Unit Total integrated heat
- Time Heat rejection rate to rejected into
{ (sec)' ultimate heat sink ultimate heat sink (BTU /hr.) '(BTU) 0 60
, 60 70 80 90 l 100 Heat 200 Rejected
- 300 through 400 steam.
.;. 500 Generator 600 .directly 0 l' 700 Atmosphere 800 for
- 900 the
- . 1,000 first 1,250 6 1,500 hours 1,750
,- 2,000 i 3,000 5,000 7,000 10,000 15,000 21,600 2.79 x 108 1.09 x 10 8.
i 30,000 4.59 x 108 50,000 9.31 x 107 1.03 x 109 70,000 8.29 x 107 1.51 x 10 9 100,000 7.47 x 107 2.16 x 10 9 4 150,000 6.53 x 10 7 3.13 x 109 200,000 5.92 x 107 3.99 x 109 ,. 300,000 5.03 x 107 '5.49 x 109 400,000 4.43 x 107 6.76 x 109 500,000 3.99 x 10 7 7.88 x 109
~700,000 3.41 x 107 .9.91 x 109
.1,000,000 2.96 x 10 7- 1.26 x 1010 1,100,000 2.91 x 107 1.34 x 1010 1,300,000.. 2.79 x 107 ~1.49 x 10 10 1,500,000- 2.65 x 107 1.63 x 1010
-1,700,000 2.51 x'107 1.77 x 10 10 2,000,000 2.36 x 107 1.96 x 1010
'2,200,000 2.32 x 10 7 2.08 x 1010 '
.2,400,000 . 2.27 x 10 7 2.21 x 1010 2,600,000 (=30.09 Days) 2.22 x 107 2.32 x 1010 s ;
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- TABLE 020.1-4: Total Heat Rej ecti .iate and Total Integrated Heat
. Rejected from Both Units. into Ultimate Heat Sink Time Total Heat rejection rate Total integrated heat (sec) from both Units rejected from both Units (BTU /hr.) (BTU) 0 0 0 60 0 0 60 1.81 x 10 8 0 70 1.81 x 10 8 5.52 x 105 80 1.82 x 108 1.06 x 10 6 90 1.83 x 108 1.56 x 106 100 1.84 t 10 8 2.07 x 10 6 8
200 1.81 x 10 7.18 x 10 6 300 1.78 x 10 8 1.22 x 10 7 400 1.73 x 108 1. 70 x 107 7 500 1.67 x-108 2.18 x 10 8 600 1.61 x 10 2.63 x 107 700 1.53 x 10 8 3.07 x 10 7 800 1.45 x 108 3.48 x 107 8 900 1.38 x 10 3.87 x 107
, 1,000 1.30 x 108 4.24 x 107 8
1,250 1.14 x 10 5.08 x 10 7 'l 1,500 9.66 x 10 7 5.80 x 107 8 j 1,750 2.63 x 10 7.16 x 10 7 8 l .2,000 2.54 x 10 8.95 x 107 3,000 2.34 x 108 1.57 x 10 8 5,000 2.07 x 10 8 8 2.78 x 10 8 7,000 1.90 x 10 3.89 x 108 10,000 1.67 x 10 8 5.37 x 10 8 8 15,000 1.44 x 10 7.50 x 108 20,000 4.08 x 10 8 9.38 x 10 8 . 30,000 2.21 x 108 1.73 x 109 50,000 1.90 x 10 8 2.88 x 109 5 70,000 1.72 x 10 8 3.87 x 109 100,000- 1.53 x 10 8 5.22 x 10 9 8 150,000 1.35 x 10 7.20 x 10 9 8 1 200,000 1.22 x 10 8.99 x 10 9 8 300,000 1.03 x 10 1.21 x 1010 400,000 9.33 x 107 1.47 x 1010 500,000 8.51 x 10 7 1.71 x 10 10 2.15 x 10 10 7 , 700,000 7.38 x 10
; 1,000,000 5.98 x 10 7 2.71x10$
1,100,000- 5.83 x 10 2.87 x 10 10
'1,300,000 5.60 x 10 7 3.23 x 10 1,500,000 5.32 x 107 3.48 x 10 10 i- 1,700,000 5.17 x 107 3.81 x'1010 2,000,000 4.96 x 107 4.18 x 1010
'2,200,000 4.82 x 107 4.44 x 1010 2,400,000 4.72 x 10 7 4.70 x 1010 2,600,000 (=30.09 Days)- 4.58 x 107 4.95 x 1010 1
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.:g : i i.{, j f, , _.;_.i .f . i= . lei . 4 __. .. . .I ._.i .l ,i _1 , g _' , I d. M lh5 $N - ~ i .I l ll' ! i ._ . go so' IO* 1 IO' IO' ME ^FTER AC'IDENT : 2'Aw (sec.) , (= 6 HRS) t s" FIGURE 020.l-6 : TOTAL INTEGRATED MEAT REJECTED INTO ULTIMATE HEAT SINK FROM THE PLANT ( FOTH UNIT.S ) 20 213.0 REACTOR SYSTEMS 213.1 Regulatory Guide 1.45 (Reactor Coolant Pressure Boundary Leakage Detection System 5/73) The monitoring of leakage from the Reactor Coolant Pressure Boundary (RCPB) is a safety function required by General Design Criteria 30. The staff recommends the monitoring of airborne particulate radioactivity for implementation of GDC 30. As stated in Regulatory Guide 1.45, at least one leakage-detection system should remain functional to assist in evaluating conditions in containment following an SSE. Describe the monitoring systems or procedures for sampling and surveillance of the containment atmosphere that are able to detect a significant increase in RCPB leakage following an SSE and before the system can be brought to cold shutdown.
Response
The Midland Project will comply fully with the requirements of Regulatory Guide 1.45 and the exception presented in our response to Regulatory Guide 1.45 is hereby withdrawn. A complete system description will be provided in the FSAR. 4 9 i 21
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310.0 ACCIDENT ANALYSIS 310.1 Regulatory Guide 1.4 (Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors Rev. 2 6/74) The source term to be used by the applicant is not clear. The 4 source terms specified in Regulatory Guide 1.4 should be used for the Midland LOCA analysis.
Response
The source terms specified in Regulatory Guide 1.4 will be used for the Midland LOCA analysis. i
.c2
310.2 Regulatory Guide 1.4 (Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors Rev. 2 6/74) An analysis of the iodine removal of the containment sprays should be provided,_ indicating the fraction of each form of iodine that will be removed from the containment atmosphere following a LOCA.
Response
The analysis of iodine removal efficacy of the containment spray system will be provided in the FSAR. 23 , -
310.3 Regulatory Guide 1.4 (Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolanc Accident for Pressurized Water Reactors Rev. 2 6/74) The applicant's proposal of meeting Guide 1.4, Paragraph C.2.3 is not in conformance with current NRC practice. This should be revised to meet this Regulatory Guide. The exposure doses from the LOCA should be based on a semi-infinite cloud for's 479.
Response
The applicant will calculate beta and gamma immersion doses frem the postulated LOCA based on a semi-infinite cloud. The wording in the Midland response to Regulatory Guide 1.4 Paragraph C.2.e will be changed from " finite" to " semi-infinite". However, the semi-infinite cloud equations presented in Paragraph C.2.e of Regulatory Guide 1.4 assume that radionuclide concentrations are uniformly distributed and everywhere equal to the concentrations at the receptor point. Integration of these uniform concentrations over the cloud yields an infinite inventory of radionuclides which contribute to the dose rates at the receptor point. This is in violation of the fundamental law of conservation of mass; that is, the dose rate at the receptor point can in reality only result from the finite inventory of radionuclides in the semi-infinite cloud (i.e., to the radionuclides which have been released and their subsequent daughter products minus the radionuclides which have decayed after release). The applicant will calculate gamma dose rates in a manner consistent with radionuclide conservation by taking into account the spatial distribution of radionuclide concentrations in the cloud (i.e., Atomic Energy and Meteorology Equation 3.115). The simple equations in Regulatory Guide 1.4 Paragraph C.2.e overestimate the gamma dose rates by roughly an order of magnitude because they fail to account for the fact that radionuclide concentrations decrease rapidly over distances comparible to the cean free path of the gammas. Such unrealistically conservative results are not deemed appropriate. On the other hand, beta immersion dose rates will be based on a semi-infinite cloud with a uniform concentration distribution. This is sufficiently accurate due to the short range of beta particles. Credit will be taken for an inert skin thickness of 2 7 mgms/m as per WASH-1258 using the recent paper " Beta-Ray Dose in Tissue-Equivalent Material Immersed in a Radioactive Cloud" by M. J. Berger in Health Physics, Volume 26, pp 1-12, 1/74 to predict energy loss as a function of tissue depth. 24
310.4 Regul .ory Guide 1.4 (Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors Rev. 2 6/74) t It is noted that the applicant states, "If charcoal filters are
. employed in the (fuel) building ventilation system. . .
ItLis recommended that the spent fuel building be provided with
, an ESF grade charcoal filter system which is automatically actuated by a high radiation signal.
Response
i The project intends to provide the filter system recommended by the Staff. 4 4 1 , .1 i a w i A ( l I [- 25
r p n 310.5 Regulatory Guide 1.54 (Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants 6/73) Consumers Power Company asserts that the coatings to be used within the Midland containment building will meet the intended purposes of Regulatory Guide 1.54 adequately, although some documentation as to the qualification of the personnel applying the coatings will not exist due to agreements with labor unions made prior to the issuance of that guide. To substantiate that essertion, the applicant should identify and estimate the quantity of all protective coatings applied within the containment. Significant amounts of coatings which will enter the containment on equipment to be instalJed there should also be identified and estimated. "Significant" is to be interpreted such that the total mass of unknown polymeric material within the containmen* is likely to be no greater than 100 kilograms, and appears as small surfaces with a typical dimension less than about 10 cm. Precoated small items may be identified by the resin base of their coatings, e.g., glyptal and phenolic heateured resin coatings. Thermosetting (" baked enamel") coatings having a phenol or phthalic acid base are likely to withstand LOCA conditions on small surfaces, even though non-rated. A description of the "phosphating" surface treatment of 56 valves should also be supplied.
Response
The applicant wishes to make the following points in response to NRC statement 310.5:
- 1. The applicant wants to insure that no misunderstanding exists relative to difficulties with compliance to the letter of Regulatory Guide 1.54. The NRC staff's statement that ". . .some documentation as to the qualification of the personnel applying the coatings will not exist" is correct but is only an example of the documentation not available.
This example should not be considered as all inclusive of documentatica required by Regulatory Guide 1.54 which is not available due to previous purchase order agreements. All LOCA qualified paints used on Midland have been qualified by the paint vendors.
- 2. The applicant does not agree with the NRC statement that all protective coating applied within containment be identified and the quantity estimated. As stated in the 26
310.5 etnt'd () 'is ' Response to Regulatory Guide 1.54 the applicant intends to identify and provide estimates of total surface area of non-LOCA qualified paints. The applicant can see no possible benefit to the staff or the applicant in providing this information for LOCA qualified paint.
- 3. The NRC staff's statement defining "significant" is understood by the applicant to refer to the amount of non-LOCA qualified paint used inside containment. It is the applicant's position that upon completing the tabulation of the amount of non-LOCA qualified paint used inside the Midland containment, should this amount be below the 100 kilograms mentioned by the Staff, the amount would be acceptable. In the calculation of this total weight, only components exposed to the LOCA environment will be considered. If upon completion of the tabulation of the amount of non-LOCA qualified paint used inside containment, this amount exceeds 100 kilograms, it is the applicant's position that the following will be considered in conj stnction with system characteristics (sump screen size and spray nozzle openings) to determine acceptability or necessity of any corrective actions:
1
- a. The areas of paint exposed to the containment LOCA atmosphere
- b. The mode of paint failure (i.e., flaking or solubility)
- c. The potential of the paint reactants reaching the containment sump, considering possible surrounding traps such as insulation, etc.
Current information indicates that use of non-LOCA qualified coatings is or may be required on small components, such as valve operators, switches, etc., and equipment trim. Many of these items, although small in terms of surface area, have dimensions in excess of 10 cm. Information on final surface areas will be available since it is a factor in the genera-tion of the total surface area discussed in 2 above.
- 4. The "phosphating" surface treatment of the 56 valves consists of immersing the valves, after caustic cleaning and rinsing, in a solution of Detrex 15 H or equivalent, containing 3 1/2 gallons of Detrex 15 H to each 100 gallons of solution. Detrex 15 H is a liquid chemical concentrate used to produce a corrosion resistant, zine phosphate coating on iron and steel producing a coating of weight 1000 to 3000 mg/sq. ft. After rinsing, a final oil dip into Parolac #17 allows this low viscosity oil to permeate into the porous structure provided by the phosphate coating.
27 ;}}