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Docket No. 50-358 August 20,19 71 REPORT TO THE ACES Wm. H. Zinsner Nuclear Power Station Unit I Construction Permit Review l
Division of Reactor Licensing U. S. Atomic Energy Consnission
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A&FFHGAltNTSE ONET i l
I ABSTRACT
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The Cincinnati Gas and Electric Company, Columbus and Southern Ohio l
Electric Company, and the Dayton Power and Light Company has submitted '
an application for a construction permit for the Wm. H. Zinmar Nuclear Power Station. The site is located near Moscow, Clermont County, Ohio, 25 miles southeast of Cincinnati, on the Ohio River.
The Zimmer Station will utilir.e a single-cycled, forced circulation, boiling. water reactor unit with a rated thermal output of 2436 MW and a gross electrical output of 807 MW. All safety systems and analyses have been evaluated at the design power of 2540 MW thermal and 840 MW electric.
The Cincinnati Cas and Electric Company has retained Sargent and Lundy to perform the architectural engineering services. The single-cycle, boiling water nuclear steam supply system will be furnished by the General Electric Company and the construction contractor will be the Kaiser Engineers Inc.
On the basis of our evaluation of the proposed facility and the j spplicant's qualifications, and subject to the resolution of the matters identified in Section 1.4, we have concluded that the Zimmer facility can be built and operated at the proposed location without undue risk to the health and safety of the public.
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OFFHCHALE&TONLY ii TABLE OF CONTENTS Page Abstract......................................................... 1 Tab l e o f Co n t en t s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11
1.0 INTRODUCTION
AND
SUMMARY
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1.1 Introduction.......................................... l' I 1.2 Princip al A rea s o f Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 l 1.3 Comparison with Similar P1 ants . . . . . . . . . . . . . . . . . . . . . . . . 8 1.4 Summary............................................... 14 2.0 S ITE C CHARACTERISTIC S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 2.1' Geography and Demography.............................. 16 2.2 Me t eo ro l o gy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 2.3 Hy d ro l o gy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 2.4 Geology. Seismolo gy, and Soil Me chanics . . . . . . . . . . . . . . . 26 2.5 Ecology............................................... 30 2.6 Environmental Radia tion Monito ring . . . . . . . . . . . . . . . . . . . . 31 2.7 Air Traffic........................................... 32 2.8 Railroad and Rive r Traf f1c. . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 3.0 R EACTO R D ES IGN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 3.1 Genera 1............................................... 34 3.2 Nuclear Dasign........................................ 34 3.3 The rmal and Hyd raulic Desi gn . . . . . . . . . . . . . . . . . . . . . . . . . . 36 3.4 Reactor Internals..................................... 37 4.0 REACTOR.................................................... 44 4.1 Sys tem Quality Group Classification. . . . . . . . . . . . . . . . . . . 44 4.2 Rea ctor Coolant Pres s u re Bo unda ry. . . . . . . . . . . . . . . . . . . . . 45 4.3 Reactor Core Support Structures-Design. . . . . . . . . . . . . . . . 59 4.4 Frac ture Toughne ss Crite ria. . . . . . . . . . . . . . . . . . . . . . . . . . . 59 4.5 Reactor Vessel Material Surveillance Program. . . . . . . . . . 60 4.6 Leak Detection........................................ 61 4.7 In se rvic e Ins pe c tion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 4.8 Reactor' Coolant System Sensitized Stainless Steel .. . . . 63 4.9 Fo r e i gn P r oc u re me n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64 4 .10 Elec t ros la g We1 din g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 64
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tii Pape 5.0 CONTAINMENT................................................. 65 S'1 General Containment
. Design............................ 65 5.2 P ri mary Co n ta inme n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 65 5.3 secondary Containment................................. 71 6.0 ENGINEERED SAFETY FEATURES.................................. 75 6.1 Core Standby Cooling System........................... 75 6.2 Residual Heal Removal Sys tem (RHRS) . . . . . . . . . . . . . . . . . . . 92 6.3 Post-LOCA'.iydrogen Contro1............................ 95 6.4 Containment Inerting.................................. 96 6.5 Long Term Cooling Water Supply........................ 97 7.0 PROTECTION, CONTROL, AND EMERGENCY ELECTRIC POWER SYSTEMS.. . 99 7.1 Genera 1............................................... 99 7.2 Protection Sys t ems Gene rf.c It ems . . . . . . . . . . . . . . . . . . . . . . 100 7.3 New Design Items...................................... 105 7.4' Ele c t ric Powe r Sys t e ms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 110 8.0 RADWASTE SYSTEM............................................. 117 8.1 Genera 1............................................... 117 8.2 Li q ui d Wa s t e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 117 8.3 Gaseous Wastes........................................ 119 8.4 Solid Wastes.......................................... 121 9.0 AUX IL I AR Y S Y S TE MS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 123 9.1 Genera 1............................................... 123 9.2 New Fuel Storage...................................... 123 9.3 Spent Fuel Storage....................................
124 9.4 Service Water System.................................. *125 10.0 STATION STRUCTURES AND SHIELD 1NG............................ 127 '
10.1 Classi fication o f Structures and Equipment . . . . . . . . .. . . 127 10.2 S t ruc tural Analysis and Des ign. . . . . . . . . . . . . . . . . . . . . . . . 128 10.3 Machanical Analysis and Design. . . . . . . . . . . . . . . . . . . . . . . . 133 10.4 Seismic Quality Assuranca............................. 136 s
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.4 GECLMrUSE UNLY iv Page 11.0 CONDUCT O F OP ERATIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 7 11.1 Organization and Responsibility . . . . . . . . . . . . . . . . . . . . . . 137 11.2 Station Management................................... 138 11.3 Emergency P1ans...................................... 140 11.4 Flant Securicy....................................... 142 11.5 Training............................................. 143 11.6 Operational Review and Audits . . . . . . . . . . . . . . . . . . . . . . . . 144 12.0 PLANT SAFETY ANALYSIS...................................... 145 l
12.1 Genera 1.............................................. 145 12.2 Loss-of-Coolant Accident............................. 145 12.3 Fuel Handling Accident............................... 147 12.4 Cont rol Ro d Drop Acc iden t . . . . . . . . . . . . . . . . . . . . . . . . . . . . 148 12.5 Ma in S t ea m Li ne Broe k . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 148 12.6 Instrument Line o r Process Line Break. . . . . . . . . . . . . . . . 149 13.0 QUALITY ASSURANCE.......................................... 153 14.0 TECHNICAL SP EC IFICAT IONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 156 15.0 CONFORMANCE TO GENERAL DESIGN CRITERIA. . . . . . . . . . . . . . . . . . . . . 15 8 l
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, V LIST OF TABLES P, age 1.1-1 Chronology for CC&E Application. . . . . . . . . . . . . . . . . . . . . . . . .
1.3-1 3-Major Design Changes Incorporated into the 1.3-2 Wm . H . Zimme r S t a t ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 Comparison o f BWR Design Parameters . . . . . . . . . . . . . . . . . . . . . 11 1.3-3 Comparison of Containment Design Parameters . . . . . . . . . . .
3.1-1 12 Thermal and Hydraulic and Nuclear BWR Design Parameters. 38 4.2-1 Reactor Coolant System, Pipe Whip Protection............ 48 12.1-1 Calcula ted Expos ure Do ses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
12.1-2 151 Mixing Re d u c t ion Fa c t o rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 152 LIST OF FIGURES 2.1-1 Cumula tive Population 0-5 M11es . . . . . . . . . . . . . . . . . . . . . . . . . 18 2.1-2 Cumulative Population 0-40 Miles . . . . . . . . . . . . . . . . . . . . . . . . 19 4.2.4-1 Recirculation Sys tem - Elevation Isometric . . . . . . . . . . . . . . 51 6.1-1 Clad Temperature Response Followin ACCIDENT. . . . . . . . . . . . . . . . . . . . . . . . . . g a LOS S-OF-COOLANT
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1.0 INTRODUCTION
AND
SUMMARY
1.1 Introduction
- On April 6,1970, he Cincinnati Gas and Electric Company (CG&E),
Columbus and Southern Ohio Electric Company (C&SOE), and The Dayton Power and Light Company (DPL), the applicants, filed an applict.; ion for licenses required' for the construction and operation of the pro-posed Es. H. Zimmer Nuclear Power Station. We Zimmer str. tion will consist of a single-cycle, forced circulation, boiling vater reactor i unit, constructed near Moscow, Clermont County, Ohio, 25 miles south-east of Cincinnati, on the Ohio River, he Cincinnati Gas and Electric
~ Company (CG&E) ic responsible for the design, construction and opera-tion of the Zinener Plant and is also authorized to act as agent for i
C&SOE and DPL in all details of construction, including licensing.
We Zissner facility will have initial power operation of 2436 MW thermal and a net electrical output of 807 MW. All safety systems and analyses have been evaluated at the design power level of 2540 MW thermal and 840 MW electric. Sargent and Lundy will perform the architectural engineering services. The General Electric Company will design, fabricate and deliver the single-cycle, boiling water nuclear steam supply system, and will also fabricate the first core of nuclear fuel. We construction contractor will be Kaiser Engineering Inc. We turbine generator unit will be supplied by the I Westinghouse Electric Corporation. A single hyperbolic, natural draf t, i
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OFRC:AL USE Ol&f cooling tower will _be used to dissipate waste beat to the atmosphere.
Cooling water for the service water system and makeup for the cooling tower is drawn from the Ohio River. All return water is routed to a settling pond prior to its return to the river.
The secondary containment, the reactor building, provides an addi- !
tional barrier to prevent the release of airborne radioactive materials, and is substantially the same as for other BWR stations.
The primary containment design is similar to that of Shoreham and Limerick Nuclear Power Stations wherein the "over-under" vapor suppression containment concept is used. The containment differs in that the drywell and wetwall are constructed frem prestressed concrete. This is the first BWR facility to use a prestressed )
concrete primary containment.
I The gaseous and liquid radwaste system will be designed to limit the processed affluents to an annual average concentration for routine discharges of less than 1% of 10 CFR Part 20 limits.
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l A chronology of the principal events during the review of the Wm. H. Zimmer Nuclear Power Station is given in Table 1.1-1. l I
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TABLE 1.1-1 i
Chronology For_ the CG & E Application
- 1. April 7, 1970 Submittal of PSAR Volumes 1 thru 5 and License Application for Units 1 and 2 (Docket Nos. 50-358 and 359).
- 2. May 12, 1970 Inf tial meeting with the applicant to discuss preliminary status of DRL' review of Zimmer Station.
- 3. May 21, 1970 Meeting with applicant to discuss signifance of Environmental Report and development of information provided in PSAR. j i
- 4. July 30 & 31, 1970 Technical meeting with applicant to dis-cuas meteorology, hydrology, geology, seismology, general site items and containment.
- 5. August 7,1970 Technical meeting with applicant to dis-cuss seismic design criteria, and structural analysis of Class I design for all forces (seismic, accident, wind, etc.)
- 6. August 25, 1970 Technical meeting with applicant to discuss hydrology - maximum probable flood design.
- 7. October 13, 1970 Request for additional information trans-mitted to applicant.
- 8. November 3, 1970 Amendnent No.1 - Submittal of geology, j seismology and site foundation information.
- 10. November 19 & 20, 1970 Technical meeting with applicant to discuss the station auxiliary systems and the quality assurance program, 1
- 11. November 20, 1970 Amendment No. 2 - Submitted information on i hydrology and containment.
- 12. December 16, 1970 Technical meeting with applicant to discuss instrumentation, control and electrical systems.
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-13. Decentier 17, 1970 Amendment- No. 3 - Revised PSAR to reflect change to a single unit design. ;
14 .- January 5, 1971 Technical meeting with apolicant to discuss seismic design values and liquefaction analysis.
~15. January 12,1971 Amendment No. 4 - Received response to DRL
- request .for additional information dated October - 13, 1970, . also included were revised ;
accident analyses and design changes 'to SGTS system.
- 16. January 22,.1971 Applicant submitted Environmental Report.
- 17. January 26, 1971 Meeting with the applicant to discuss back-filling technique under the reactor building, and service water pipes (between reactor building and pump house).
- 18. Feb ruary . 3, 1971 Technical meeting with the applicant and J
' February 10, 1971- General' Electric Company for a generic discussion of 1969 product line flow control valve, instrumentation, and control.
- 19. February 8,1971 Meeting with applicant to discuss liquefaction potential of soil under service water pipes and pump house.
- 20. February 12, 1971 Anaendment No. 5 - Completed revisions of PSAR :
as outlined by Amendments 3 and 4. )
21 February 25 & 26, 1971 Technical meeting with applicant to discuss i ECCS, pipe whip criteria and radiological l consequences of accidents.
- 22. February 16, 1971 Request for additional information transmitted !
to applicant.
l_ 23. April 3,1971 Amendment No. 6 - Revised Section 7 of PSAR l Instrumentation and Control and updated
- j. system classification of Pressure Boundry
{ Integrity Criteria, Appendix A.
[- 24 April 9, 1971 Request for additional information on the balance of review items transmitted to the applicant.
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- 25. ' April 19,1971 Amendment No. 8 - Provided anti-trust infor- l mation.
- 26. April 21,1971 Request for additional information on Environmental Report transmitted to the applicant.
- 27. April 27, 1971 Amendment No. 7 - Received response to DRL request for additional information dated October 13 and February 16, 1971 and revised quality assurance program.
- 28. May 17,1971 Amendment No.1 - Received response to request for additional information dated j April 21, 1971.
- 29. May' 17, 1971 Amendment No.10 - Provided information for notification to newspapers for legal notices.
- 30. May 17, 1971 Amendment No. 9 - Revised PSAR to incorporate new Appendices H.O - Resolution o f AEC-ACRS Staff Concerns, I.O - Seismic Analysis of Critical Structures Systems and Components, and J.0 - Analysis of Under-Ground Service Water Pipes and Supporting Piles.
- 31. .May 28, 1971 Amendment No.11 - Received response to DRL request for additional information dated April 9,1971, also provided primary containment hydrogen, orygen and fission product saspling system.
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- 32. June 15,1971 Amendment No.12 - Received response to balance of DRL request for additional information, also provided pipe whip criteria.
- 33. July 15, 1971 Amendment No.13 - Provided revision to DRL questions, and also provided conformance to AEC ECCS interim acceptance criteria and flood protection capabilities.
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- 34. July 16,1971 Request for additional information on generic CSCS questions transmitted to the applicant.
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35.- July 23, 1971 Amendment No . ' 14 - Revised - PSAR to incorporate additional details on inservice inspection program and radwaste off-gas treatment system.
- 36. July 27, and 28, 1971 Technical meeting with applicant and General" Electric Co. to discuss balance of instru-mentation, control, and electrical review.
- 37. July 30,1971 Amendment No.15 - Revised PSAR to incorporate split loop design of auxiliary system also included information on flow control valve, reactor nozzle failure, and primary contain- i ment design and leakage.
- 38. August 4, 1971 Amendment No.16 - Received response to request for additional information dated July 16, 19 71.
- 39. . August 9,19 71 Technical meeting with applicant, DRL, DRS, and our consultants to discuss balance of review on reactor building and service water .
pipe soil foundation.
- 40. August. 11, 1971 Amendment No.17 - Provided an analysis to -
show that the recirculation flow control valves do not have a safety significance.
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-GFFECAL UShMMLY 1.2 Principal Areas of Reviev The Zinsner Nuclear Power Station is the first facility to incorporate the General Electric 1969 product line. Many features of the nuclear steam supply system and principal protective systems for Zimmer are the same as the 1967 product line BWR plants; a comparison of these features served as a basis for identifying areas of continuing review of generic problems. Design features and potential problem areas unique to Zimmer were identified and evaluated. Euphasis was placed upon the following items during our review:
(1) site related items, (2) recirculation loop flow control valve used instead of a variable speed pump to regulate core flow, (3) compactness of the recirculation loop piping system (20 inch diameter versus 28 inch diameter previously used),
(4) primary containment prestressed concrete design (first BWR to use pres tressed containment) , -,
(5) the new 69 product line core stand-by cooling system, (6) the new solid state reactor manual rod control system,
( 7) new instrumentation, control and auxiliary power sys tems necessitated by items 2 and 5 above, (8) internal pressure of the primary containment following a pos tulated loss-of-coolant accident, OFMCEAL4HE ONLYm
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(9) adequacy of the service water system and Ohio River to supply l essential cooling water and (10) generic problem ~ areas and areas of ACRS concerns expressed in previous BWR reviews.
1.3 Cosparison with Similar Plants The major systems have been compared with those of other Boiling Water Reactor Units for which construction permits have already been issued or will be issued prior to Zimaner ACRS Review. The proposed Wm. H. Zimmer Nuclear f acility is similar in many respects to the Edv'n I. Hatch Nuclear Plant No.1, now under construction by the Georgia Power Company. Table 1.3-1 provides a comparison of signifi-cant differences between the Wm. H. Zimmer and Edwin I. Hatch facilities.
Table 1.3-2 provides a comparison of major BWR design parameters for Zinener and other similar BWR facilities. Comparisons of containment Design Parameters between Zimmer, Limerick and Hatch are given in Table 1.3-3.
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J0FMCAL USE-ONLY Table 1.3-1
, Major Design Changes " Incorporated Into the Wm. H. Zimmer Station Parameter Wm. H. Zimmer Edwin I. Hatch
- Reactor Coolant Recirculation kop Design Flow Control Throttling With An Recirculation Pump Additional Valve Speed Pump Flow Race, gpm/ pump 33,880 45,200' Total Core Flow Rate, 6 6 lbs/hr 78.5 x 10 75.5 x 10 Nominal Pipe Diameter 20 28 inches Core Spray System-Number of Systems 2** 2 Pump Flow Rate and Pressure, gpa (psid)*** 4625 (119) 4625 (120) 1330 (1130)
High Pressure Coolant Inj ection System Coolant Injection Mode Spray Flood Injection Method Directly into core Indirectly into core "la core spray sparger via feedwater sparger Flow Race, gpm (psid) 1330 (1130) 5000 (150-1130)
Pump Motive Type Motor with separate Steam Turbine Diesel generator
- The Edwin I Hatch Nucicar Pacility has the same equipment arrangement as other BWR facilities such as Browns Ferry, Peach Bottom, and Enrico Fermi-2.
- A high pressure core spray system is included and serves as a redundant low pressure core spray system.
- psid - pounds per square inch differential between reactor vessel and primary containment.
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i Low Pressure Coolant Inj ection System No. of Pumps 3 4 Flow Rate, spm/ pump (psid) 4970 (20) 7700 (20)
Loop Selection Logic None Sense Break Location i Injection Method Top of Core Recirculation Loop Piping Primary Containment
' Concept Over & Under Pressure Pressure Suppression Suppression (Lightbulb and torus).
Construction Type Prestressed Concrete ASME Steel Pressure.
Steel lined Vessel Drywell Geometry Frustum of Cone Light Bulb Shaped Pressure Suppression Cylindrical Torus Chamber (PSC) Geometry i
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1.4
SUMMARY
p Our review and evaluation of the Zimmer application is complete. ;
Some matt ers have been identified for which supplemental information must be provided during the construction permit stage and several-l matters have not been resolved yet. These are summarized below: !
1.4.1 Matters Involving Significant Regulatory Positions and/or Potential for Significant Design Changes l
- 1. Control of hydrogen concentration in the primary containment i
following a design basis LOCA (6.3)
- 2. Inerting of the primary containment atmosphere (6.4)
- 3. Main steam line isolation valve leakage (4.2.3)
- 4. Capability to perform a preoperational vibration test (3.4.4)
- 5. Independence of redundant channels of the reactor protection system (7.2.1)
- 6. Annunciation of a bypass condition for a channel of tiie engineered safety feature instrumentation (7.3.4)
- 7. Protection against anticipated transients without a scram (3.3)
- 8. Instrumentation for post-incident monitoring (7.2.3)
- 9. Seismic design classification of main steam line outside of the primary containment (4.2.6) reettcr_Am nmn a a **- a u>-- -
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2.0 SITE CHARACTERISTICS f 1
2.1 Geography and Demography j 1
The William H. Zimmar nuclear power station site is situated on a 635 acre tract of land in Washington Township, Claremont County, Ohio, and is located approximately 24 miles southeast of Cincinnati, Ohio and 1/2 mile north of Moscow, Ohio, on the Ohio River. n e minimum I exclusion distance. as defined by the applicant, is 380 meters and the established low population zone (LPZ) distance is 4,827 meters.
There are no military installations or missile sites within 300 miles of the plant.
The topography in the western portion of the reactor site is relatively level at about 500 feet above mean sea level (MSL) and is primarily farm land. he remainder of the site is hilly and partially wooded, with the highest elevation running up to 800 feet MSL. A small stream with a bed elevat".on of 470 feet MSL runs east to west through the site to the river. The reactor plant grade will be at 520 feet MSL.
We nearest population center with a population in excess of 25,000 people is Covington, Kentucky which is located 20 miles northeast of the plant. New Richmond, Ohio, with a 1960 population of 2,800 pro-jected to be 8,300 by 1985 is within 10 miles of the reactor site.
L l he+ cumulative population distribution in this same area (0-10 mile I
radius) is very low, with 21,500 persons in the 1960 census and is
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projected to be 30.100 by the year 1985.
The population distribution within the LPZ for the 1960 and 1985 (estimated) census is less tha 1,900 and 2,800, respectively. One third of the LPZ population live in Moscow, Ohio.
3:
Figures 2.1-1 and 2.2-2 give the cumulative popula-tion distribution (projected to 1985 census) versus distances for Zimeer, Zion, Indian Point, and Newbold Ieland nuclear stations . As shown in the figures, the Zinuser site is well within the envelope of sites previously considered acceptable.
There are no residences or facilities located within the exclusion radiva (380 meters), except a small manufacturing plant employing a labor force of approximately 75 persons.
The applicant has made legal arrangements with the owner of the manufacturing plant, to evacuate the area immediately or take whatever action is deemed necessary to prevent re exposu to these employees.
Several transportation facilities provide limited access to the site.
Normal access to the station will be U.S. Route 52 which trave rses the site about 1/2 mile east of the plant location
. An access road with a bridge over route 52 will be provided.
There is ne direce rail access to the site from the Ohio side of the Theriver.
Chesapeake and Ohio Railroad line is located on theside Kentucky of the river within several hundred feet of theAshore. barge unionding and' landing facility will be constructed to handle rge la
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s equipment. Commercial barge and boat traffic exists on the river throughout the year. The applicant has indicated that there are no )
recreational activities near the site.
Based on the analysis of the popula' tion distribution in this region, and on our evaluation of the calculated potential offsite doses dis-cussed in Section 12.0 of this report, we conclude that the proposed exclusion radius and low population zone distance meet the 10 CFR Part 100 guidelines and are acceptable.
2.2 Meteorology The applicant has presented five years of meteorological data from the Creater Cineinnati Airport (CVC) which is located in Covington, Kentucky. The prevailing wind at the airport is from the south-southwest with an average wind speed of 4.0 meters per second. The data indicate that calms occur approximately 3% of the time, and Pasquill type C conditions occur 5.44% of the time with an average wind speed of 0.75 meter per second. The applicant has conducted a short-ter:n meteorological program at the plant site. Since the onsite data consisted only of wind direction and speed at two locations, data collected at Creater Cincinnati Airport were used for the analysis leading to the short time period accident dilution factors. From this analysis the applicant concluded that a Pasquill stability Class F (determined from isolation and wind speed) accompanied by a i
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wind speed of 0.5 meters /second was justified. ne applicant's comparison between wind speed at the site and wind speed at the 1
airport showed that the airport speed was 2.5 times the site speed j on the average. We wind speed was corrected for this difference. i j
i ne accident dilution factor evaluation was therefore based on j i
Pasquill stability Class F conditions and wind speed of 0.2 meter /second.
Assuming a ground level point source with a building wake correction 2
factor (cA=900 m ) which is limited to a factor of three, Pasquill T stability and wind speed of 0.2 meters /second, the resulting dilution factor at the nearest site boundary (380m) for the short term release (0-2 hours) is 5.2 x 10~3 sec/m3 . ' he same diffusion equations and meteorological assumptions were used to calculate a dilution factor at the low population zone distance (4827m) for the first eight hour release period. This resulted in a valua of 2.96 x 10 ' sec/m3 . ~
For time periods greater than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, we used the meteorological paraineters listed in Safety Guide No. 3, Page 3.5 figures 3 (A) and 3 (B).
It should be noted from the PSAR tables showing the radiological effects of variot's accidents that the calculated results do not reflect terrain festures. At the ties of the operating permit review, when the applicant and the staff have valid onsite meteorological data, the effects of riger valley channeling will have to be considered in determining doses ,at the sitt boundary and at the low population tone distance.
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C b.hbb hbb he applicant planned originally to use the sigma theta method for determining the Pasquill' conditions. We informed them that this i method is not valid during periods of low wind speed. Me re fore .
i we requested that at measurements in addition to sigma theta type measurements be provided. This same was requested for the Beaver Valley and the Trojan reactor sites. De applicant has agreed to -
include At measurements in their onsite meteorological program.
We conclude that the calculated dilution factors are adequately con- _
servative.
he onsite meteorological program proposed by the applicant will check the adequacy of the values.
2.3 Hydrology The plant is on the east side of the Ohio 1tiver about one mile north i
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of the village of Moscow and approximately 25 miles southeast of l j
Cincinnati. The natural ground at the site is about 40 feet above the normal reservoir level of Markland Lock and Dam (elevation 455 ftf MSL). Plant grade is elevation 520 ft MSL. i i
he historical flood of record (1937) has been estimated to have produced a maximum runoff rate of 830,000 cfs and a river level of approximately '515 'fc MSL, which is about 60 feet above " normal" river level, he applicant has estimated that a probable maximum l
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. a uw a su oab U%"Jii Yv flood (PMF) would produce a peak runoff rate of 1,980,000 cfs and a corresponding river level of about 546 f t MSL. Considering the possibility of wind-wave action the applicant has established an elevation of 550 ft MSL as the level for which flood protection will be provided. The applicant has also evaluated the effects of the potential feilure of upstream dams. He has concluded that those located along the main stem of the Ohio River are too low to cause a higher flood level at the site than the PMF. His investiga-tion of the many tributary reservoirs indicates their relative )
location and size would allow flood waves caused by failures to be almost fully attenuated before reaching the site, and in no case cause a flood level greater than the PMF.
The applicant will provide flood protection for all plant structures to elevation 521 ft MSL. Safety related plant structures will be permanently flood protected to elevation 546 ft MSL. He has also investigated the effects of potential wave action above the flood protection level and has concluded waves may reach approximately elevation 550 ft MSL. Safety related structural openings will be protected to an elevation of 550 ft MSL.
The applicant has stated the essential cooling water requirements for Unit 1 are 28 'efs. The historical minimum instantaneous low flow of 2100 cfs in the area was recorded at louisville, Kentucky, 1
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OFF10AL USE ONLY
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on August 20, 1930. Since then dans have been constructed which can be comted on to augment low flow for all but the most severe drought conditions, he applicant has concluded that suf ficient flow exists in the' Ohio River for emergency requirements. He has also evaluated the effects of zero flow condition and has concluded that sufficient storage exists in Markland Reservoir above the proposed service water intake elevation of 439 ft MSL to provide sufficient a
water supply storage for safe shutdown. he applicant has also evalu-ated the ' extreme low flow condition in the Ohio River (equivalent to the local minimum flow of record in the area) coincident with complete loss of Markland Dam. His analysis indicates that the resulting water level would be of sufficient height to assure adequate suction head on the pumps to maintain omsegency shutdown. .
Ground water in the area is drawn from aquifers near the surface. The upper twenty to thirty feet of material is unconsolidated silty and clayey sand. Beneath the alluvium and to depthe ranging from sixty to ninety. feet beneath the surface are located sand and gravel glacial outwash deposits. Both the alluvium and glacial outwash deposits provide a usable ground water resource. Perched water table conditions were found where a few discontinuous thin silt and clayey lenses occur above the normal water table. Bedrock composed matly of limestone and shale are found at depths below ninety feet. Both rock forma-tions are relatively impermeable, and are not generally used as water
- - ^ -
.0FMC%L USE ONLY supply sources since well yields are generally less than one gallon per minute. These rock deposits also act as aquitards against down-l ward water movement. The applicant has estimated permeabilities decreasing from 463 gal / day /ft to 122 gal / day /ft between depths of 30 to 60 feet, respectively. Surface infiltration is low and has been estimated by the applicant at about 0.4 gal / day /ft . Ground water gradients are normally toward the river, although at high river levels flow reversal occurs.
Eleven towns or water districts within 25 miles of the site serve between 500 and 14,000 persono at rates ranging from 0.05 to 1.57 mgd. Four public users, generally down gradient from the site. are located at distances ranging from about 5 to 17 miles from the site.
They are, however, along the banks of the Ohio River, which can be expected to intercept most ground water flow passing beneath the site. We closest public well is on the same bank of the Ohio River as the site and about three miles upstream. There are 10 private wells within one mile of the site, and a total of 12 wells within three miles of the site. The applicant has determined that there are no public or private wells presently in the direction of ground water movement which could be affected by conditions at the site.
The applicant has, investigated the effects of both surface and ground wa?. r spills o f radioactive material. The applicant has concluded that m
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- OIPMC3L EE U?i.~.,Y a release of the entire contents of the liquid radwaste collector, floor drain and chemical waste tanks (35,900 gallons) without any holdup in the plant would encounter low permeability in the layer of clayey silt inusediately beneath the site. In addition, the ion exchange capacity of the soil can be expected to reduce the activity by as much as 80 percent before such a slug would reach the Ohio River. Finally, the applicant has assumed mixing of the activity with an average 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Ohio River flow of 10,000 cfs and has determined that the effective concentration at the Cincinnati water intake 25 miles downstream would h less than 10-30 pCi/cc.
We conclude that the flood protection provided for the facility is acceptable and that sufficient water is available in the Ohio River for emergency purposes. We also agree with the applicant that no existing public or private ground water or surface supplies facilities are likely to be affected by conditions at the site.
2.4 Geology, Seismology and Soil Mechanics The site is located within the Central Stable Region of North America . .
an area characterized by broad, circular to oblong erosional domes and . ,
sedimentary basins. A system of arches connect the domes and separate the basins. The site lies near the top of the Cincinnati Arch. At the site the Precambrian crystalline rocks are mantled by a' proximately 3,500 feet of sedimentary rocks of Paleozoic age. Bedrock consisting anmnennn nv m m 2[.
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. OEEICMJL USZ (C of Ordovician shale and limestone ranges in depth from 83 to 88 feet beneath the Ohio River Valley alluvium.
Faulting has not been identified within the sedimentary strata in the vicinity of the site. The nearest known fault to the site is the Maysville Fault approximately 30 miles southeast of the site. The age of this fault is considered to be Paleozoic, and definitely older than middle Tertiary.
Major faulting in the eastern Missouri-southern Illinois-western Kentucky region, more than 200 miles from the site, is not structurally or tectonically related to the geologic structure at the site. In addition and of greater consequence to the site is the Cincinnati Arch, also ~ referred to as the Indiana-Ohio ristform, which has a history of earthquake activity. The applicant attempted to demonstrate that geologic conditions are such that the seismic activity at Anna, Ohio, {
i should not be related to the proposed site location. Our consultants at USGS agreed with the applicant's interpretation that any activity which may occur on the major structures south and west of the plant would be subordinate to seismic activity occurring in the Cincinnati Arch. However, due to the lack of geological information about the geologic structure of the Anna area and the Cincinnati Arch, our consultants were not able to agree with the applicant's contention that the seismic activity at Anna can be precluded from occurring
~. p
OFMCIALUSE'OETT 29 -
conditions. However, the assurance given in the form of computed l factors of safety were dangerously, if not unacceptably, low.
Newmark and Associates performed independent analyses of the Class I structures assuming the same ' conditions described above. The resulting analyses showed that the Reactor and Auxiliary Buildings would move laterally and the service water pipeline which is to be supported on pile bents would move toward the river due to drag from the liquefied soils tending to move downslope. Since the pump house is so massive and because it will be supported on a caisson founded at bedrock, it was felt that its stability is assured. Also, tha t portion of the service water inlet at the river will be protected by a sheet pile bulkhead. The bulkhead design would be acceptable provided it is designed for earth pressures at rest, for dynamic pressures, and that these pressures be assumed to act during low water levels.
Our consultants Newmark and Associates indicate that the fix to assure stability of the reactor and auxiliary buildings requires a clay blanket which will envelope the foundations of these structures to
, prevent excessive pore water pressures and consequent sliding of structures in the event of a DBE and high water.
The applicant has agreed to provide the clay blanket and to make observations prior to the operating license review both inside and TUPUAU a awn WL 1T FC" UD U 9 / ul/f y
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UAh.o li outside the clay blanket of piezometric levels and rate of change of water levels under partial flood conditions to determine whether pumping will be needed to prevent saturation of the foundation soils within the clay blanket. Provisions will be made so that if pumps are needed they will be installed.
Regarding the service water pipeline, our consultants reconnended batter piles to. provide for lateral forces both downslope and trans-verse of the order of 360 kips per bent downslope and 100 kips or less in the transverse direction. The applicant has agreed to perform an analysis to determine the number of pile bents and battered piles that are required to resist these forces. We find this acceptable.
2.$ Ecology The topography in the vicinity of the site is relatively level with sparsely wooded sections. The land is predominantly used for farming and information on the types of acreage and yields of pertinent local crops has been provided. The applicant has already performed an ecology study of the area in the vicinity of the proposed site. The purpose of the program was to determine the biological and relevant physicochemical conditions near the site prior to construction in order to assess the environmental impact of the facility. The use of the cooling tower and settling pond is expected to minimize the effect on the aquatic ecology.
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_ s a ma 6' C iFa U A L 2 T 4 The ecology study was' performed in July through October,1970 and included daily and seasonal distributions, and abundance of various .
life. forms of the terrestrial and aquatic communities. The aquatic conusunity study determined the composition o'f the plant and' animal components and the physical and chemical characteristics of the aquatic ecosystem at the site. The species studied includes Benthos.
Periphyton, Phytoplankton, and Fish. The terrestrial study documented-plaut and animal diversity, distribution, and relative abundance of-
- s. oils vegetation, invertebrates, amphibians and reptiles, birds, and mansnals. . The. data accumulated will be used in the pre and post environmental monitoring program proposed for the site.
Based on our review of the information provided we find that the ecology program is acceptable. We and our consultants, the Fish and Wildlife Service of the Department of the Interior, conclude that the program is acceptable.
' 2. 6 Environmental Radiation Monitoring The applicant has . submitted hic environmental monitoring program that will be conducted at the Zimmer facility.
The proposed environmental radiation monitoring program will determine the levels of radioactivity that exist at the site and the surrounding environment. The program will be initiated two years prior to fuel loading and will continue for at least two years I
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during plant operations. This program will include collection and
. radiometric an'alysis of airborne ' particulate, gansna radiation, sur-L l
face water, bottom sediments, bottom organisms, slime, well water, precipita!. ion, fish, soil, vegetation, milk, wildlife, and miscel-laneous food items, on and off-site in both directions along the-
- 1. Ohio' River. The area within 10 miles of the site' will be monitored for radioactive gases released to the atmosphere. Air samplers will be placed in population areas greater than 100 within five miles of the site and in population areas of 1000 or greater from five to ten miles from the site. Based on the data contained in the PSAR.
we conclude that the applicant's preoperational environmental monitoring program is acceptable.
2.7 ' Air Traf fic Two small airport facilities are located about 10 and 15 miles north of the site. These airports do not offer scheduled commercial flights and their use is limited to one-engine light aircraft for business or personal use. The nearest air facility that accommodates large com-merical aircraf t is the Greater Cincinnati Airport located 25 to 30 miles northwest of the facility. Based on guidelines used on previous reviews, we have concluded that the probability of an aircraft crash affecting the Zimmer facility is sufficiently low that no special design provisions are needed.
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2.8 Railroad and River Traffic i
There is no railroad access to the plant from the Ohio side of the ]
1 river. The Chesapeake and Ohio Railroad traverses the Kentucky side of the Ohio River, within several .hundred feet of the shore. The applicant plans to provide a barge unloading and landing facility to accommodate shipments that cannot be transported by truck. 'Ihe rail-road car to barge transfer point will be located on the Kentucky side of the river. The existing transfer f acility at the J. M. Stuart Station (a fossil plant owned by applicants) will be used or a new installation will be constructed opposite the fseility. W e Ohio side landing facility will also be constructed at the river's edge near the plant.
The applicant has analyzed the potential effects of accidents involving barge traffic and its associated cargo. Included in the analyses were effects of barge traffic and potentially dangerous cargo such as toxic chemicals, explosives and flamable materials. The applicant concluded and we agree that the facility could be maintained in a safe condition -
following postulated accidents associated with the barge traf fic.
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l 3.0 Reactor Design ,
i 3.1 General l
The reactor design of the Zimmer Nuclear Power Station, in many respects, is the same as that of several other BWR plants previously reviewed.
Tables 1,3-2 and 1.3-3 provide a comparison of major reactor design parameters for a number of these facilities. Although Table 1.3-1 identifies significant differences between the Zimmer end Hatch facilities, comparison of the tabulated data shows that the nuclear fuel, reactor vessel and the thermal-hydraulic characteristics are basically the same. Our review and acceptance of the nuclear, thermal and hydraulic design was based primarily on the comparison with previously reviewed and approved Boiling Water Reactor characteristics.
3.2 Nuclear Design The nuclear fuel and its arrangement in the Zimmer core is substantially the same as that of previously approved plants.
The nuclear reactors compared in the Tables 1.3-2 and 1.3-3 use reactivity control syste.ms that are all identical. Control of the core reactivity is provided through movable cruciform control rods and a variable recirculation flow rate which automatically accommodates demand for increase or decrease in power. A standby liquid control system is provided that injects into the core a solution of water and sodium pentaborate, a strong neutron absorber. The method used Y \(lti L
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l tio regulate the recirculation flow rate through the core has changed significantly. Previous BWR's utilized a variable speed recirculation pump while Zimmer is the first nuclear plant to utilize a flow control valve and a constant speed pump in the recirculation loop piping system. Mechanical and electrical devices associated with limiting positive reactivity insertion into the core include control rod velocity limiters, epecified control rod operating patterns, limited control rod drive speed capability, a control rod worth minimizer program and its I rod blocks (to inhibit selection, withdrawal, or insertion of out-of-sequence rods during startup, shutdown, or low power operation), and ! control rod drive housing supports. The red patterns administrative 1y permitted will limit individual rod worths to less than 0.01 dk/k. ne nuclear fuel is designed for 19,000 WD/MTU (averaged over the initial core load). Enrichment in U-235 for the fuel charge will not be known exactly until about eighteen months before the initial fuel loading. Typical fuel enrichments for a fuel assembly are 1.10 w/% to 2.60 w/%. He highly enriched fuel rods ato designed and made with large diameter ) end plug-shanks that can fit only in the proper location in the uppsr 1 tie plate of the fuel assembly. The design permits lower enriched rods to be placed in the locations for higher enriched rods accidentally, but , this would result in a fuel assembly with less than the standard fissionable loading. The fuel elements used in the Zimmer core will nDiFlirinCilAIL UsE nimN > < > , . - . '
4 JWFCLWUSE ONLY _ 36 _ employ burnable poison rods (Gd 23 0 g-M ) to control long tem reactMty [ [ changes . The use of these burnable poison rods were reviewed and determined to be acceptable during the Quad Cities operating license review. The applicant has indicated that the General Electric Company will obtain inservice data on use of the fuel. The analysis will provide the inservice operating limits for reactor operation. The applicant indicated that there'are no plans for exposure experimentation using full length production fuel rods. We conclude the information submitted provides a suitable basis to expect safe performance of the reactivity control mechanisms and the nuclear fuel under normal and - accident operating conditions of the Zimmer facility.
. 3.3 hermal and Hydraulic Design _
The thermal and hydraulic characteristics of the Zinsner and Hatch - reactor cores are nearly identical. They are very similar to the higher power density cores of Newbold Island, Browns Ferry, and Peach Bottom. Figure 3.1-1 provides a comparison of characteristics for these facilities. Core cooling systems for the facilities are identical: two recirculation loops and twenty jet pumps for each reactor. The i recirculation flow rate for the Zinener facility is less than the flow ) rate of similar BWR facilities. However, the total developed head is
. higher so that the total jet pump flow is about the same. This could q af fect the jet pump performance characteristics. This area is being rO 5 5IIEII$\{L {L55 5 $ 5k?G 5 0RL)a, ,
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1 _OFFHCHAL1JSE4NhY ; 1 investigated to assure that the jet pump design falls within the experimental design limitations. During normal steady-state operation the thermal hydraulic design of the Zinsner core will assure a minimum critical heat f1'u x ratio (MCHFR) of 1.9 or greater and the maximum linear heat generation rate will be maintained below 18.5 kW/f t. Average core power density is 51.2 kW/ liter. 3 General Electric submitted a topical report " Analysis of Anticipated Transients Without Scram" (NEDO-10349) in March,1971. Our review of this report was submitted to the Committee on July 30, 1971. We plan to require the applicant to provide the capability to trip the two recirculation pumps on coincident high pressure and high neutron flux level signals to assure acceptable limits and consequences of an ATWS in the Zimmer plant. 3.4 Reactor Internals 3.4.1 General The design of the reactor vessel internals will be in accord-ance with the intent of Section III of the ASME Boiler and I Pressure Vessel Code. The material used for fabrication of l most of the reactor vessel internals is solution-treated unstabilized Type 304 austenitic stainless steel conforming
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to ASTM specifications. Weld procedures and welders will be l qualified in accordance with the ASME Boiler and Pressure Vessel Code. t I MWS MU U@D Ammm1T f\ 17 ftT@ E g -
\ ll0[!dSEEJJLY 41 -
(Proprietary) to the Quad Cities docket and referenced in Amendment 7 to Zimmer. W e additional information required to complete our review of Quad Cities Amendment 19 will be submitted as a topical report. Final acceptance of the procedures proposed for determining the dynamic loading due to normal and upset operating conditions for the Newbold. Island reactor internals will be withheld until we complete our review of Quad Cities' Amendment 19. We expect to complete this review in the near future and will apply the results to Zime r . If necessary, confirmation of the procedures employed to determine the normal and upset condition design dynamic loadings for the Zimmer reactor internals structures can be obtained during the pre-operational vibration test program for this plant by the use of additional instrumentation and analyses beyond that currently contemplated. h is matter is discussed in more detail in the following Section 3.4.4. Design loadings for the postulated Loss-of-Coolant Accident (LOCA) will be determined by computing the response of each structural member to the calculated peak pressure differential applied as an equivalent static load. In response to our con-cerns regarding the validity of this static analysis the appli-I cant has stated that the natural' frequency of the BWR internal structures is more than ten times the calculated frequency of i f\m L f
L ~n m m em um ILo V U LCo U A\ 1 ff : submitted, we believe that the applicant should take all measures nee'escary during design and construction to provide the flexibility to implement a preoperational test program. The applicant has not agreed to provide the capability to conduct a vibration test program. "Dtis program should measure the responseN of the reactor internals to determine the flow-induced forces and the related dynamic forcing functions for all significant modes of normal reactor operation. 'Ihe data obtained by these measurements on reactor internals should be sufficient to verify that the steady state and cyclic stresses in the components, as determined by analyses, are within the acceptable design limits set forth in the design specifications and code requirements and that the results meet the acceptance criteria of the vibration test program. In the event that the Zimmer plant is not the first of the 69 product line to receive an operating license and has conducted a satisfactory vibration test program, we may relax our requirements, but as a minimum a confirmatory type vibration testing program will be required. l 1 I ! N Frequency and magnitude of vibration (in terms of displacements, velocities. and accelerations) . 1 l bbbNlls bbb bb a g.g[v I
_ OFFlCLWUSE ONLY ~ 44 - 4.0 REACTOR 4.1. System Quality Group classifications The applicant has applied a system of code classification groups to those pressure containing components that are part of the reactor coolant pressure boundary and other fluid systems important to safety. I These classification groups generally correspond to the tentative code classification groups A, B, C and D developed by the regulatory 1 staff. We and the applicant are in general agreement on the code classification groups for the reactor coolant pressure boundary and the majority of those fluid systems important to safety. For those systems, portions of systems, or components whers the applicant's classification group-ing dif fers from ours, the applicant has upgraded his classifications to a quality level substantially equivalent to the applicable staff
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classification code groups. Clarification of these upgraded classi-fications is documented in Appendix A.0 of Amendment 6 and in Amendment 7 to the PSAR. The ASME code for pumps and valves does not adequately cover pump and valve pressure retaining castings for lines over 2 inches up to and I including 4 inches in systems which are classified A & B and in ( Group D for main steam and turbine bypass lines. To cover these castings adequately volumetric examination (radiography or ultrasonic A
4 OFfRC. ALJJSE ONLY - testing) is necessary. Where size or configuration does not permit
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volumetric examination, surface examination (magnetic particle or 11 quid penetrant testing) may be substituted. The applicant has agreed to conform to the ASME code for pumps -and va~1ves including added testing outlined above. Group Classification Diagrams, Figures A.2-1.1 and A.2-1.2 in the PSAR are acceptable, however, we will request that the applicant provide the piping and instrumentation diagrams identifying the detailed boundary limits of each classification group during the construction stage. We ' find that the system quality group classifications as described are acceptable. 4.2 Reactor Coolant Pressure Boundary 4.2.1 gag The reactor coolant pressure boundary will be a Class I (seismic) system designed fabricated and inspected in accordance with the requirements of the applicable codes delineated above in System Quality Group classifications _; the stress limit criteria specified for the normal and upset operating condition categories of the applicable codes will l-apply for all normal loads and anticipated transients including , the Operational Basis Earthquake. l __ _ _ _ _ da C
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( '_ l These design, fabrication, and inspection criteria are consistent j l 1 with those accepted for a11'recently reviewed plants of this type and we find them acceptable. ! Under the loads that result from the Design Basis Accident, l l the Design Basis Earthquake, and the combination of these, ,l
'1 the components of the reactor coolant pressure boundary will l
. i be designed to the applicable emergency and faulted operating condition category limits of the appropriate codes. Where the appropriate codes do not provide explicit design limits for these operating condition categories, these components will be designed to the criteria submitted in Appendix C of the PSAR.
The criteria of Appendix C as modified by Amendments 12 and 13 I 1 are consistent with comparable component code criteria. We find these criteria acceptable. ] 4.2.2 Pipe Whip Criteria The applicant states that he will examine all lines which are part of the reactor coolant pressure boundary for their potential to whip. In no case will reactor coolant system pipe whipping be allowed to cause any damage to the reactor coolant system, essential equipment or to the containment. Pipe whipping will be prevented by installing restraints and/or placing check valves or closed motor operating valves
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' ao close to the reactor vessel as possible. Protection against the effects of longitudinal as well as circumferential breaks will be included in the design. Table 4.2-1 outlines the pipe whip protection provided for each line. We find the pipe whi'p criteria acceptable.
4.2.3 Main Steam Line Isolation Valve Leakage . l The general subject of the effects of direct leakage from BWR steam and feedwater systems under accident conditions (i.e. , I . leakage of fission products that bypass the engineered safety features and escape directly to the environs) has been discussed with GE and various licensees over the past several years with-out any apparent satisfactory resolution.
.v At the time of the Dresden 2 review, GE provided the results of an evaluation of Icakage through the lines assuming that the associated steam line, turbine and condenser, remained intact following a LOCA to provide holdup time. In operating BWRs' technical specifications,a leak rate of 11.5 scfh at 25 psig .g for each valve was accepted as an interin action. This rate ., would be unacceptable if the steam tiysten was assumed not to )
I be intact. The results of an evaluation by GE assuming that l the system is not intact indicates that the calenlated LOCA doses at the site boundary would be 12,000 rem and the 30-day y LPZ dose would be 3,000 rem. prmw en pn - - - n-- r , ,
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' U ? II5 [ A- { { Q q The subject continued to be ' discussed on subsequent licensing 9
cases with GE claiming a continued evaluation of'other fixes. Applicants have proposed various design modifications to resolve the problem (e.g. , a seal water system) . It is apparent that the steam line isolation valves by themselves are not sufficient to provide an adequate leakage barrier. We do not agree with GE on its simple calculational model regarding an intact steam system. We plan to inform the applicant that a solution to this problem is required that would not place complete reliance on the maintenance of low leakage valves. 4.2.4 Reactor Coolant Recirculation Loop Flow Control Valve 4.2.4.1 General Zimmer is the first of the GE BWRa to incorporate a flow control valve in conjunction with a constant speed pump to vary the recirculation loop flow rate. This differs from the previous BWR product line plants in which a pump speed was varied by a motor-generator set to accomplish the same purpose. Each of the two recirculation loops in the Zimmer Station has a motor ~ driven constant speed, vertical, centrifugal pump which develops 845 feet of head at a flow rate of 34,000 l . i
f Eltok i gps, The flow control valve in each loop is 20 inches in - i diameter and is of the 45'Y-pattern type. Valve motion is accomplished by an electro-hydraulic actuator. Each loop also contains two 20-inch-diameter gate valves which serve as block valves, and a six-inch-diameter globe valve which is .j d located in a bypass line around the flow control valve. The arrangement of these components is shown in Figure 4.2.4-1. The flow control valve can be operated in either the manual or the automatic mode. The automatic mode is operable over a 35% range, normally from 65% to 100% power. In the automatic mode, turbine load / speed and steam line pressure signals are input into the master controller. The power demand output of 1 the master controller is input to the neutron flux controller ' and compared with the APRM signal which is also input to neutron flux controller. The neutron flux controller provides a total recirculation loop flow demand signal to a flow controller which, in turn, provides a flow demand signal to each recircu-lation loop. l There are also various interlocks that provide pump protection - l l AgainSt cavitation. These can cause the pump to trip, prevent 1 j pump startup, cause the control valve to close, and/or cause L I .. 1 l i Ab b A f*l h W
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below. , 4 4.2.4.2 Test Program l The valve and actuators are being developed currently. Tests have been performed on an eight-inch-diameter valve and the data are being used to design the full size valve. The valve actuator and some electronic components currently are being developed and tested. The tests include performance capability, failure modes, and the determination of the maximum valve ! strexing speeds (as limited by the hydraulic design). The current development schedule contemplates testing of the i nt taaster contro'41er, flux controller, flew controller, and associated lirdtens.daring mid-1972 The fully assembled valve,, actuator 'and control system will be tested with the' Zimdt recirculation pump at the end of 1972. These tect will be conducted in a test loop under actual operating temperatures. , We have discussed the possibility of accelerated life testing with the appIfcant and GE. However, we have not established M 3m E } $ I R O I M ll 3 L i! h. V
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a position. in this regard and ' expect to continue our discussions - i in conjunction with the planned testing program. 4.2.4.3 Safety Evaluation
-The safety evaluation of.the recirculation loop flow control considered valve malfunctions, and the interaction of valve motion with a' loss-of-coolant accident.
The applicant has analyzed the consequences of the flow control valves in both recirculation loops opening at maximum speed and closing at maximum speed. Thermal limitations were not exceeded in either case. The applicant also analysed valve cycling. It was found that a high flux scram would occur before thermal limits are approached. In the event that the flow-rate-sensing signal in one recir-culation loop were lost while operating in the automatic mode, the valve in the other loop would receive a signal to open. This situation was analyzed at various power levels by the applicant. Again it was found that satisfactory thermal margins were maintained. Motion of the control valve and of the pump discharge block valve is not pnecluded following a loss-of-coolant accident (LOCA). 'In fact, the normal response of the controls and l 9
bb bbb hIk.:$ interlocks would cause the control valve to close at a rate of 10 percent /second following a LOCA._ Valve motion would probably be initiated several seconds after the pipe break. l If a failure of a valve actuator hydraulic line were to occur,
' the valve could close at a rate of 20 percent /second. The interlocks on the pump discharge block valve would cause it to close approximately 30 seconds after the occurrence of the LOCA. These valve motions would have the effect of decreasing the effective size of postulated breaks in the recirculation lines. This, in turn, would reduce the rate of reactor pressure decrease and thus affect the performance of the ECCS. The applicant is currently analyzing this potential problem assuming a design basis break whose size is reduced following the LOCA.
Thc. applicant has provided the following information orally concerning the preliminary results of the analysis. (a) Valve closure initiating after lower plenum flashing could af feet the performance of the ECCS to the extent i that maximum cladding temperatures are increased above ' those that would be calculated for the design basis break if no credit were taken for the reduced flow area
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assoc'iated with the actual valve dimensions. l I (b) Valve closure would only affect the consequences of pipe { breaks occurring downstream of the valve; i.e. , between ) l ( ___ ___
0 5 .E0 h Y 5Eb) hkl,.:$ locations A and B shown on Figure 4.2.4-1. For pipe breaks in these locations the analysis can be based on 1 2 the flow area of the valve (1.3 ft ) rather than the flow area of the recirculation pipe (2 ft ) . Using this assumption the calculated maximum cladding temperature would not exceed 1800*F. Resolution of this item is i
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swaiting completion of our review of Amendment No.17 that l contains the applicant's analysis of these accidents. It is expected that satisfactory resolution of this matter will be achieved prior to the ACRS meeting.
, 4.2.5 primary System Pressure Relief The objective of the pressure relief system is to limit any over-pressure of the reactor coolant boundary (reactor vessel and recirculation lines) that might occur from abnormal opera-tional transients. In addition, the automatic depressurization feature of the system is used in the event of small primary system breaks to depressurire the primary system for low pressure coolant in,jection system operation. Eight safety-relief valves are provided that discharge to the suppression pool and perform the following functions (a) in the relief mode the valve is opened by a signal generated by high primary system pressure to limit primary system overpressure to below the self-opening l
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@@W"C" A 17 "mm e?- - S u u uwruu 4/D)6 UAD pressure of the safety and safety-relief valves, (b) in the safety mode the valves are self actuated by primary system high pressure to augment safety valve capacity and (c) to auto-matica11y depressurize the primary system the valves are opened from signals resulting from a loss-of-coolant accident.
There aire five safety valves that discharge directly to the drywell and function in conjunction with the safety-relief to prevent overpressurization of the primary system. The safety-relief valves are spring-loaded valves equipped with air cylinder operators that. permit remote, manual or automatic opening of the valves at pressures below their self-opening set pressure. The safety valves are the conven-tional spring-loaded type valve. Safety valve capacity for this system is based on a pressure rise resulting from a main steam flow stoppage (turbine trip) at operating conditions, turbine pressure 980 psig (103% rated), no steam bypass of the turbine and a reactor scram due to high pressure. The analysis indicates that a design safety valve and safety-relief valve capacity of approximately 100% of reactor rated flow is capable of maintaining an adequate pressure margin (approximately 60 psi) below the peak ASME Code allowable pressure of 1375 paig. Eight of the thirteen
1 0EC AlSE 0bY" safety and safety-relief valves will limit the peak pressure of the reactor coolant boundary to 1375 psig. Based on the above, we conclude that the system design is acceptable. 4.2.6 Seismic Desistn of Main S team Lines The main steam line (MSL) piping from the reactor pressure vessel out to and including the ML outer contair. ment isola-
. tion valve is designed, fabricated and inspected to the requirements of Class I (seismic) including Quality Group A.
The main steam lines from the outermost valve up to the tur-bine, however, are designed, fabricated and inspected to the requirements of Class II (seismic) including Quality Group DI. Consequences of postulated failures of this portion of the main steam line are presented in Section 12.0 of this report. Although the calculated radiological consequences of such acci-dents are well below the 10 CFR Part 100 guideline values, they are significant. For other recent applications, most recently Newbold Island, we have required that the seismic classification of the main steamline (from the outer EL isolation valve to the turbine stop valve) be upgraded to reduce further the i probability of a steam-line-break accident. The specific ) i i requirements we have used are discussed below.
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- he MSL from the outer MSL isolation valve up to end including the turbine stop valve and all branch lines 2al/2 inches in diameter (IPS) and larger up to and including their first valve '(including their restraints) should be designed by the use of an appropriate dynamic seismic analysis to withstand i the OBE and DBE loads in combination with other appropriate loads, within the limits of the ANSI B31.1 piping code. The dynamic input loads for design of the MSL should be derived from a time history modal analysis (or an equivalent method) of the Auxiliary, Reactor, and applicable portions of,the Turbine Buildings.
The Class II Turbine Building, housing the MSLs may undergo some plastic deformation under the DBE, however, the plastic deformation should be limited to a ductility factor of 3 and an inelastic multi-degree-of-freedom system analysis should be used to determine the input to the MSL. The stress allowable and associated deformation for piping should be limited to 1.2 times the stress allowable for OBE and 1.8 times the stress allowable for DBE. The MSL supporting structures (those por-tions of t,he Turbine Building) should be such that the MSL and its supporte can perform their safety function under the Class I (seismic) loading conditions. b(([C3A[, [$TMdy J)g v
1 l' OE2LfSLDEY We have concluded that these requirements should be applied to the Zissner plant. We plan to inform the applicant of these . req uirements . 4.3 Reactor Core Support Structures-Desian For all operating condition categories, i.e. , normal, upset, emergency and faulted, the core support structures will be designed to stress, . deformation and fatigue limit criteria which are consistent with the criteria of the code for core support structures currently in preparation by ASME. We find these criteria acceptable. 4.4 Fracture Toughness Criteria The reactor vessel will be designed in accordance with the ASME Boiler and Pressure Vessel Code, Section III. Current ASME Section III Code rules permit that a vessel be pressurized only above a temperature equal to the sum of the Nil Ductility Transition (NUT) temperature and 60*F. The NDT temperature, according to paragraph N-331 of the Code, can be obtained by either the dropweight test (DWT) or the Charpy V-notch (C ) impact tes t. y Recent fracture Eoughness test data indicate that the current ASME Code rules are not always sufficiently conservative, and may not guarantee adequate fracture toughness of ferritic materials. While the Charpy V-notch tests continue to be useful in measuring the upper shelf fracture energy value, the C, specimens , generally, do not predict correctly the NDT. temperature. The latter, therefore, must be obtained i
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_ .-- 17 7 ww II uu.o @ wana "2 M W from other tests, such as the DWT test. Quite often, also, considerable difficulty exists in defining from the Cytest curves the transition temperature region in which fracture toughness of ferritic materials increases. rapidly with temperature. In addition, this transition tem-perature region shifts to higher temperatures when the thickness of the specimen tested is increased (size effect). We have described the proposed AEC fracture toughness criteria to the applicant and requested that adequate fracture toughness data be available
. to establish appropriate heatup and cooldown limits for this plant at the time of the operating license review.
In addition, we have asked the applicant to provide' us with the fracture toughness data for all pressure-retaining ferritic components of the , i reactor coolant pressure boundary. The data submitted by the applicant l! meet the current requirements of Section III of the ASME Code, but are ; i not adequate to establish compliance with the proposed AEC fracture touphness criteria. We intend to review the fracture toughness data l I available at the time of operating license review, and we will apply )
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. the AEC fracture toughness criteria to establish appropriate heatup and cooldown limits for this plant. I t
4.5 Reactor Vessel Material Surveillance Program i The material surveillance program is consistent with ASTME-185-66 f l
- l. which has been accepted on previous similar BWR plants. The reactor ,
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a hkb b hhh)L _ 3 vessel material surveillance program submitted by the applicant -includes provisions with respect to total number of specimen capsules placed in the reactor vessel, number of capsules scheduled .to be withdrawn and tested, archive' material available for additional specimens if required later in the service life of the vessel, and material chemistry docu-mentation. We conclude that this program will adequately monitor radiation-induced changes in material fracture toughness properties of the ferritic materials of the Zimmer reactor vessel, during its service life.
-4.6 Leak Detection The applicant's proposed reactor coolant pressure boundcry leak detection system within the drywell is improved with the use of airborne particulate sampling equipment. The equipment is similar to that now used at Dresden 2 and 3 and is considered acceptable for leak detection. .The air sampling points in the drywell and suppression chamber are monitored continuously by drawing samples outside the con-tainment and measuring gross beta activity. The drywell air sampling system supplements other leak detection equipment that permits measure-ment of drywell pressure, temperature, and sump level. Air samples taken manually from specific areas provide a means to determine the approximate location of the break.
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b4 .a bb[b Obi.sY i Temperature, pressure, and flow sensors with associated instrumentation and alarms are provided beyond the limits of the reactor coolant pres-sure boundary for leak detection of vital fluid-carrying systems external to the primary containment. We conclude that the leakage detector systems proposed by the applicant are acceptable.- I 4.7 Inservice Inspection 4.7.1 Inservice Inspection Program for Reactor Coolant Pressure Boundary The access for inservice inspection of the reactor coolant pressure boundary will be in compliance with the ASME Boiler i and Vessel Code; Section XI: Rules for Inservice Inspection of Nuclear Reactor Coolant Systems. Access is provided for each applicable component within the reactor coolant pressure 1 boundary in accordance with the requirements for inspection given in Table IS-261 of Section XI. , I 1 The applicant has provided the necessary equipment for both remote and contact inspection of the reactor coolant pressure 1 boundary. A procedure has been prepared to coordinate the equip-ment development with the reactor plant design. Procedures and systems are being developed for recording and gathering the inservice inspection data and for comparing the data taken in inservice inspections with that taken in the baseline inspection. ' - ~ ~ m m___ _ __ m m e i,a ,
t OFJECHAL USZ6LY We conclude that the inservice inspection program is acceptable. 4.7.2 Inservice Inspection program for Group B & C Fluid Systems The applicant's program includes provision for inservice inspection, for the Group B and C fluid systems, to the maximum extent feasible. Engineered safety features will i receive functional tests and inspection to assure integrity and operability. We conclude that the Group B and C Fluid Systems will be designed to permit periodic inspection and that the proposed access provisions are acceptable. 4.8 Reactor Coolant System Sensitized Stainless Steel 1 The applicant states that the sensitization of non-stabilized stainless steel (S.S.) during fabrication will be avoided. The precautions used to assure this include rapid cooldown from solution heat treatment 1 temperature, the use of low carbon S.S. where exposure to temperatures above 800*F will be experienced due to postweld heat treatment operations, the use of low carbon welding electrodes, and the limiting of weldment 1 interpass temperatures to be below 350*F. Non-sensitized stainless ! steel nozzles and pipe safe-ends will be affixed after the final vessel stress relief. We conclude that the planning to avoid sensitization of austenitic stainless steel during the fabrication period is acceptable.
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RECTIALJLJSE ONLY
- 64 -
4.9 - Foreign' Procurement The applicant has not yet selected all of the suppliers of reactor coolant pressure boundary components. The applicant has stated that no foreign procurement is anticipated and that if foreign procured com-ponents become necessary, a tabulation of components would be provided. 4.10 Electroslag Welding The applicant has not yet selected all of the suppliers of reactor coolant pressure boundary components. The applicant was asked to provide a tabulation of all components using the electroslag welding process. In response to our query, the applicant stated that the use of electroslag welding is not anticipated. However, should the process be used, the applicable components and process information will be identified and provided. _ _ _ A_ s
v- _ b hf'hh.a 5.0 CONTAINMENT 5.1 General Containment Design The containment systems include the primary containment using the pressure suppression concept and the secondary confinement which includes the reactor building, its recirculating (atmospheric ventilation) system, and the standby gas treatment system (SBCTS). The containment configuration is similar to that for the Limerick ! I and Shoreham nuclear power stations. The drywell is a steel-lined prestressed concrete vessel in the shape of a frustum of a cone. The vapor suppression chamber is a steel-linad prestressed concrete The right circular cylinder located directly beneath the drywell. drywell and wetwell are separated by a reinforced concrete floor penetrated by eighty-eight vent pipes. A low-leakage Reactor Building surrounds the primary containment to serve as a secondary barrier. A comparison of the containment design parameters for the Zimmer Nuclear Power Station with those of other BWR's is presented in Table 1.3-3. Both primary and secondary containments .will meet, among others, the criteria for Citas I seismic design. Structural aspects of the containment design are discussed in Section 10.0, Station Structures and Shielding. 5.2 Primary Containment The vapor suppression concept for the reduction of pressure inside the primary containment following a LOCA has been used l l l nomeneaw w wara n e w m ilj,. _
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An d .b hjb Od in the Zimmer design as* in other BWR f acilities. This is a departure from the traditional BWR structural design in that the dryvell and wetwell configuration is of the "over-under" type: the drywell is constructed above the vetwell and together they form a continuous, single structure for the primary containment. The codes utilized in the design of the primary containment are ACI 313-63, " Building Code Requirements for Reinforced Concrete" and the AS4E Boiler and Pressure Vessel Code, Section III, Subsection B. The latter code governs design of the drywell head, locks, penetrations. and other steel structures having pressure vessel functions. Where appropriate, the latter code also applies to the steel liner; however, the liner plate is designed to function only as a leak tight menbrane. Containment liner material is A-516 steel, Grade 60 to SA 300. Connecting the drywell and wetwell are eighty-eight straight pipe vents, which are approximately forty-four feet long and two feet in diameter. The vents project a short distance above the reinforced-concrete drywell floor and extend into the suppression pool to provide a flow path for uncondensed steam into the water. Each vent opening is shielded by a steel deflector plate to preve:t overloading any single vent by direct flow from a pipe break near that particular vent. The steel deflection plate will double as a temporary seal during the pressure test of the dryvell floor a:d also prevents m egenAP g ygg gwgg y g j g) ,, , , - 3
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foreign objects from entering the vent. The vents are structurally tied together by supports to provide resistance against forces which will be developed during the postulated LOCA. Vacuum breakers are provided to equalize the static pressures between the suppression j chamber and the drywsil and provide a controlled return flow path from the suppression chamber to the drywell to assure design operation of the suppression chamber in the event of a small steam leak. The applicant has supplied the results of the LOCA containment pressure transient based on the new vent flow model. The new model is documented in the General Electric Company Topical Report NEDO-10320, "The General Electric Pressure Suppression Containment Analytical Model," April 1971, and Supplement 1 to NEDO-10320, May 1971. Additional information j i supporting and extending application of the model has been presented in the Limerick ACRS Report. During our review of the primary containment a number of areas were 1 reviewed in detail. These areas are discussed below Post-LOCA Containment Pressure _: As a result of the continuing (a) review of the method of calculating the post-LOCA peak drywell containment pressure, a revised analytical model of the contain- ! I taent pressure transient was developed. The applicant recalculated l i the LOCA peak drywell pressure using this refined vent flow model. l The recalculated peak drywell pressure and peak deck differential I
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pressure increased only 'slightly (less than 2 psig) over the l 1 earlier calculations. The pressure increase for Zimmer is small j 1 because the suppression system has a large vent area to break area ratio (twice as much as for Limerick), which makes the vent flow model less sensitive to the size of the primary system ] rupture. ] l An extensive review of the containment design and. blowdown model led us to the conclusion that a 15% pressure margin should be added to the peak calculated drywell pressure and a 30% pressure margin should be added to the calculated peak deck differential pressure. The margin will provide for a limited number of unknowns such as a somewhat different accident from the design basis accident and errors associated with the calculation and asstamptions used. These design margins were applied to the Zimmer containment system and the following containment design parameters were found acceptable Calculated Design Margin ; 1 Containment Peak 38 45 20% ! pressure psig l Peak Deck Differential 15 25 55% pressure psig l
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. II( N.I A. n I wea wa (b) Drywell Deck Design: The-drywell deck mus: be designed to minimize the likelihood of bypass of blow &wn steam from the drywell directly into the air chamber above the suppression pool. Short circuiting of the pool would pro' duce significant higher pressures in the suppression chamber and the drywell.
Several potential bypass areas have been identified. These are the periphery joint of the dryvell floor to containment wall, joints between the downcomers and the dryvell floor, and cracks in the reinforced concrete floor. The cocerete drywell floor will have minimal amounts of cracking as a result of 60 Kips per foot of hoop prestressing on the slab. The prestressing and thermal expaneion will tend to seal the containment wall periphery joint and any developed crack. The downceeer vents that penetrate i the dryvell floor will be equipped with welded seal plates and anchored to the concrete at the top and bottom of the slab. In addition they will be epoxy coated to improve ' bonding capability and reduce leakage potential. Based on the drywell deck design, we conclude that the potential for bypass leakage has been minimized, (c) Wetwell-to-Drywell Vacuum Relief Valves: The relief valves are positioned in t.he downcomers and are designed to withstand dynamic loads associated with a LOCA. However, failure of the valve represents a possible leakage path from the drywell to e
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1 h {..$h .N$s 0kY i I the suppression chamber air space. The valves are continuously 1 monitored and. remotely operable for testing from.the reactor i building. Failure of a valve in an open position could result ' in a potential bypass area of'about 0.42 ft with a consequent 4 psi increase in peak containment pressure. This pressure increase is well within the design margin of the containment. (d) Drywell Deck LeakaRe Test: A leak rate test program will be developed to verify the leak tightness of the concrete deck. We will review the detailed test program during the operating . license review to assure that an adequate test program is developed.
.(e) Primary Containment Penetrations: Penetrations of the primary containment are in accordance with current design criteria.
The applicant's instrument line isolation system is designed in accordance with AEC Safety Guide number 11 and therefore, is acceptable. (f) The concrete containment structure is designed to accommodate substmospheric pressures of approximately 5 psi absolute. Studies showed that the maximum cooldown rate of the post-LOCA drywell atmosphere by containment spray produces a 2.5 psig pressure dif ferential. Accordingly, the applicant is not pro-l l viding vacuum relief valves between the inside of the primary containment and the reactor building atmosphere. 1 i
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1 ppg o n -- y n iw t h 4.8&lk %#fL_; ! 5.3 Secondary Containment The secondary containment structure (Reactor Building) is ; designed to minimize release of airborne radioactive materials and provides for a controlled release of building atmosphere so that offsite doses from the postulated design basis acci-dents will be below 10 CFR Part 100 limitations. The Reactor Building will enclose the reactor and its primary containment. The Reactor Building exterior walls and superstructures up to the refueling floor are constructed of reinforced concrete. Above. the ' level of. the refueling floor, the building structure is fabricated of industrial steel members, insulated siding, and a metal roof. Joints in the superstructure panelling will be caulked during installation to assure leak tightness. Penetrations of the Reactor Building are designed to have leakage characteristics con-sistent with leakage requirements of the entire building. 1he design criteria is to provide a leaktightness that would limit in- ' leakage to 100% building volume per day at 1/4 inch water (vacuum) while operating the Standby Cas Treatment System. The following f sections present infomation on design features and equipment which contribute to the leakage control capability of the Reactor Building. 5.3.1 Reactor Building Recirculation System 1 1 The reactor building recirculation system is provided to assure that mixing of the reactor building atmosphere occurs in the M _U M A R_ N e r na 'i 11.
3sva wert.v A r-w e.Q r a& n.isw tr vm 4 4 A_%xJ ., 1 y-1 event of an accident that requires reactor building integrity (postulated loss-of-coolant accident or the refueling accident.) Within ten seconds after receipt of an appropriate signal (high pressure in the drywell, high radiation in the reactor building exhaust ventilation duct, or manual initiated signal from the control room) the reactor building is isolated from the outside atmosphere. The recirculation system and the standby gas treatment system (SBGTS) both automatically start on isolation of the Reactor Building. The recirculation of Reactor Building air, following an- l isolation signal, utilizes the normal ventilation system ductwork which has been sealed to prevent outleakage from the building. The normal ventilation fans are shutdown and one of the two 100% capacity redundant reactor building recirculation fans is started. The fan' suction draws air from the areas above and below the refueling floor and discharges the air throughout the reactor building. The flow rate in the main duct to the fan is about 80,000 CFM. A small fraction, about 2300 CFM, of air is exhausted to the standby gas treatment system (SBGTS) in order to maintain a slight negative pressure (1/4" water vacuum) in the reactor building. l l l
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(f Dl (((M f/ 4,iky , , 3 , 5.3.2 Standby Gas Treatment System The Standby Gcs Treatment System (SBGTS) consists. of two parallel process systems designed to meet seismic Class I requirements. Each. process system will have a minimum capacity of 100% of reactor building volume per day. Each equipment train of the ' E dual process systems have: a demister for removing excess moisture; a prefilter capable of removing 80 to 85 percent of ; particulate'; electric heating coils to reduce the relative humidity of the gas entering the absorber bed to less than 70 percent; a high efficiency particulate filter (HEPA) capable
- of removing 99.97 percent of particulate matter that is 0.3 micron or larger in size; an iodine filter (impregnated activated carbon bed) capable of removing not less than 99
)
percent of iodides; a mixing device, utilizing the vortex principle to assure is homogeneous mixture of air and the filtered gases; an additional iodine filter identical to the one above; and an additional HEPA filter identical to the one described above. Both standby gas treatment trains will be started automatically following receipt of an appropriate signal ) (high drywell from high radiation or by manual activation). The applicant has indicated that the f odine filter efficiency j l of each filter will not be less than 99 percent as a result of - d
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i & -I.f. II li k,i7..,v. 4 MDv.V. the charcoal bed increased depth and due to a gasketless, welded seam type design that eliminates the bypass of air around the charcoal bed. The maximum filter efficiencies for removal of iodine allowed on previous BWR facilities has been 95%. The new filter design has been evaluated and a filter efficiency of 99% was found acceptable. We conclude that the design of the SBGTS is acceptable. ; l
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JQFFECHAL URE NLV-l l ' 6.0 ENGINEERED S AFETY FEATURES 6.1 - Core Standby Coolina Systems l 6.1.1 System Description GE's 1969 product line (CSCS) is significantly different from j other BWR (1967 product line) plants for which construction permits and operating licenses have been issued. A comparison of the emergency core cooling systems for '69 and '67 product line plants is presented in Table 1.3.1. The 1969 product .; I line core standby cooling systems coraist of the High Pressure Core Spray System (HPCS), the Automatic Depressurization System (ADS), the Low Pressure Core Spray System (LPCSS), and the Iow Pressure Coolant Injection System (LPCIS) which is one mode of operation of the RHR System. The various systems are initiated automatically by a high dryvell pressure signal or a low reactor vessel water level signal, with the exception of the ADS system which requires coincidence of the two signals. I The CSCS is designed to provide adequate core cooling for the complete break spectrum up to and including the design basis break which is the complete double-ended circumferential rup-ture of a recirculating pipe. The design basis break for Zimmer is 2.2 ft which is significantly smaller than the approximate 5 ft design basis break for the 1967 product line plants. Apprn A3L;3 tan nutst a ! .h s -
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t OECliAL US&etW u The salient features and significant diffs;rences from previous plants are.' described below. The new HPCS system combines the function of the former high pressure coolant injection (HPCI) system and one of ' the core spray systems (CSS) of previous BWR Plants. A single motor-driven pump- and associated pumping and instrumentation is used : to perform both ' functions if offsite power is lost. Power for the HPCS pump is provided by a separate diesel generator. HPCS injection coolant enters the vessel and is. piped to a
.J sparger over the reactor core via two entry points near the top of the shroud. Nozzles spaced around the sparger spray the water over the top of the core and into the fuel assemblies.
This injection mode, through the core spray sparger rather than through the feedwater sparger, represents a significant departure from previous BWR designs. The system's core cooling capabilities are designed to function over a wide range of reactor coolant' system pressures and break sizes. For small breaks that do not depressurize the reactor vessel rapidly, the system will maintain reactor water level and depressurize the vessel. For large breaks, rapid depressurization occurs and the HPCS cools the core in the spray cooling mode until sufficient inventory is accumulated to terminate the transient.
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'ha a b/k\lla Udlb hk The pump characteristics are selected to satisfy requirements -
for both the high pressure and low flow rate for small breaks and low pressure and high flow rate for large breaks. flhen the cooling system is activated, the initial flow rate is established by primary system pressure. As the reactor pressure I falls, the flow rate will increase until required core spray flow rate is achieved when the vessel reaches 200 psid (dif-ferential pressure between the reactor vessel und the primary containment). The HPCS systems NPSH requirements are met I without reliance on a pressurized primary containment. The ADS remains the sama as for previous BWR systems. The system rapidly reduces reactor pressure and enables the low pressure standby cooling systems to function. The ADS utilizes six of the relief-safety valves in the nuclear pressure relief system. These valves are actuated by coincident signals of low water level and high drywell pressure. The ADS will not be activated unless either the LPCS or LPCI system is operating. The LPCS system consists of a centrifugal pump powered by 1 either normal offsite power or by the onsite diesel-generator. The system is identical to other BWR LPCS systems except; that only one loop is provided. The HPCS system operating in the low pressure mode serves as a redundant core spray loop. When A emm ^-e m m_ _m- m _ _ _ _ _ _ - h. a
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- the reactor vessel pressure is low enough, water from the suppression pool is' piped to a spray sparger above- the core.
This sparger.is separate and distinct from the HPCS spray sparger.- The LPCS pump is located in the reactor building below the water level in the suppression pool to assure posi-tive pump suction. NPSH requirements can be met without l reliance on a pressurized containment. j i l i The LPCI system injects water from the suppression pool directly l 1 into the core region through three separate nosales to flood-the core. Previous BWR designs used the recirculation loop piping to inject the'LPCI cooling water. The LPCI system is initiated , by the same signals as the HPCS and LPCS systems and operates independently to cool the core. The low pressure coolant injection system is one mode of operation of the Residual Heat , Removal (RHR) system. The pumping system is designed to provide both adequate head and coolant flow capacity to meet flooding ! I requirements for the entire break spectrum. The use of l suppression pool water establishes a closed loop for recir-culation of LPCI water. Because the LPCI supplies water directly to the reactor vessel, the recirculation loop l i l l _ _ _ _ _ - - _ _ _ _ _ - - - a
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i' l OFFKL% MSE ONLY selection logic used on previous BWR designs to sense break locations is not required. 6.1.2 Discussion of ECCS Review The procedures used by GE to analyze the consequences of a LOCA depend upon the particular break size and the location being evaluated. It has been shown by CE that the ' worst-case' situation (i.e. , highest cladding temperature) arises for a break in the coolant recirculation lines because they have the potential for causing the coolant mass loss from the vessel to be more rapid and more extensive than for breaks in other lines in the reactor coolant system that would carry either steam or two phase fluid mixtures. Based upon our review of the GE analyses we have found that the ' worst-case' situation, with regard to assessing the performance of the CSCS, would' be for an instantaneous break of a large recirculation outlet line. For the purposes of analyses, the changing thermal and hydraulic phenomena that are associated with a design basis loss-of-coolant accident (LOCA) may be described in five phases: (a) temperature changes and heat removal during reactor blowdown with associated flow coastdown, (b) achievement of critical heat flux at any point on the fuel rod cladding and associated temperature rise t --------~~m_ _ , r
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of fuel and clad material, (c) lower plenum flashing causing - a temporary resurgence of core flow,1/ (d) temperature rise of 1
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cation, and (e) temperature changes and heat removal during l ECCS operation. Figura 6.1-1, is a diagram showing temperature i as a function of time for the five phases of a design basis loss-of-coolant accident. The analysis of each of these phases f of' the LO6. originally were performed using calculational models and edw& ques different from the models and techniques currently used by the applicant (and GE). Since the original analysis additiona) information and results of tests related to the performance of the CSCS systems also have become avail-able. As discussed in the following sections, we have met with GE on many occasions to review details of the codes, models, and analyses. In addition, independent checks of certain porcion6 of these calculations using different codes have shown reasonably g>od agreement with the GE results. The first' phase of the LOCA is the short-term blowdown during which energy is removed from the core by coolant passing through j the core and exiting through the postulated break, causing the reactor coolant system pressure to decrexec rapidly. Initially y
-1/This phase occurs when the fluid in the lower plenum reaches a saturation condition resulting in a rapid expansion of the fluid causing a large flow increase in the core region.
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l 4 conditions are nearly the same as during normal operation and nucleate boiling continues undisturbed. During the nucleate boiling regime, a heat transfer coefficient of about 3 x 10' Btu /hr-ft 'F is calculated to exist during this period of time which is slightly less than 9 seconds. A short time later, the core flow and system pressure decrease ! sufficiently that nucleate boiling cannot be sustained and
- the heat transfer rate decreases markedly. In previous analyses a " dry-out" model was used to calculate the time at which the degradation in heat transfer occurs. This model was based on the results of tests in which the flew in a heated test i
section was stopped simultaneously with initiation of depressurization. Because the flow in a BWR core after a pipe break is expected to coast down rather than to stop ins'tantly, degraded heat transfer will occur later than would be predicted using the " dry-out" model. In the current analyses, the time at which the departure from the nucleate boiling occurs in determined using empirical correlations based on the results of steady-state critical heat flux (CHF) tests. We requested additional information from CE to confirm the conclusion that the use of a steady-state correlation is l appropriate. GE has performed transient tests to demonstrate f the validity of their steady-state CilF correlations during
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o C 03%t?? transient conditions. The tests were performed in which flow and pressure reduced separately and concurrently. Comparison of the CHF measured in the tests with the CHF predicted by the steady-state correlation show that the use of the correlation gives results that are conservative. The short-term blowdown phase ends when the coolant flow through the core is assumed to stop as the water level in the downcomer region reaches the inlet of the jet pump. Even though flow actually would continue at a very reduced rate, for conservatism in the analysis, the flow in the core is assumed to stop at this time. During this period of flow stagnation, the heat transfer coefficient used in the analysis { is assumed to be equal to zero. This implies that no heat transfer by conduction or convection from the fuel rods to l the coolt.nt occurs and the fuel heats up adiabatically except l for the heat loss due to thermal radiation from the fuel j surfaces which is assumed to take place during this phase; which exists for about 3 seconds. The GE analysis yields a ! maximum calculated cladding temperature during this period of about 1500'F. Similarly, GE calculates that the pressure in the primary system during this second phase of the postu-lated LOCA is continuing to decrease because of mass loss through the break, and the depressurization rate during the 111, J
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.c short-term blowdown regime is of the order of 10 psi /sec.
We conclude that the use of a heat transfer coefficient
. equal to zero for this period is very conservative.
In the third phase of the LOCA, lower plenum flashing occurs. This is a flow phenomenon during the blowdown wherein a sudden transient increase in. the core flow begins a few seccnds af ter - the core flow has decayed to near zero. The increase in core i flow results when the liquid level in the vessel drops below. the recirculation line suction nozzles causing the flow out the break to change from a liquid phase to a steam phase and increasing the rate of depressurization of the system (to about 30 to 40 psi /sec). This rapid depressurization results in a rapidly changing U thermodynamic state of the fluid in the primary system. Because the fluid in the lower plenum beneath the core was ) initially in a subcooled state (by about. 24 Btu /lb), it does , not change thermodynamic state during early blowdown as does the rest of the fluid system; however, when the system pressure decreases to the level where this fluid flashes to 1 steam a large increase in steam flow through the core results. l This period of the LOCA is thus called " lower plenum flashing". l
-i Calculation of flows, temperatures and pressures during this phase depends on the knowledge of the flashing process, the i,
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. effact' of' flow maldistribution, the resistance to flow of a two-phase mixture through the core and jet pump dif fusers, and the rate of blowdown through the break.
During this period of increased core flow, GE assumes that nucleate boiling is . reestablished and that. relatively large heat transfer coef ficients result. Although nucleate boiling may be reestablished,' there is insuf ficient experimental evidence to support this assumption. Therefore, we asked GE to perform the analysis with the assumption that only stable film boiling occurs, with greatly reduced values of heat transfer coefficient, as determined by the Groeneveld correlation, during the period of lower plentan flashing. This assumption is in accord with Appendix A, Part 2 of the Commission's June 19, 1971 Interim Policy Statement on emergency core cooling systesw. Following the period of lower plenum flashing it is conservatively assumed that no convection cooling occurs. Heat generation, produced by the radioactive decay of the fission products, and thermal radiation among the fuel rods causes the core to heat up. The results presented in the applicant's August 4,1971, submittal show the calculations of fuel clad temperatures in the core for four fuel rod groups. We have reviewed the calcu-l 1ations for,this period and conclude that the predicted thermal l responses calculated during this phase are conservative. P b 2___ _ . _ _ _ _ . _ . _ _ _ _ _ _ _. _ L _
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.of the expected accident conditions and core configurations 1
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l as possible. The tests did not simulate the blowdown phase l , of a LOCA, but used the expected conditions at the time of initiation of the low pressure core standby cooling systems as initial conditions for the tests. Most of the tests were q run using top spray and bottom flooding concurrently. GE has developed a heat transfer correlation based on data i obtained in the stainless steel bundle tests. This correlation is reported in the GE topical report " loss-of-Coolant Accident and Emergency Core Cooling Models for General Electric Boiling Water Reactors" (NEDO-10329) . The correlation accounts for the reduced heat transfer in the central pins of an assembly , ( more realistically than previous correlations, predicts the
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clad temperatures more accurately, and forms the basis for applying the stainless steel test results to Zircaloy bundles. We have reviewed the GE correlation and its application, and a core spray model developed independently by the Aerojet Nuclear Corporation (ANC).* In both models, heat transfer i by thermal radiation from the rods is the controlling phenomenon in cooling of the rods, and heat transfer by convection plays
, only a small part.
- Formerly the Idaho Nuclear Corporation
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BFFITC~~A ~ wwutt wanu.e it vercs nv- er - The convection portion of the heat transfer is inferred from. the results of the tests on stainless steel bundles by sub-tracting the thermal radiation heat flux from the total heat flux 'as measured in the tests. The thermal radiation flux is calculated using the geometrical view factors and experimentally determined emissivity. In the GE' analysis a value of 0.6 was used for the emissivity of both the stainless steel rods and channel box. Subsequent tests made to measure the emissivity of the stainless steel bundles indicated that a value of 0.9 would be appropriate. Since the use of the larger value of test bundle emissivity results in smaller values of the empirical convective heat transfer coefficient, we have required CE to perform the analysis of the accident using convective heat transfer coefficients derived using the larger value of emissivity. This matter is treated in accordance with Appendix A, Part 2 of the Consnission's Interim Policy Statement on emergency core cooling systems. Since the cooling is mainly by thermal radiation from the fuel to the channel box, which is the outer container of the fuel bundle, the peak fuel clad temperatures reached are strongly influenced by the tempera-I tures of the Zircaloy chsnnel. Until the channel is wetted, by action of the spray, it is at a high temperature and is a relatively poor heat sink. CE and ANC use different heat
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transfer coefficients derived from the same data to calculate the cooling of the channel prior to quenching. GE uses.a correlation based on theoretical analysis to calculate the time to wet or quench. ANC estimates a larger quench time from , 1 available data. In order to assure that the channel quench times are conservatively calculated, we have required the calculation of the fuel rod temperatures using a channel quench time which is 60 seconds longer than that detemined by the theoretically-based (Yamanouchi) correlation. This require-ment is -in accordance with Appendix A, Part 2 of the Commission's Interim Policy Statement on emergency core cooling systems. The peak cladding temperature calculation assumes that cooling by spray action persists until the accumulation of water is sufficient to terminate the temperature transient by flooding action. The applicant and GE have performed t.hese analyses for the entire pipe break spectrum, up to and including a double-ended l severance of the largest pipe of the reactor coolant pressure bound ary . In the limiting case of a postulated double-ended break of a primary coolant system recirculation loop pipe with the simultaneous failure of the LPCI system, the calculated maximum fuel clad temperature is 1800*F, using the AEC evaluation model described in Appendix A, Part 2 of the Interim c r '{uh :
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- r Policy Statement. For smaller breaks the calculated maximum 2
clad temperature is less; for example, a break of 0.05 f t l would result in a calculated maximum clad temperature of 1400'F. l l The cladding-water chemical reaction is calculated to be less than 0.12% of the total amount of cladding in the reactor for-all break sizes. 6.1.3 Other Considerations The LPCI and HPCS pump capacities specified for the GE 1969 product line plants are lower than the LPCI and HPCI pump capacities specified for the 1967 product line plants. The various pump capacities are presented in Table 1.3.1. We requested the applicant to perform a sensitivity study of the LPCI and HPCS pump characteristics, the results of which are presented on page 6.5-1 of Amendment No. 12. These studies ] were not performed using the evaluation model set forth in l Appendix A, Part 2; however, the general trends are considered sufficiently accurate to evaluate the sensitivity of peak clad temperature to changes in pump characteristics. The effect of LPCI flow rate on peak clad temperature is shown in Figure 1 on page 6.5-4 of Amendment No. 12. As indicated , on this figure, an increase in LPCI flow rate from the design flow rate for Zimmer of 14,900 gpm to 23,100 gpm (the design ;
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[ A flow rate for '67 product line plants), a decrease in peak clad temperature of 150*P would result. The design basis LOCA peak clad temperature is insensitive to flow rating of the HPCS at hip,h pressure (1,000 psig) since the reactor vessel depressurizes quickly (about 50 sec) for a large break and about 30 see is required to start the HPCS pump. Consequently, the HPCS flow capacity for a large break is effectively determined from its icw pressure rating. However, the break size for which no clad heatup occurs is sensitive to the HPCS flow rate at high pressures. CE analyses show ; that the break size for which no temperature rise occurs increases by about 0.05 ft 2per 1,000 gpm of flow at 1000 psig. If the HPCS flow rate of 1,860 gpm @ 1,000 psig were increased to 4,250 gpm @ 1,000 psig the break size for no clad heatup would 2 increase from 0.2 ft2 to about 0.32 ft , We recognize that the changes in pump characteristics described above can improve the performance characteristics of the CSCS proposed for the Zimmer facility. Ob vious ly , larger changes in pump characteristics than those described above or the addition of more pumps could result in even greater improvement. However, we conclude that the CSCS proposed for Zimmer meets all of our criteria and thera. fore is acceptable.
' ^
ll EluSLON:Y-i The CSCS' pumps are arranged and located to assure that an adequate NPSH is available without reliance on primary containment pressure.- As an additional backup featu're the Zimmer ECCS includes a permanently installed, normally isolated crosstie to the service water system. 6.1.4 Conclusions We conclude that the design of the Zimmer emergency core cooling system is acceptable based on the analysis which shows that-the consequences of, the loss-of-coolant accident are such that (a) the calculated maximum fuel rod cladding temperature does not exceed 2300 F (b) the amount of fuel rod cladding that reacts chemically with water or steam does not exceed 1% of the total amount of cladding in the reactor, (c) the clad temperature transient is terminated at a time when the core geometry is still amenable to cooling, and before the cladding is so embrittled as to fail during or af ter quenching, and (d) the core temperature is reduced and decay heat is removed for an i extended period of time. j L 6.2 Residual Heat Removal System (RHRS) I l The RHRS is designed for five modes of operation as discussed below. j The system consists of two heat exchangers, three main system pumps, three service water pumps, and associated valves, piping, controls . 1
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J31C AdSE'[D 93 and instrumentation necessary to form three independent closed loops. The pumps are arranged and located to assure that adequate NPSH is available for all modes of operation. The RHR system loops are not interconnected so that a failure in one loop cannot cause a failure of another. In addition, two of the three . independent loops and associated equipment are located in a separate protected area of the reactor building to reduce the pos-sibility of a single physical event causing the loss of the entire RHR system. The separation provided by the above arrangements assures that the various operating modes of the RHR system are always functional. 'Ihese modes are as follows: (a) The shutdown cooling and reactor vessel head spray operating mode is placed in operation during normal shutdown and cooldown. The shutdown cooling system is espable of maintaining the nuclear system at 125*F during reactor refueling and servicing. The head spray maintains saturated conditions in the reactor vessel head volume by condensing steam being generated by the hot vessel and internals. This assures a high water level which limits thermal stresses in the vessel during cooldown. (b) The suppression pool cooling mode of operation is placed in operation immediately following the design basis accident (LOCA)
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. to limit the suppression pool temperature to less than 170*F. This l- suppression pool temperature provides assurance that complete condensation of blowdown steam is obtained.
l. (c) The containment spray cooling system provides additional redundancy to. the standby core cooling systems for post accident conditions. Spray headers located in the drywell and above the suppression pool are capable of removing the' energy from the drywell atmosphere by. condensing the water vapor. h e coolant spray is collected in the bottom of. the drywell until the level reaches the downconers vents and drains back to the suppression pool. A' spray ring in the suppression chamber cools any noncondensible gases collected in the free ~ volume above the pool. The spray-headers of the RHRS do not become operable until the low pressure
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conditions are met. (d) The RHR system operating in the condensing mode is only one method of operation of the reactor core isolation cooling system. The amount of decay heat that can be dumped to the suppression pool during RCICS operations is limited by the maximum allowable temperature of the pool (170*F). We RCICS turbine steam ( exhausted to the suppression pool results in a 3*F/hr temperature l-rise. Therefore, decay heat can be transferred to the service
- water instead of the pool by operating the RHR heat exchanger l
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ElfLiEON:y-as a direct steam condenser. The cool condensate is either dumped to the suppression pool or returned to the suction side of the RCICS pumps. 1 l (c)- The low pressure coolant injection (LPCI) system operating mode l is'used to maintain the reactor vessel coolant inventory following a LOCA. We conclude that the design of the RHRS is acceptable. 6.3 Post-LOCA Hydrogen Control i The applicant presented a discussion of the potential for accumulation l 1
)
of hydrogen in the primary containment following a loss-of-coolant l accident and the proposed techniques for limiting the concentration to preclude the potential hazard. In response to our question 5.10, the applicant cited " Hydrogen generation in BWR's" as discussed in .f l Dresden 3, Amendment 23 as a basis for including provisions for venting the containmht through the atandby gas treatment system. They also stated that they are following the program underway by the General Electric Company to develop a hydrogen-oxygen recombiner for resolution of potential hydrogen hazards. There is presently no commitment by the applicant to install recombining equipment in the containment. However, they have stated that the station design veuld not preclude the addition of a recombiner to dispose of radiolytic hydrogen if venting is deemed
'1 unacceptable or if it becomes a requirement.
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.The applicant is aware that we require at the construction permit stage the design bases and conceptual design of systems, other than venting, that are intended- to sample, mix, and control the concentration of hydrogen.in the' containment following a LOCA due to metal water reaction and radiolytic decomposition of wster.- For. design purposes, the appli- -j cant vas requested to' utilize hydrogen generation assumptions consistent with the assumption given in AEC Safety Guide 7. In response, the I applicant stated that the assumptions used in AEC Safety Guide 7 were too conservative. The applicant also reaffirmed their position that venting is an acceptable means for control of hydrogen generation.
This matter remains unresolved. 6.4 containment Inerting During a meeting the applicant stated that inerting the containment as means for the potential solution to the hydrogen problem will detract from the safety of the facility due to the reduction in containment accessibility by pe rs onnel . Accordingly, the use of nitrogen to inert the containment during reactor' operations would not be adopted. . In a letter dated April 9,1971, we infor1oed the applicant that we have concluded that provisions for inerting the primary containment should be included in their design. This design provision is consistent with action taken on all previous BWR plants. In response the applicant stated that provision for inerting the primary containment would not be precluded from their design and reaf firmed design position that the 4FFCIAHJS?0NtY UlW - _
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. containment would not be inerted unless we require it. We intend to inform the applicant that use of the inerting equipment will be l required.by the Technical Specification.
I 6.5 Long Term Cooling Water Supp12 f Ileat removal capability during operation, for safe shutdown or during the accident mode of operation (including loss of outside power) is 1 provided by means of the combination of the Service Water System (SWS) and the Reactor Building Closed Cooling Water System (RBCCWS). The Service Water System will aise remove heat rejected by the Reactor Building Closed Cooling Water System. The SWS consists of four motor-driven horizontal centrifugal pumps, which take suction from the Ohio River. Three of the four pumps are required for normal operation, while one pump is capable to shutdown the plant and keep it in a safe condition af ter a loss-of-coolant ! accident. In Amendment No.15, the applicant modified the design of the SWS to provide for a split header system to assure adequate cooling capability even in the event a failure of an active or passive component is postulated. j i The RBCCWS is a heat sink for various essential and non-essential equipment located in the reactor building and in the auxiliary bev. l Three 50% capacity pumps and two 100% capacity heat exchangers are l l
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provided. In Amendment No.15, the applicant modified the design of RT,CCWS to provide for a split header system to assure adequate cooling I in the event of postulated failures of pressure components. One pump and one heat exchanger are sufficient to provide the cooling capability for plant shutdown following a LOCA concurrent with loss of off-site power. The RBCCWS will be monitored continuously to detect leakage from reactor associated systems and components. The SWS and RBCCWS which supplies essential equipment will be designed and constructed in accordance with Class I (Seismic) and Quality Group C requi rements . We have concluded that the provisions made to assure a continuous 1 I supply of long term cooling water in the event of a design basis accident are adequate. In addition, the potential for leakage of activity t'o the Ohio River is acceptably low. I
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r . l i JEEICbOSHlNtY--- 99 _- ' 7.0 PROTECTION, CONTROL AND EMERGENCY ELECTRIC POWER SYSTEMS ~
- 7.1 General' our. review encompassed the reactor t rip and control systems, the I engineered safety feature circuits, and the emergency electric power system.
. The Commission's General Design Criteria (GDC), the Proposed IEEE Criteria for Nuclear Power Plant Protection Systems (IEEE-279) dated August 1968. . and Safety Guides 6 and 9 served, where applicable, as the bases for evaluating the edequacy of these des igns .
The reactor trip system is substantially the same as that of the Pilgrim Nuclear Power Station. However, the reactor control system and emergency core cooling system instrumentation and controls i are new and represent the first of- a new product line design (69 product line) of the nuclear steam supplier, General Electric Company. The energency electric power system design is unique to this s tation.
. Our review has emphasized protection system generic items, new design areas, and items unique to the Zimme r Station. Specifically, the review encompassed the following:
A. Protection System Generic Itean '
- 1) (
Independence of Redundant Otannels
- 2) Capability for Pe riodic Testing 3)
Incident and Accident Surveillance Ins trumentation.
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B. New Design I*cas
- 1) Control Sys tems a) Solid State Manual Control System b) Solid State Rod Position Circuitry
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c) Recirculation loop Valw. Control System.
- 2) Reactor Trips a) Deletion of Condenser low Vacuus Trip and Addition of APRM St art up Re actor T rip at 15% Power.
- 3) Core Standby cooling System Initiation and Control Ins t rume ntation
- 4) Annunciation of Engineered Safety Feature Bypasses.
C. Items Unique to the Zimmer Station
- 1) Eme rgency' Electric Power Sys tens
- 2) Physical Separation Criteria for Protection and Emergency Power Systems
- 3) Seis mi c , Radiation, and Environmental Tes ting.
- 7. 2 Protection Systene Generic _ Iteen 7.2.1 Independence of Redundant Oiannela One of the applicant's safety design bases (SDB) requi res:
"'lhere shall be sufficient electrical and physical separation between channels and between logics monitoring the same variable to prevent environmental f actors, electrical t ransients.
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and physical events from impairing the ability of the system to respond correctiv." Our review of the PSAR. revealed several areas of the protection system design that do not s atis fy this requi rement, nor those of IEEE-279. These areas fall into two general categories. The firs t category concerns cohnection of redundant protection channels to single switches and terminal boards in control room panels. The second cat ego ry concerns location of redundant protection system switches on control room panels within a few inches (less than 6 inches) of each other. This comnromise of protection channel independence increases their vulnerability to single random failures. Addi tionally , a basic premise of all conanon mode f ailure studies is the neces sity for independence, both physically and electrically, of redundant protection system channels. i We have asked the applicant to review his preliminary des ign in this regard and propose design modifications that will assure the maintenance of redundant erotection channel i ndepe ndence. The applicant's response is that the evaluation as presented in the General Electric Company Topical report "Com11ance of Protection Systems to Industry Criteria" (NEDO 10139) shows ORBAttr00 l //, _ _ _ . _ _ _ _ _ _ _ . . - - - - - - - - - - - - - - LJw T,
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( i that single failures do not preclude automatic protection system action. The applicant has identified four instances where redundant protection system channels are connected to a single switch. He has not responded with respect to minimum physical separation of redundant protection system switches and components on panels. This mat ter remains unresolved. We plan to require that the design be considerably improved in this respect over that found marginally acceptable in recentiv 5 reviewed BWR operating license applications (e.g. , Pilgrim Nuclear Power Station Docket No. 50-293), since Zimmer is the first of a number of new plants that are expected to use the new GE Protection Sys tem design. 7.2.2 Capability for Periodic Tes ting Past BWR designs did not include in all areas of the reactor protection and engineered safety feature circuits the built-in capability to permit periodic tes ting at power. The designs were marginally acceptable because of backfit considerations, added redundancy, de tailed written procedures requiring manual interference with circuits (removal of fuses), or the perturbing of reactor parameters to assess the operability of the instrumentation. N hj y a _ _ _ _ _ _ _ _ - _ _ _ ____-_ - - - _ - 1
I I L WIALUSFON:Y' l 10 3 - The reactor protection system and the engineered safety feature circuit ' designs for Zimner will include the inherent capabilf ty for complete periodic testing during power operation. An area of doubt was removed when the applicant stated that the p rinare containment isolation system design is changed from that us ed in previous BWR's. The applicant will document,the preliminary design in a forthcoming amendment. We will rev' iew this new design and report orally at the ACRS meeting. 7.2.3 Incident, and Accident Surveillance Ins t_runentation The BWR reactor protection and enginee red sa fety feature instrumentation channels generally use blind sensors and, therefore, do not provide continuous readout in the cont rol room of the parameters being monitored. The neutron monitorine and main s team line radiation monitoring svstems are excep tions . The other vital parame te rs, howeve r, are monitored by ins t rtment channels associated with control systens. As such, these information readout channels are not desianed to satis fy protection system criteria and are not included in the technical ; I s pe ci fi cat ions . j Information readout channels are requi red by the operator to assess plant conditions subsequent to an accident in order that he may determine whether to inte rvene in the operation of the Automatic Depressurization System (ADS) or to initiate the p';lCPUStDNQ dhh 1
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I I l Containment Spray System (CSS) . Addi tionally , the recent Dmaden t' nit 2 incident emphasized that thin re adout and I I recorded infomation is needed to mitigate and assess the cor.secuences of operational occurrences. l l The information received from the applicant on surveillance int.trumentation was not complete. We advised the applicant th at , an a minimum, we would mquire the monitoring, readout
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and recording of the following paramete rs in the control I room:
- 1. Full range drvwell pressure
- 2. Full range reactor water level 4 l
- 3. Reactor pressure I
- 4. Torus water level
- 5. Torus water temperature
- 6. Drywell temperature
- 7. Neutron monitoring systems
- 8. Control rod indication
- 9. Im ak detection sys tem (portions) .
The applicant was further advised that we will require . I I redundancy of instruments monitoring each narameter, that the ins truments be capable of withstanding the environmental i l 1 1 m m.s W __ _ l
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l 1 conditions resulting from anticipated operational occurrences 1 l without loss of function, and that they otherwise compiv with the requirements of IEEE-279. We consider diverse means for monitoring plant conditions (such as rod sosition determination by neutron monitoring and rod position indicators) as meeting the requirement for redundancy. Although this position is similar to that taken on Vermont Yankee. Newbold Island and Lime rick, the applicant does not agree with our position. 7.3 New Des ign I te m_s 7.3.1 Control Sys tems A. The reactor manual control system acts on the elect ro-hydraulic control rod drive system to position control rods in the core. This system has been modified from relay logic control circuitry to self monitoring solid-state ci rcuit ry . The elect ro-hydraulic drive system, howeve r, re mai ns the s ame . This change in design will substantially reduce field wiring and should improve operability th rough the self monitoring feature being incorporated. Our evaluation is primarily concerned witt, assuring that the reactor protection system is capable of preventing fuel damage with control system failures or during operational occurrences. _g g g E S? 0 R ~~~~ Ut2W
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1 The applicant has. identified the primary design changes l l l in the proposed solid-state circuitrv from those for the previous BWR relay logic design. He has s tated that single f ailures will not result in t ransients that are beyond the ; l cap ability of the reactor protection system. Additionally. he has described the prototype qualification tests that have been completed and those still to be performed. We have concluded that the applicant's design criteria an'd I I tes ting program are satis factory . The technique of using 1 sequential coding signals is new to the nuclear power plant control systein designs and therefore we will perform an in-depth evaluation of the design during the operating license review. B. The control rod position information circuitry has been ch anpe d f rom rel ay logic t o s olid-stat e circuitry . Agai n the motivation for this change is the reduction in field wiring and inn ta11ation p roblems . The rod position 1 de tection and display remain the name except for an i apparent increased reliance on the plant computer. The applicant's response to our concern regarding the significance of the computer to plant safety was to state
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that it has no safety significance. Discussion did reveal th at reliance on the computer as an information bank and operational aid to the operator appears to be increasing. l We nian to investigate this aspect of the use of' the computer further during the operating license review. C. The rectreulation loop flow in previous BWR design was I varied by a variable speed pump motor. This system is ! being replaced by a constant speed pump, a control valve, i bypasa piping, and additional valves to permit isolation and/or maintenance. The circuitrv for this new system ! di f fe rs in two maj o r areas . The first area concerns the tase of APRM flux (protection system) signals to derive an error signal from the mas ter power controller which is fed to the flux controller. The flux controller then changes flow in each loop (and reactor power) by modulating the control valve. The second area concerns the design of the recirculation pump and loop isolation valve interlock ) I ci rcuit ry . j i Our review of the information presented in the PSAR with rer.ard to the use of APRM flux nignals for control purposes revealed that if the preliminary design is proper 1v implemented in the final design, the requirement of , 1 Section 4.7 o f IEEE-279 would be satis fied. 1 l' J tbs ~ g
01UllJHNT 10 8 - The applicant's crite ria governing the design o f the recirculation loop pump and isolation valve interlock ci rcuit ry are not yet formulated. This matter in discussed in Section 4.2.4. 7.3.2 Reactor Trips The applicant proposes to delete a reactor trip derived from instruments monitoring condenser low vacuum. Condenser low vacuum reactor trip is provided in this design by an indirect me thod. Condenser low vacuum initiates closure of the turbine s top valve which, in closing, t rips the reactor. Th e applicant was requested to provide information to assure that the condenser vacuum instrument system will satisfy the req ui re-ments of IEEE-279. This will be reported on orally. i Additionally , the applicant plans to modi fy the APRM channels i by extending their e effectiveness into the startup range and ) including a reactor t rip at 15% power. In previous BWR designs . the APRM channels were made e ffective only in the Run mode. The applicant has stated and will document that these changes have no safety significance and are being included in the design to preclude the necessity for recalibration of the Intermediate Range Mont tors (IRM), due to changes in responses res ulting from control rod motion in the vicinity of the IRM cQ h war '
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10 9 - y de te ct o r. This commitment is satisf actory for the cons truction spermit review and will be followed closelv during the operating,
, li ce ns e revi ew.
y . y .7.3.1 Core Stan,dby Cooling System (CSCS) Initiation and Control Ins trume nt ation The. criteria that gove rned the redesign of the ECCS were s tated
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by the applicant to be IEEE-279 and -308: t he rene ral Den t.m Cri te ri n : and reneral Elect ric's criteria for separation, e t es tab i li ty , and qualification testing. The ECCS is to be divided into three redtodant divisions such that the loss of any one will not preclude adequate core cooline over the complete range of break sizes. This arrangement will . i f p rope rly implemented, result in a balanced 3 diesel - 3 bus elect rical distribution system. Additionally, the initiating i sensors are uniquelv assigned to each division thereby avoidi ng the interconnections found in past desicna and assuring separation and independence of these redundant divisions . To provide further unurance of neparatton, the location o f sensors with rennect to elevations and azimuth around the containment was considered. The relays and logic circuitrv vill be ,
} arranged in three panels, one for each division.
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This new ECCS design contains diversity in both the initiating s ign als and t he subsys te ms . We conclude that this system can be designed to the criteria and is acceptable. 7.3.4 Annunciation of Engineered Safety '1sture Bvnass Our review of preliminary design revealed that annunciation o f the bypass of an engineered safety feature resulting f rom a deliberate operator action was not included. The applicant s tated that he considered administrative control as an e ffective and adaquete means to identify these bypasses . We do not consider administrative controls an acceptable substitute for the indications required by I EEE- 279. We, thereforn , vill requi re that the design of these circuits include control room annunciators to indicate whenever operator actions result in the loss of an ESF function or a reduction in system redundancy. This mat ter remains unresolved. 7.4 Eleet ric Power Sys tems 7.4.1 Emergency, Elect ric Power System 7.4.1.1 Offsite Power The Zimmer Nuclear Power Station will be interconnected to the t transmission system through 345 kV circuits. Power f rom the unit 's ge nerator is fed via a ningle circuit containing the main atep-un t rans former to the 345 kV switchyard. Th e 345 kV switchyard in arranged in a b reaker-and-a-half confi guration.
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11 1 - Two 345 kV transsiission circuits emanate from the switchyard. A third 345 kV circuit will be added prior to plant operation. The 345 kV system (through the switchyard and reserve auxiliatv t rans former) represents one source of of fsite power to the engineered safety feature and safe shutdown loads. This offsite newer source is automatically connected to the emergency buses supp1ving these loads. A second source of offsite nower will be supplied from the 69 kV t transmission system via a circuit which connects to the energency bus es through a separate reserve auxiliary t rans fo rme r. This second source of offsite power is also capable of supplying all e ngineered s afe ty feature and s afe shutdown loads . , The applicant has completed studies concerning the stability of his system with respect to the effects of the loss of the unit of this s tation or the largest remaining generatine unit. From these s tudies he has concluded that offsite power to the e nginee red s a fe ty feat ures will not be los t. Our review of the applicant's response to cues tions concerning l the mutual independence of the offsite power circuits between ; ! 1 the grid and the emergency b us es reve aled that a ningle f ailure i l V B ArUSt0iu . J u k' !< b - a
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of a 345 kV line or tower could also result in the loss of the 69 kV circuit, since a portion of the 69 kV circuit crosses beneath the 345 kV lines. Our review indicates that, except for the lack of independence of the o f fsite power t transmission lines oreviously discussed, the offsite power systems satisfy the CDC 17 and are accentable. We will report orally on the resolution of this matter at the ACRS seeting. 7.4.1.2 Onsite Power The design of the standby (onsite) emernency nower system utilizes the solit bus concept in accordance with Safety Guide 6.
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The redundant engineered safety feature equipment t rains are divided into three divisions with each division assigned to an eme rgen cy 4 ,160 volt b us . One diesel generator set is connected t o e ach b us . The three redundant diesel generator sets are each individually housed in a seismic Class I room such that th e re is complete physical separation between units. Auxiliary systems for each of these machines are physically and elect ri-cally independent of each other and houted in Class I roons . The fuel suppiv system will be of suf ficient capaci ty for operating all three diesel engines at req ui re d l oads f o r 7 days . c:FF 3 A. I GLY Oli~' y
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The applicant has stated that the criteria for selection of diesel generators will include the regulatory positions 'of S a fe ty cuide 9. We expect that the diesel renerator requi red to power the HPCS will not meet the maxisssa allowable voltage drop s tated in regulatory position 4 of' Safety Guide 9. We have informed the applicant that if this is so we will require that the applicant either substantiate that the starting reliability, equipment f atigue and wearout are equivalent to the other plant diesel generators or obtain a diesel gener stor with suitable capability. Three d-c systems are to be provided. The 125 vole d-c sys tem will consist of three redundant batteries, each with its cwn charger and distribution system. Further, each batterv is located in an independent Class I room with separate venti-lation systems . A separate portable battery charger will be provided which can be substituted manually for any normally ins talled ch'arger. The batteries will have a capacity to supply all assigned loads for a minimum of one hour. Th is 125 volt system is designed to be compatible with the three-division engineered a sfety feature load grouping discussed p re vio us ly . I i 1 l
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- The remaining d-c systems consist of 1) two redundant 250 volt battery systems and 2) two redundant 24 voit batterv sys terms . The 250 voit system will supply power to larger d-c loads such as ptimps and large valves. These redtmdant 250 volt battery systems will be assigned to two of the three engineered safety feature load divisions such that single f ailures will not preclude minimum safety engi neered I fe ature operation. The two redundant 24 volt battery systems supply power to the neutron monitoring system. The 250 voit and 24 volt battery system designs and installation will be identical to that of the 125 volt system.
The applicant has stated that he will document in a forthcoming amendment a corsnitment to meet the requirements of GDC 18 with { respect to providing the capability for periodic tes ting and inspection of his emergency power systems. We conclude that this is acceptable for the construction permit re vi ew . Our review indicates that the standby (onsite) pcuer systets (both a-c and d-c) satisfy the regulatory positions of Safety Guides 6 and 9 and the requirements of GDC 17 and are accep tab le . I I
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- 7. 4 . 2 . Cable Ins tallation Crite ria for Protection and Elect ric Power Systenn Our review of these criteria have found them to be acceptable l
in all areas except cable t ray seismic criteria. The applicant l is reviewing the design to assure that these criteria are i ncluded . We have informed the applicant that their criteria should be documented for review and evaluation. We will report orally in this regard. 7.4.3 Seismic, Radiation, and Environmental Tes ting Our review of the instrumentation, control and electrical system environmental testing criteria ascertained that all safe ty related devices supplied by General Electric would be seismically tested with the exception of the primary pressure boundatv devices which would be analyzed. The applicant has subsequent 1v agreed to test or provide test data for all devices supplied by S arge nt and Lundy and the primary pressure boundarv devices. This , when documented, will be an acceptable connitment. The criteria also includes the qualification testing of all s a fe ty related equipment located inside the primary con t ainre nt at the extreme combined tenverature and oressure conditions of the design basis los s -o f-coolant acci de nt . We have reviewed the criteria and conclude that they are acceptable. E DIA rilS N N;Y - 9s~ x
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' The criteria for radiation qualification of all safety I
related devices have been reviewed. We conclude that they a re l the same as those for previously licensed niants and find them " acce pt ab le . l 1 l
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8.0 RADW AS TE _S_Y,S, TEM, 8.1 Ge_ne ral The applicant has described the system to be provided for the Zimmer f acility for t reatment of gaseous and liquid radioactive was te discharge to reduce the radioactive e ffluents such that the annual averace con-centrations are less than 1% of 10 CFR Part 20 limits for each systec . The liquid and solid waste disposal systems nrovided are substantially the same as those systems orovided for other BWR's. The gaseous radio-active vaste disposal system differs in that an additional cas eous hold-up system will provide a minimum decay of 24 hours for krypton isotopes and 18 days for xenon isotopes prior to their release to the atmosphere. At the present time , the applicant has no t provided suf ficient informa-tion on the additional holdup system proposed for the gaseous dis pos al sys tem to enable us to determine independent 1v if thev can meet the limits set on 10 CFR Part 20 above (see paragraph 8. 3 for det ails) . 8.2 1.1, qui d Was tes The liquid radwas te system collects , monitors , processes, s tores , and provides cont rolled discharge of all radioactive liquid was te. Th e system components consis ts of s torage t anks, vas te demineralizers , filters , and evaporatom similar to those used on other BWR f acilities. The liquid radwns te system is divided into three main subsys tems , h i rh p u ri ty , low p u ri ty , and chemical waste so that waste f rom va ri o us
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sources can be kept segregated for separate processing. Cross connections among subsystems nrovide additional flexibility for processing the waste by alternate methods. High purity (low conductivity) liquid was tes are primartiv collected - from equipment d rain stunps . Liquids from this source are processed by filtration and ion exchange through the waste filter and det:iner-alizers and then transferred to the condensate s torace tank for reuse as makeup water. Reprocessing of high conductivity or highlv radio-active liquida is accomplished by the demineralized train or by the radwas te evaporato rs . Law puri ty (hich conductivity) wastes from floor drain sumps are collected in the floor drain collector tank. These was tes generally i have low concentrations of radioactive impurities and are processed j in the same manner as high puri ty was tes . Chemical and detergent wastes normally are processed through the radwas te evaporators . diemical distillate is demineralized before being discharged to the 011o River and the concentrates will be processed by the solid radwaste system. Deterrent was tes are t rerted with anti-foami ng agents p rior to evaporation. The distillat e is trus e d i n th e l aun dry and on2v the excess in discharged from the nlant, j The plant is designed so that operation of the was te disposal system l is by manual s t art and automatic closure of discharge valve on high
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h : - 1 19 - radiation s tenal. Protect ion agains t accidertal discharge is provided hv design redundancv. ins trumentation and alarms , and procedural con t rols . Liquid wastes that are discharged to the environs are handled on a batch basis through the plants circulating water discharge n i pi ng. Th e liqui d b at ches are held for a period o f time to allow comnle te mixine, s ampling , analyzing, and processing orior to the t rans fer to the condensat e a torage t ank , solid radwas te s torace , o r to the e nvi rons . Based on the anplicant's design criteria and ,rocess objectives to allow for maxinum reeveling of liquid radwaste, and on the apolicant 's , analyses of concentration, we conclude that the activity released from the Zinner's liauid radwas te system will be less than 1% o f the 10 CFR P art 20 limits. 8.3 Caseous Was tes The off-gas t reatment system of the plant is designed to reduce the p as eous radioactive wastes released f rom the station. Caseous was tes in BWR's are generated f rom two main sources : noncondensible radio-active gases removed from the turbine gland seal condenser; and noncondensible radioactive gases remove d f ron th e mai n con de ns e r by the ai r e jector. Th e Zi mme r gas treatment system has incorporated an additional decav system in addition to the normal 30-min. holdun system. Th is repre.sents a significant change in design from orevious BWR designs . j L" 7 = . g i n,w
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I The' of f-gases f rom the turbine gland seal condenser require holdup time for decay of the radioactive constituents. This gas is removed by' a mechanical vacutra pump. Exhaus t from the pump and the off-gas from the . turbine gland seal enter a coninon two minute holdup pipe before being discharged to the of f-gas vent. The air ejector off-gas normally will contain activated gases, radio-active noble gases, and radiolytic dissociated hydrogen and oxygen. The off-gas treatment system will include a high-temperature catalytic recombiner to combine the dissociated hydrogen and oxygen continuously from the 'of f-gas s tream of the air ejector. The of f-gas e ffluent from the recombiner is condensed and passed through a 30 minute holdup for decay of short lived isotopea. The off-gas is nrocessed further by a high-efficiency filter (HEPA) before it enters the additional decay system that provides a curie reduction factor necessary to meet their design basis of the system. Before being disch rged from the of f-gas vent, it is passed through a final HEPA filter. The radiation levels at the off-gas vent are monitored continuously by two detectors.
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Should the activity of the gas being released exceed the reduction j capabilities of the processing system, an isolation valve is I automatically closed to prevent further flow of gaseous ef fluent to the environment. The applicant has stated that all equipment will be used to the fullest extent possibit: at all times to achieve their
- design basis of the system (1% of 10 CFR Part 20) .
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As a design basis for the gaseous waste system, the applicant has used a noble gas input equivalent to an annual average off-gas rate of- 100,000 pC1/see based on a 30-min decay oniv. The additional holdup system will provide for complete decay of all isotopes except long lived krypton and xenons, thereby significantly reducing the release of radioactivity to the environment. Two types of additional decay systeen are being considered by the ap pli cant . One is an ambient-temperature, charcoal bed system and the other is a cryogenic temperature system. The applicant s tates that either sys tem is capable of meeting their design basis. At the present time we have insufficient information to perfors' an independent calculation to permit us to reach a conclusion as to its adequaev. However, as a result of additional information (the design of Newbold f i Island and Limerick off gas system), we expect to determine that the applicant is capable of designing an additional holdup system so that , I the release of radioactivity from the station will meet 1% of 10 CFR Part 20. We have informed the applicant that we require additional information on the design of the decay systems. He has agreed to provide this information when the final design is selected but before l any irrevocable commitments are made. We conclude that this commitment I 1 is acceptable. I I 8.4 Solid Was tes The solid radwan te sys tem collects , monitors , proces ses, packages and 4m;AAlSE0N:Y s w0r' ~J
e 1180ALUSE-0N+ 122 - providse temporary storage f acilities for wet and dry radioactive solid waste for off-site permanent dispoeal. Wet wastes resulting from spent filter demineraliser, deep bed demineraliser resins, filter sludge, and evaporators concentrates. are passed to waste tanks that - serve as s torage and batching f acilities. Dry waste consists of air filters , paper, rags , contaminated clothing, tools and equipment patts . Compressible wastes are compacted to reduce volume and shipped in SS-gallon s teel drums as is the non-compressible was tes packaged manually. Design and operation of the solid radwaste system do not involve any musual safety problem not already previously considered on other BWR applications. ; amCIA-llEON:Y , i a, s - ,
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9 .0 Auxiliary Sys tems 9 .1 Gene ral The design bases, safety considerations, and inspection and testing requirements of the auxiliary systems have been reviewed and we have concluded that the design concepts are acceptable. The following systems were found capable of performing their intended function: Tool and Servicing Equipment; Fuel Pool Cooling and Cleanup System; Reactor Building Closed Cooling Water System: Turbine Building Closed Cooling Water System; Fire Protection System: He ating, Venti-lating and Air Conditioning System; Instrtsment and Service Air Systems ; Portable and Sanitary Water Systems : Equipment and Floor Drainage Systems: Plant Process Sampling System: Communication Systems; Lighting Sys tem: Heating Boiler; and Hake-up Water Treatment System. 9.2 New Fuel Storage The storage racks for new fuel are sufficient to hold 30 percent of the full core fuel load. The s torage racks are designed and main-tained with suf ficient spacing between the new fuel assemblien to assure that the array, when fully loaded, will limit the ef fective multiplication f actor of the array (K,gg) to not more than 0.90. The fuel storage racks will be designed to Class I seismic requirements . I We conclude that the design provisions are acceptable. ) i I
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L r 9.3 Spent Fuel Storage 1 The spent fuel storage racks provide specially designed underwater storage space for spent fuel assemblies requiring shielding and/or cooling prior to shipment. Pool storage space to acconnodate 160 percent of the full core fuel load is provided. The spent fuel racks design provides for a suberitical multiplication factor (Kgg) for normal and abnormal s torage conditions . For normal conditions, K,gg f is equal to or less than 0.90. Under abnotaal conditions, such as l dropping equipment on the racks, earthquakes and accidents involving refueling equipment, K,gg will not exceed 0.95. The spent fuel pool is lined with stainless steel to lindt the possibility of pool leakage around penetrations. No inlets, outlets , or drains are permitted that might allow the pool to be drained below j i approximately 10 feet above the top of the active fuel. Lines extending below this level are equipped with syphon breakers, check valves, and other suitable devices to prevent inadvertent pool drainage. The cool I is provided with' interconnected channel drainage paths behind the J liner weld joints to prevent pressure buildup behind the line plate, j and uncontrolled loss of contaminated pool water. Provision is made I for liner leak detection and measurements. A separate spent fuel ]
. shipping cask a torage area is provided adjacent to the spent fuel f
- pool. An interconnecting canal between these areas allows for under-l water fuel trans fer to shipping cask. The cask storage area is l
l
-0VCh0SF0N:Y l UDF - i
n 05LUSi-0N Y- I 125 - constructed of reinforced concrete and lined with stainless s teel and is capaMe of withstanding the cask drop ac ci dent . . In addition, a cask accident within this area will not affect the spent fuel pool s torage . The spent fuel s torage racks, spent fuel pool and the spent fuel , I I sh'ipping cask s torage area are designed as Class I seismic s tructures . The transferal of the crane and the shipping cask over the fuel pool ! will be prohibited by appropriate interlocks. We have concluded that the design of the spent fuel storage facility meets the intent of Safety Guide Number 13 " Fuel Storage Facility De s i gn Bas is " and , there fore. is acceptable . 9.4 Se rvice Water Sys tem The service water system (SWS) is designed to supply cooling water during normal clant operation and for all emerrency conditions . Se rvice water is supplied to the following systems', equipment , and facilities located in the reactor building and the turbine building: Reactor Building Closed Cooling Water Heat Exchange rs : Res idual He at Removal Heat Exchangers : Turbine Building Closed Cooling Water Heat Exchange rs : Turbine Generator Oil Cooler Heat Exchange rs : Turbine Ge ne rato r Hyd ro ge n Coole r He at Exchange rs : Tuttine Cenc rator St ator Wate r Cooling ilent Exchange rs : Diesel Generator Cooling Hent l l Exchangers : Service Connections in Reactor Building and Turbine l l l B uilding : and Radwas te Facility. l
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.1EEICIALUSMNtY- .
126 - E - 3. The S:4S consists. of four horizontal' centrifugal pumps;' three of the four are required for normal operation and only one is . required for safe shutdown af ter an accident condition. The SWS pumps are separated
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j into two ' flood-proof cubicles to maintain independence. The applicant's split header system design provides adequate cooling capabilities in the event of a single failure of an active or passive component. Service water will supply the cooling requirements for the control
. room air conditioning equipment and the air conditioning equipment for the auxiliary equipment room following a combined generator. trip and total lose of off-site a-c power or a loss-of-coolant accident.
The SWS and related structures are Class I seismic design, capable of safe shutdown at river flood stages, and operable at low river flow rates' (see Section 2.3 Hydrology for details). The capability of the service water system to supply long term cooling water following a i loss-of-coolant accident is discussed in Section 6.0 of this report. Based on the above information, we conclude that the SWS is designed to supply adequate cooling water for all emergency conditions and there fore, is acceptable. l
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l i l 10.0 Station Structures and Shielding a
'10.1 Classification of St ructures and Equipment The applicant defines Class I structures and components as those whose failure might cause or increase the severity 'of a loss-of-coolant accident or result in an uncontrolled release of excessive amounts of radioactivity or whose function is vital to safe shutdown and isolation of the reactor. The Class I structures are designed to withstand the applicable seismic loads resulting from'the seismic events and other applicable loads without loss of function. Class II structures and equipment are those that are inmo rt ant t o reacto r operation but are not essential to the safe shutdown and isolation of the reactor and whose failure could not result in a significant . release of radioactive material. Class III are those that are not essential to the operation, shut down or isolation of the reactor and whose f ailure cannot result in the release of radioactive material.
These de finitions are considered to be acceptable for identifying those vit si s t ruct ures and components essential to reacto r s a fety . The potential interaction of Class II or III s tructures or equipment with Class I structures in discussed in Appendix I.0 of the pSAR. This appendix has been reviewed and its criteria were found to be acceptable. 41CK.-USF0tY __ Wau- i
0-h"Jl0A_lsLON:P l h- - 128 - a 10 .2 St ructural Analysis and Design 10.2.1 St ructural Analysis The containment analysis will be performed utilizing plate and shell analyses. SOR-II developed by Knolle Atomic Power Laboratory. KALSHEL developed by Prof. Kalnins of Lehigh Unive rsity, PLFEM and TEMCO developed by Sargent & Lundy as proprietary programs and AFEM which was originally developed by Rohm and Haas will be used. The anolicant has a tated l that all of these programs have been validated by Sargent & Lundy by selecting appropriate elasticity problems for which the solutions can be obtained by manual computation or other l l acceptable methods and comparing the computer solution with the es tablished solution. On this basis we conclude that th e analysis methods are acceptable. 10.2.2 St ruct ural and FoundatioLonsiderations_ C The p l ant i s t o b e s ited on th e e as t b ank o f the Oh io Ri ve r on app ro ximately 100 feet of clayey silts underlain by' limestone b edrock . The major plant s tructure will be supported on mat ty pe foundations which will be placed on engineered fill. Such foundation structures are considered adequate to trans fer the loads to the underlying soil. l 1 i
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0FBC LUSE-0N.N-i - 129 - For the prestressed concrete portion of the containment - a tructure , the applicant has specified acceptable allowable s tresses as applicable to the various design loading combinations including seismic loads. For the yield limit loading combinations the applicant has specified the load f actors to be utilized. The other Class I s tructural compo-nents and s tructures will be designed to specified allowable
-stresses under working stress design methods and checked agains t yield limits for load combinations that include the tornado or design basis earthquake. These procedures , in (
I Reneral, comply with rules of ACI 318-63, except in certain j l de t ails . The applicant has stated that the more conservative l rules contained in a proposed revision of this code published f
. 1 in the Journal of ACI dated February and September of 1970 ]
I will be used. { l The entire containment will be pressure tested to 52 psig which in 15% higher than the design pressure . The drvwell floor will be pressure tested initially to demonstrate its l capability to withstand 25 psi differential pressure. Th es e t es ts are consistent with the practice adopted on orevious BVR applications such as Shoreham. The containment will be ins t rtvoented with s tress mete rs , s t rain me te rs , and l J
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de formation meters along one meridian and at the equipment h at ch . The s truct ural i ns t rumen tati on fo r us e b e fo re and during the structural acceptance test will consist of s tress and s train meters along with a system for measuring de fo rmat ions . These instruments generally will be placed on . two meridians with additional instruments at the large hatch. This ins trumentation program is judged to be adequate. A statistical evaluation program is underway by the applicant's engineers , Sargent and Lundy, in order to j us tify the t endon ! surveillance program. We intend to continue to work on the review and the satis factory solution to this problem in the noet-CP follow-un. The TSAR will contain the results of I' the anplicant's investigation and resulting surveillance n rogram. We~ consider this to be accentable. The Class I s teel-framed s tructures will be designed in accordance with the AISC Specification (1969). This speci fi cation provides design rules for the working s tress j method of analysis (Part I of AISC) and the plas tic design me thod of analysis (Pa rt II of AISC). We have considered the safety margins associated with these design me thods for s teel and concrete s t ructures and have found them to be adequate for the fomdation at this site.
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.10.2.3 Tornado Analysis 1 l
Design wind loads are based on the' 100 year recurrence as defined in ASCE-3269. These wind loads are conputed to produce 53 psf for heights greater than 150 feet. The 'l effects of the design basis tornado are translated into forces on the structure by the tornado model. The model consists of ' the simultaneous application of a 300 mph
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rotational velocity, and a 60 mph transnational velocity with a 3 psi pressure drop in 3 seconds. The velocities considered by the model are converted into a equivalent static pressure using the equation in ASCE paper No. 3269 i titled " Wind Forces on Structures." We conclude that l procedures to be used to design the facility against the i effects of a tornado are acceptable. 1 10.2.4 Sacrificial Shield , The applicant has presented information in response to the l l staff's question 12.3.8 ( Amendment 15) indicating that the jet forces resulting from a longitudinal split (138 in ) will not result in failure of the sacrificial shield. The applicant has indicated orally that the sacrificial shield could withstand a 2.2 f t break within the pressure vessel annulus. This will be documented in a forthcoming amendment. Design of the sacrificial shield plugs around pipe DR H MSPOLY Q&O
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penetrations will prevent the plugs f rem becoming potential missiles ~ for nrimary ruptures within t' e pressure vessel annulus . We conclude that this is acceptable. 10.2.5 Materials and Cons truction The containment liner will be construe:ed from ASW A-516, Grade 60 s teel. The reinforcing steel used in Class I s tructures will meet the requirements of ASW A-615, Grade 40 with the structural steel and its f abrication for Class I structures meeting the requirements of one of the following ASTM s pe ci fi cations : A- 36, A-441, A- 5"2, A- 53, A- 7, A- 354 , A- 325 cnd A-490. The prestressing steel will confom to AS TM - A-421, Ty p e B A. The s tructural concrete will be in conformance with ACI 318-63. The user tes ting of reinforcing s teel will be perfomed on a frequency of one full size
]
1 reinforcing bar for each 50 tons of ea::5 heat , if the heat i s- i smaller than 50 tons . Based on the information contained in the PSAR, we conclude that the materia;s and construction te chniq ues are acceptable for cons truction of Class I s t ru ct ures . Th e p ri ma ry cont ai nments d rvwell and p ressure suppression chambern are prest ressed concrete usinr the BBRV System of l
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pos t-tensioning utilizing parallel . lay, unbonded type tendons. The tendons are fabricated from ninety,1/4-inch-diameter, cold drawn, s tressed relieved , prestressing grade wires with each tendon individually encased in conduit. The criterf a and loads are acceptable and the prestressed containment is verv similar to the Trojan Nuclear plant and < others recently approved. 1he applicant 's program of quality control for the Cadweld splices meets the AEC Safety Guide Number in on Cadweld splices and is acceptable.
'10.3 Mechanical Analysis and Design 10.3.1 Seismic Input The seismic design response spectra submitted provides for an amplification f actor of 3.5 at 2% damping for the period range from 0.15 to 0.5 see and amplification f actors greater than 1 in the period range 0.15 to 0.033 seconds. The structure and equipment damping is in accordance with those recommended by N. Newmark. The modified time histories to be used for component equipment design are adjusted in amplitude and frequency to envelope the response spectra specified for the site. We and our neismic consultants l
s conclude that the seismic input criteria proposed by the . applicant provide an acceptable basis for seismic design. 1
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10.3.2- Seismic System Dynamic Analyses Modal response spectrum multi-degree-of-freedom and normal mode-time history methods are used for all Category I i structures, systems, and components. Coverning response parameters will be combined by the square rc.ot of the sum of the squares to obtain the modal maximum when the modal response spectrum method is used. The absolute sum of responses is used for in-phase, closely spaced frequencies. Floor spectra inputs to be used for design and test verification of structures, systems and components are generated from the normal mode-time history method. A vertical seismic-system dynamic analysis is being employed for all structures, systems, and components. We and our seismic consultants conclude that the seismic system dynamic methods and procedures proposed by the applicant provide an acceptable basis for the seismic design. 10.3.3 Class I (Seismic) Mechanical Fluid Systems All Class I systems, components, and eculpment outside of the reactor coolant pressure boundary will be designed to sustain normal loads, anticipated transients and the j i Operational Basis Earthquake within the appropriate code l allowabic stress limits and the Design Basis Earthquake
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within stress limits which are comparable to those associated 4FRG AEUSEON:V 1 t RWg a
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1 I with the emergency operating condition category. We l 1 consider that these stress criteria provide an adequate margin of safety for Class I systems and components that may be subjected to seismic loadings. i 1 10.3.4 Se_ismic Instrumentation
, In Amendment 12, dated June 11, 1971, the applicant has indicated that the AEC Safety Guide Number 12 on Instruments-tion for Earthquakes will be met. We evaluated this material and concluded that the instrumentation is in accordance with i
the Safety Guide and is acceptable. 1 10.3.5 Missile Protection
'Ihe design criteria presented in the PSAR that govern the design of the missile protection features of the Zimmer facility are adequate in regard to both internal and external missilea. Intertial missile protection is provided by separation of vital redundant equipment, use of missile shielding, and proper orientation, and by the consideration of design of potential missile sources. The external missile criteria specifically provide for protection of the enutain-ment, control room, and other vital plant features. In the unlikely event of a turbine failure or generation of tornado missiles, the applicant's design criterion in that the plant can be shut down safely and maintained in the safe shutdown
<0 C AL-USEONJ Mr'
'9 23 C A.-USHN:Y-
. - 136 - j .. l condition. This criterion is acceptable. The secondary confinement building proposed for Ziruner provides for external missile protection for the drywell and fuel pool that is comparable with other BWR designs. We conclude that the missile protection criterion for the Zimmer Station is acceptable. 10.4 seismic Quality Assurance The quality assurance requirements for Class I (seismic) structures, systems, and components are specifically stated in Amendment 4. We have concluded that these quality assurance provisions that were implemented for all items designated as seismic Class 1 for design, comply with the requirements of Appendix B "Ouality Assurance Criteria ~for Nuclear Power Plants" of 10 CFR 50 j i
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,) 11.0 C u,iuct of Operations t i
'11.1 Organization and Responsibility a
The Cincinnati Cas and Electric Co. (CC&E), Columbus and Southern j Ohdo Electric Co. (C&SOE), and the Dayton Power and Light Company (DPL) l l are the applicants for the construction permit and facility license for the Wm. H. Zimmer Nuclear Power Station. The CG&E is responsible for the design, construction, and operation of the facility and is author-ired to act as the agent for C&SOE and DPL in all details of construc-tion including licensing. The applicants have no previous experience associated with generating electricity through the use of nuclear fuel. However, the utilities (specifically the CG&E Company) have acquired considerable" experience in the design, construction, and I operation of numerous fossil fuel power plants. Various subcon-tractors and consultants assisting the applicant have been actively involved in the nuclear energy field. The technical support for the design and construction effort of the facility is provided by the CG6E Company's Engineering and Construc-tion Department staf fed by over one hundred engineers. Responsibility for the review and approval of design features of the plant rests with the CC&E technical staff. In addition, the CC&E staf f will implement , with the ansistance of consultants, a quality assurance program and will follow field construction of the plant until its completion.
-onchus Hng
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Complementing the applicant's technical staff will be the engineer-ing staffs of: the architect-engineer Sargent and Lundy, which will i provide the coordination of all aspects of procurement of equipment, systems, and construction services through the use of specifications and drawings that will integrate the total station design; the General Electric Company which will provide the nuclear steam supply system, the nuclear fuel, and technical personnel for guidance and support during start-up operations; the Westinghouse Electric Corpor-ntion which will supply the power conversion equipment; and the con-structor, Kaiser Engineers, Inc. , which will perform the services of construction manager. Other subcontractors and consultants have been engaged to provide expertise in fields of a specialized nature such as: meteorology, hydrology, scismology and geology. We conclude, based on the results of the Zimmer review, that CG6E retains a technically competent and safety-oriented engineering organization that can ef fectively manage, design and construct the Wm. H. Zinner Nuclear Power Station. 11.2 Station Management Responsibility for operating the Zimmer Unit will be with the Superin-tendent of Electric Production Department who is responsible to the Manager of Electric Production for all electric power production activ-Itics in the Cincinnati Cas and Electric Company's system. The Plant Superintendent will report to the Superintendent of Electric Production and has the principal responsibility for all phases of operation and
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i i maintenance of the facility. Functionally organized under the plant $ Superintendent are the Operating, Maintenance and Technical Sections. !! The operations group is headed by the operating supervisor who is responsible for the supervision of the plant operations and will work closely with the other disciplines in the facility, n e Shift j Supervisor reports to the Operating Supervisor and will be in charge of the plant operation on a shift basis. The operating shift will { consist of Shift Supervisor, Control (Licensed) Operator, Assistant Control (Licensed) Operator, Auxiliary Operator, anu Assistant Auxiliary Operators. The auxiliary operator and assistant will be l qualified to perform the health physics technical activities. Five ) shifts are provided in order to maintain adequate twenty-four hour ( 4 per day, seven days per week coverage. We conclude that the appli-cant's operating shif t crew size, for single unit operation, is acceptable, I i i The Maintenance Department will perform all electrical and mechanical maintenance in the station in accordance with approved written i i operating and maintenance procedures. The Technical Engineer supervises the activities in the technical section and reports directly to the Plant Supervisor. D e Technical Engineer is responsible for the three specialized disciplines: nuclear engineering, instrumentation and control, and chemistry / health
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physics. The personnel in each of the above disciplines provide 3 specialized technical skills and knowledge to assure safe reactor 1 0
-operations. Technical assistance may be requested from the General l Engineering Department of CG&E or from Sargent and Lundy retained as consultants, j
- With respect to personnel qualifications, the applicant has estab-lished the minimum requirements for the selection, qualification, and training of key station personnel. These requirements are in l
, accordance with ANSI N-18.1 " Proposed Standard for Selection and Training of Personnel for Nuclear Power Plants" dated June 22, 1970, and are acceptable.
Based on the above information and the applicant's commitment to meet the requirements of AEC Safety Guide number eight " Personnel l Selection and Training" we conclude that the CG&E Company will have an adequate station organization and capable staff with which to operate the Zimmer facility safely. 11.3 Emergency Plans The applicant has presented in Amendment 11 a description of the emergency plans that will be developed for the facility for use in ' the event of an accident involving the release of radioactive mate-rial to the environment. The plan will rely on the use of the j station organization with the shift supervisor responsible for l
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initiating corrective action to limit the consequences of the l accident. Supplementing the normal operating erev for conditions that exceed the capability of the shif t crew will be a designated l emergency organization. In addition, an emergency coordinator will be designated who will be in complete charge during the emergency. At predetermined action levels the emergency coordinator will decide on the need and extent of offsite assistance required in case of an emergency involving the general public. The plan will be coordinated I with local agencies such as South Claremont Disas ter Committee (formerly the New Richmond Flood Committee), the Red Cross, law enf6'fcement agencies (i.e. , the Ohio State Patrol, the Kentucky State Police, and the Claremont County Sheriff), Fire Departments. U.S. Coast Cuard, Ohio and Kentucky Health Departments and other government agencies. Medical support will be coordinated through the applicant 's Medical Director. The applicant's description of the proposed emergency plan for the facility conforms satisfactorily with the guidance and criteria provided in Appendix E,10 CFR Part 50. A primary concern is the proximity of the facility to Moscow, Ohio and the Moscow Elementary School. The elementary school has an enrollment of 170 students and is 2650 feet from the Reactor building. Moscow, Ohio, has a population of approximately 700 and is 2700 feet south of the facility. The applicant has contacted j
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the superintendent of the Richmond School District with regard to evccuation procedures for the school. Procedures for evacuating and quartering residents in the areas of Moscow and Richmond, Ohio for Ohio River Floods have been in existence for some time, The South Claremont County Disaster Committee (formerly the New Richmond Flood Committee) and the Red Cross have decided to increase the scope to include all types of disasters. The applicant contacted the Red Cross and the South Claremont County Disaster Committee to discuss inclusion of the facility in the overall emergency plan for the local area. The applicant has documented the results of agreements reached with the various agencies contacted. We conclude that the emergency plan proposed for the site and surrounding area is accept-able for the construction permit stage of reviev, however, we intend to continue our review of this matter during plant construction to assure that the procedures for implementation of the program are adequate. 11.4 Plant Security A fence will be erected around the plant operating areas to preclude unauthorized entry of personnel into critical plant areas. In response to our query, the applicant stated that they recognized the need for plant security and that several schemes are being studied and evaluated. The protective requirements include, in addition to general access control, the use of monitoring techniques both inside and outside the building. This aspect will be reevaluated during the 4 3EllSE0N:Y p
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operating license review stage and we will incorporate any additional requirements that may evolve as a result of criteria to be developed on this subject. ' We conclude that the above measures are adequate at the construction permit stage to furnish reasonable assurance of plant security. ; L 11.5 Training ne CG6E Company nuclear training program is basically the same as the one developed by the General Electric Company for BWR plants. De program is designed to meet the specific needs of each man selected to perform a specialized task and to provide hkn with sufficient knowledge and experience to perfom their function. Senior Reactor Operator candidates will be trained to enable them - to receive their license before initial fuel loading. Other key i l personnel will also be assigned prior to initial fuel lo-ading so that they may participate in the design and construction phase and the formal training program. Le applicant has indicated in Amend-ment No. 13 that the proposed training program will meet the require-ments of ANSI N-18.1 as discussed in section 11.2 of this report. Based on the applicant 's commitment, we conclude that the training i program vill be acceptable to provide the facility personnel with the required capabilities to operate the Zimmer f acility safely. 2
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11.6 ' ' Operational Review and Audits The plant Superintendent has the responsibility and authority to operate the plant within the facility license and Technical Specifications and to approve procedures and their revisions within these limitations. The applicant has described his intent to establish a routine review function and a Operations Review Committee (ORC). Routine reviews will be conducted at the plant level under the direction of the Plant Superintendent. The ORC will be composed of at least five personnel who for the most part are not members of the operating organization. The ORC will meet at least quarterly and will be responsible for review of design changes involving unreviewed safety questions, and changes to the Techhnical Specifi-cations or other changes which might affect nuclear safety. The Consnittee will also review facility operation, equipment, and personnel performance to determine adherence to license requirements. The results of such reviews vill be reported to the Executive Vice President of CC&E. Based on the applicant's commitment to meet DRL's Operational Safety Guide No. 2, " Supplementary Guide to the Contents of Technical Specifications (Administrative Section)", we conclude that the plans for an objective review and audit are adequate. 1 I i i ___ - - _ - - m- ;
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12.0 PLANT SAFETY ANALYSIS 12.1 General The four standard BWR design basis accidents have been evaluated in the course of our review of the William H. Zimmer Nuclear Power Station. 'Ihe results of 'these analyses are shown in Table 121-1 The accidents analyzed are the: loss-of-coolant, refueling, control - rod-drop, steamline-break accidents. A break of an instrument or process line was also considered. None of the accidents analyzed resulted in calculated doses in excess of 10 CFR Part 100 guidelines. It .should be noted that we and the applicant have used a more conser-vative two-hour meterological assumption than the Pasquill Type F and a one meter per second wind speed assumption normally used in calcu-lating doses at the time of the construction permit. Table 12.1-2 presents the mixing re. duction factors for 50% and 80% mixing as a function of time after postulated accident. 12.2 Loss-of-Coolant Accident The design basis loss-of-coolant accident is the same as that used in j the analysis of other BWR's in that a double-ended break in the largest i I pipe, the recirculation line, in the reactor coolant system is assumed, f Our analysis is consistent with the AEC Safety Guide Number 3. A i l primary containment leak rate of 1.5%/ day was assumed. l This leak rate is consistent with the values being used for recently licensed operating facilities. A leak rate of 0.5%/ day is specified 1 4 CEllSFDEE )
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as the design leak rate for the containment; however, our analysis was based on a leak rate of 1.5%/ day to provide adequate margin. for operating flexibility. The primary containment leak rate will be reviewed again during the operating license review at which time the on-site meteorological data and containment mixing values will be available. The basis for the secondary containment is internal mixing in 50% of the free air volume of the reactor building and a 99% filter effi-ciency for the two SGTS charcoal filters in series which exhaust to the environment at ground level. The mixing reduction factors for 50% and 80% are presented in Table 12.1-2. The recirculation and mixing credit assumes that activity leaking from the primary contain-ment is mixed in 50% of the volume of the reactor building by means of the 80,000 cfm recirculation fan and associated ductwork with the release through two charcoal filters to the environment at ground level. This is a departure from the previous standard BWR containment concept where the activity was assumed to pass directly from the primary containment through a single charcoal filter to a high stack for an elevated release. Mixing by the 80,000 cfm recirculation fans in conjunction with the 2300 SCTS fan exhaust provides a two-hour site boundary thyroid dose reduction factor of 9.1 in addition to the reduction factor of 100 obtained by the two charcoal filters (in series) in the SGTS exhaust. This recirculation and mixing credit is similar to l )
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that assumed-in' the analysis for the Shoreham Station (Docket No.
.50-322). The' applicant has provided an analysis that indicates that greater credit for mixing in the reactor building could be , assumed-in the Accident Analyses. However, we consider it prudent to use the assumption noted above ~(50% mixing) until the f acility is constructed and suitable demonstration tests are performed.
The effect of better mixing can be inferred from the data presented in Table'12.1-2. The' resultant calculated doses for both the exclu-sion distance and the low Population Zone distance are within the 10~CFR Part 100 guidelines. 12.3 Fuel Handling Accident In this accident, it is assumed that a fuel assembly is dropped by the refueling crane into the reactor core or spent fuel pool from the maximum height permitted by the fuel handling equipment design. Bas ed on the anticipated energy dissipation distribution and the mechanical energy absorption capability of a fuel rod, the applicant has calcu-lated that 111 fuel rods would be damaged, releasing 10% of the contained noble gases and 5% of the iodine.to the surrounding water. A 1.5 radial peaking factor, 24 hours cooling time and a water decontamination factor of 100 for elemental iodine is assumed. As in the LOCA analysis, credit for mixing and filtration by the SGTS is assumed before releasing the effluent to the environment through the building vent which we assume is at ground level since it does not meet the 2-1/2 factor height criterion. The applicant has 4FFIBIAdEON3 - _ - -. JW
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provided an analysis to show that, following a refueling accident, there is adequate time available to isolate the reactor building and actuate the standby gas treatment system before the first " puff" of contaminated airborne particulate and gases reaches the vent 11a-tion ducts. The analysis is acceptable to justify the assumption used above. The resultant calculated doses are within the 10 CFR j Part 100 guidelines. 12.4 control Rod Drop Accident For this accident the doses resulting from the failure of 330 fuel rods were calculated using our standard model. Isolation of the main steam lines is assumed to occur in 5.5 seconds following the rod drop accident. The calculated doses in Table 12.1-1 are based on the standard set of BWR accident assumptions. The applicant has j stated that automatic isolation of the mechanical vacuum pump follow-ing a rod drop incident will be provided. The calculated doses presented for this accident are well within the limits specified in 10 CFR Part 100. 12.5 Main Steam Line Break j The amount of radioactivity released is calculated from the activity of noble gases and individual iodine radioisotopes that pass through the open break assuming a 5.5 see closure time for the MSL isolation l valves. m in.d!SH tY (pp- .
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' The applicant assumed the noble gas is dischar%ed at a 0.1 curie /
second rate after a 30. minute holdup while we assume a 1.0 curie / second rate. We believe that the use of the higher discharge rate to compute primary coolant activity is justified since the addition of the new offgas treatment system for Zimmer will reduce the dis-charge rate and consequently permit higher activity concentrations in the primary coolant'and main steamlines. Furthermore the higher discharge rate will provide margin for developing technical speci-fication' 11mits on the " treated" off-gas release rate, and the allowable activity level in the primary coolant which will be established during the operating license review. 12.6 Instrument Line or Process Line Break The consequences of a break or valve body failure on an instrument line or process line outside the primary containment results in an "unisolatable" line. This prompted the applicant to adopt the suggested methods of Safety Guide No. 11 to reduce the consequences of the failure of the instrumentation lines penetrating primary containment. At the time of the operating license review, the limit on primary coolant concentration of radioiodine will be established to assure that releases to the atmosphere from a break in any live steam line will not result in calculated thyroid doses at the site boundary in excess of 30 rem. Such a limit in Technical Specifications eV CMUSE~0N:.~Y-U%W <
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combined with the applicant's agreement to conform to Safety .Cuide No. 11 result in' an acceptable solution to the instrument line or process line break problem. k h k. Y~
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.i.9 hhhD TABLE 12.1-2 MIXING REDUCTION FACTORS T_IME 50% MIXING
- 80% MIXING
- 0-2 Hours 9.12 12.70 0-8 Hours- 2.81 4.44 8-24 Hours 1.55 2.37
> 24 Hours 1 1
- Mixing reduction factors based on 80,000 CFM recirculation flow rate and 2,300 CFM exhaust flow rate through the SGTS.
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1 13.0 . QUALITY ASSURANCE The' Cincinnati Gas and Electric Company' described their Quality
' Assurance '(QA) Program 'to be implemented in the design and construc-
-tion of the Wm. H. Zimmer Nuclear Power' Station in' Appendix'D of the I
'PSAR., The applicant has stated that the proposed QA program will
- satisfy the criteria of: Appendix B,10 CFR Part 50 and information identifying how the applicant intends to meet each of the 18 QA criteria has been addressed in the PSAR.
The Cincinnati Gas and Electric Company (CG6E) is' responsible for implementing the over-all QA program. The other principles involved are.Sargent-and Lundy, the' architect engineers; the General Electric Company, supplier of the Nuclear Steam Supply System; and Kaisers Engineers Incorporated, the constructors. The organization for the Zimmer project consists of a. project manager, an assistant manager who supervises eight full-time engineers (two from each of four disciplines that conduct only design reviews) and a separate full-time office QA group consisting of a section head and four engineers. This
. of fice QA group reviews drawings and specifications to assure that QA requirements, including design review by appropriate in-house engineers and. outside contractors, have been conducted. The applicant also has a field QA organization, consisting of three engineers who perform
- field QA work, such as equipment receipt inspection and equipment installation checking. The applicant stated that additional part-time f - -
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q QFROAL USRONW - 154 -
' office. QA personne1' are available if required and that 'a number of additional! field QA personnel would be assigned to the site.
JThio proposed organizational arrangement 'of QA functions relative
~
' to the project ! function and the planned extent of staffing is
. acceptable. - The applicant's QA plan, the criteria of Appendix B,.10 CFR 50, .and th'e 'draf t copy of their QA manual '(a working document to define in ~ greater detail the QA plan and responsibilities
-of the principles involved in their QA program) were inspected by Region III, Division of Compliance- in March 19 71. Compliance found that the. applicant's QA program in terms of criteria commitment in PSAR and dra.f t QA manual were unsatisfactory with respect to Part 50, Appendix B. During- the audit and at the management meeting at the conclusion of the Compliance-inspections, the applicant made commit-ments to revise the unsatisfactory sections of the PSAR. In amendment numbers 7. 'and 13, the applicant revised these unsatisfactory sections.
Additional. information describing their QA programs planned imple-mentation of each of the 18 QA criteria was provided. Based on the revised information, we believe that the applicant's quality assurance program is acceptable to satisfy the requirements of Appendix B of 10 CFR Iart 50. _ A re-inspection was performed by Region III, Division of Compliance on July.7 and 8, 1971. The appropriate , corrective action taken by the applicant was found to be acceptable. l l l L f/'ht?"90 ty"3" oP 'T & "O 1%? W V" J D$k 1 u( Ju 3 JOh:reji/i AlkJ W*CM DJi Y u.,1 h
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4 The architect-engineer (Sargent & Lundy) is responsible for the development of detailed specifications, drawings and coordination of the design interfaces. Sargent and Lundy's internal review program and the extent of its design review of GE-supplied equipment is acceptable. The General Electric Company's Quality Assurance-I program is identical to that performed on other BWR facilities; it is~ acceptable. The constructor (Kaiser Engineers, Inc. (KEI)) is responsible for conducting a Quality Assurance and Quality Control (QA/QC) Program that meets the requirements of the CG&E QA program and the ' intent of Part 50, Appendix B, as applied to the functions of the constructor. The KEI QA/QC manager has the authority to suspend construction activities at any time work continuation could affect-the safety related functions of the facility. Region III, Division of Compliance conducted a special inspection in January 1971 to review the KEI construction QA/QC program as related to the Zimmer project. Compliance found the Kaiser QA/QC programs acceptable and, more important, found that they demonstrate a sound knowledge of Part 50, Appendix B, requirements and fully understand the needs for implementing a comprehensive QA/QC program during construction. Based on the above information, we have concluded that the applicant's over-all Quality Assurance program for the Zimmer project is acceptable. qp73"im1* 6 rr "
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t I L l OFFFCEAL USE ONLY
- 156 -
14.0 TEQlNICAL SPECIFICATIONS Our review of the applicant's proposed Technical Specifications was limited to the coverage'and depth required at this stage of'the licensing process (10 CFR 50.34a(5)). We have informed the applicant of our position' that a section on the Sites Environmental Monitoring Program describing the concept and procedures for the program should be included in the Technical Specifications. The applicant was requested to provide more information regarding probable content. We suggested that at the construction permit stage the applicant could develop this information from review of a recently issued set of Technical Specifications and identify the areas where there are variations resulting from differences between the companys' operating practices or plans and from differences in design of the facility relative to the referenced facility. The applicant stated that the Technical Specifications conform to 10 CFR Part 50.36(c) by including the items in che categories specified. They also stated that Technical Specifications for other BWR facilities (including Dresden 3) had been reviewed before submitting the proposed content of the Technical Spec-ifications for the Zimmer Station. It is the Applicant's position that 1 the outline proposed for the Wm. H. Zimmer Nuclear Power Station is more j i appropriate at this stage of design and that fundamental differences j between other BWR facilities and the proposed Zimmer Station are not 1 sufficiently comparable to provide a basis for an incisive and detailed j comparison of the Technical Specifications. I naawanaw .___w- o. i
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4FJECHAL USE OW
'- 157 -
The applicant will use as a guide the most recently issued BWR Technical Specifications in preparing their detailed Technical Specifications. These specifications will be reviewed in depth at the operating license stage and will contain requirements consistent with Technical Specification that are acceptable to us, however, we conclude that the requirements of 10 CFR 50.34a(5) are met. 4 e l l l 1 ODSU@U A 17 I?T@7
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- 15.0 CONFORMANCE TO GENERAL DESIGN CRITERIA
. Based on our evaluation of the preliminary design and design criteria for the proposed Wm. H. Zimmer Nuclear Power Station, Unit I, we have concluded that the applicant plans to meet the intent of the General Design Criteria for Nuclear Power Plants, published May 21, 1971, as Appendix A to 10 CFR Part 50 in the final design of the station.
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