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Docket Nos. 50-352 and 50-353 July 20, 1971 REPORT TO THE ACRS Limerick Generating Station Unita 1 and 2 t-m 4
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I TABLE OF C0! GENTS P, age
. ABSTRACT i
'1.0' INTRODUCTION AND
SUMMARY
1 1.1 General 1 1.2 Maj st Areas of Review 3 1.3 Summary 3 1.3.1 Areas Not Completely Resolved 4 1.3.2 Areas of Continuing Review 6 1.3.3 Comparison of Limerick and Newold Island' 9 2.0 SITE QlARACTERISTICS 13 2.1 Geography and Demography 13 2.2 He teo'rology ' 15 2.3 flydrology 17 i
2.4 Geology, Seismology, and Soil Mechanics 21 2.5 Ecology 24 2.6' Environmental Radiation Monitoring 24 2.7 Air Traffic 25 2.8 Railroad and River Traffic 26 3.0 REACIOR DESIGN 27 3.1 General 27 3.2 Nuclear Design 27 3.3 Thermal and liydraulic Design 28 3.4 Reactor Internals 32 3.4.1 Design 32 3.4.2 Dynamic System Seismic, Operating and LOCA Analysis 33 3.4.3 Vibration Control 34 4.0 REACTOR COOLANT SYSTDI 35 4.1 General 35 4.2 Reactor Coolant Pressure Boundary 36 1
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i TABLE OF CONTEICS (cont.)
P, age, 4.2.1 Design 36 4.2.2 Pipe Whip Criteria 36 4.2.3 Main Steam Line Isolation Valve Leakage 38 4.2.4 Seismic Design of Main Steam Line Piping 38 4.2.5 Primary System Pressure Relief System 39 4.3 Application of AEC Classification Groupa 40 4.4 Reactor Vessel Material Surveillance Program l and Fracture Tought.eas Criteria 42 1
4.4.1 Surveillance Program 42 l 4.4.2 Fracture Toughness Criteria 42 4.5 Inservice Inspection 44 )
4.6 Reactor Coolant System Sensitized Stainless Steel 44 4.7 Electroslag Welding 45 4.8 Foreign Procurement 46 4.9 Leak Detection 46 l 5.0 CONTAINMENT AND STRUCTURAL DESIGN 48 5.1 Containment Design and Comparison with Similar Plants 48 I 5.1.1 Primary Containment- 48 5.1.2 Secondary Confinement 55 i
5.1.2.1 Reactor Building Recirculation I System 56 1' 5.1.2.2 Standby Cas Treatment System 58 5.2 Structural Design 59 5.2.1 Class I (seismic) Structures 59 5.2.2 Environmental Effects 60 5.2.3 containment Structural Design Analysis 60 5.2.4 Testing and surveillance 61 5.3 Seismic Design 63 5.3.1 Seismic Input 64 5.3.2 Seismic System Dynamic Analyses 64 )
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TABLE OF CONTENTS (cont.)
- f. age, 5.4 Comparison Between Limerick and Newbold Island containment Concepts 65 6.0 ENGINEERED SAFETY FEATURES 68 6.1 Emergency Core Cooling Systems (ECCS) 68 6.2' Hydrogen Generation in Primary Containment Following a LOCA 73 6.2.1 Hydrogen Control System 74 6.2.2 Containment Inerting 74 6.3 Corrosion Cracking of Pipe Metal 75 7.0 INSTRUMENTATION, CONTROL AND ELECTRIC POWER SYSTEMS 76 7.1 Instrumentation and Control Systems 76 !
7.1.1 BWR Ceneric Problems 76 7.1.2 Post Accident Monitoring Instrumentation 77 7.1.3 Environmental Testing 77 7.1.4 Engineered Safety Feature Testing 78 7.2 Electrical Power Systems 78 7.2.1 Offsite Power 78 7.2.2 Onsite Power 79
- 7. 3 Design criteria for cable Installation and Identification 81 8.0 AUXILIARY SYSTEMS 83 8.1 General 83 8.2 Radioactive Waste Systems 84 8.2.1 Liquid Radwaste System 84 8.2.2 Gaseous Radwaste System 87 8.2.3 Solid Radweste System 90 8.3 Spent Fuel Storage 91 8.4 Emergency Service Water System 91 8.5 Main Control Room Ventilation System 92 8.6 Residual Heat Removal Service Water System 93 l
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TABLE OF CONTENTS (cont.) l P,sg j
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. 9.0 ACCIDENT ANALYSIS 94 9.1 Ceneral 94 9.2 Loss-of-Coolant Accident 95 9.3 Puel Handling Accident (Refueling Accident) 97 9.4 Control Rod Drop Accident 97 9.5 Main Steam Line Break 98 9.6 Control Room Exposure Doses During Accidents 99 9.7 Main Steam Line Isolation Valve Leakage 100 9.8 Instrument Line or Process Line Break 101 10.0 CONDUCT OP OPERATIONS 105 j 10.1 Technical Qualifications 105 10.2 Organization of Plant Management 106 10.3 Operating Procedures 107 10.4 Startup, Reoperation, and Power Tests 108 10.4.1 Construction and startup tests 108 10.4.2 Preoperational tests 108 10.4.3 Startup and power test progras 109 10.5 Emergency Planning and Plant Security 109 10.5.1 Emergency plans 109 10.5.2 Plant security 110 10.6 Independent Safety Review of Operator Actions 111 11.0 QUALITY ASSURANCE 112 12.0 TECHNICAL SPECIFICATIONS 115 13.0 ACRS CONCERNS 116 14.0 CONFORMANCE TO CENERAL DESICN CRITERIA 124 List of Tables 1.0 Chronology 10 3.1 Comparison of NWR Design Parameters 34a 4.3 AEC Code Classifications 40a 5.1 Comparison of Containment Design Parameters 48a 5.3 Comparison of Limerick and Newbold Island 66 6.1 Comparison of ECCS Capabilities 69 9.1 Assumptions for Accident Dose Calculations 102 9.2 Calculated Exposure Doses 104
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.BifNN ABSTRACT i
on February 26, 1970, the Philadelphia Electric Cornpany submitted an application for a construction permit for.a dual-unit nuclear power plant facility to be situated on the Schuylkill River. The site is 1.7 miles from Pottstown, Pennsylvania, which has a population of j approximately 27,000 'and is about midway between Reading and Philadelphia, Pennsylvania. -The distance from Limerick Generating Station to the city limits of Philadelphia is about 21 miles.
The Linarick Generating Station vill use two identical boiling water reactors each with a rated thermal output of 3293 W and a gross electrical output of 1152 W at the generators. The General Elec-tric Company will supply the nuclear steam supply systems, nuclear fuel, and the turbine generators. Bechtel will furnish engineering and construction services for the design and construction of the 1
plant. . Fabrication and subsequent onsite assembly of the reactor q 1
vessels will be accomplished by the Chicago Bridge and Iron Company.
Two natural draft, hyperbolic cooling towers will be employed to dissipate waste heat to the atmosphere. Cooling water for makeup '
to the main circulating water system will come from either the Schuylkill River or the Delaware River, depending on' the water usage permit obtained by the applicant. Delaware River water, if l i
used, will be routed via pipelines and open channel to the site.
Discharge of liquid and gaseous radioactivity will be limited so that annual average concentrations will be less than 1% of the 10 CFR Part 20 values. .
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1.0 INTRODUCTION
AND
SUMMARY
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1.1 General 8f s
on February 26, 1970, the Philadelphia Electric Company filed an application for a construction permit for the Limerick Generating Station Units 1 and 2 to be located on the Schuylkill River about 1.7 miles southeast of Pottstown, Pennsylvania'and 20.7 miles northwest of Philadelphia. Pottstown is the nearest " population center" (population about 27,000). The earliest dates for operation are indicated to be March 1975 for Unit 1 and March 1976 for Unit 2.
The designed thermal power for each of the identical boiling water reactors is 3440 MW. Initial operation is proposed at a rated j thermal power of 3293 MW with a gross electrical output of 1152 MW at the generators. ,
The design of both reactor units is similar to boiling water reactors previously approved for construction. A comparison of some of the design characteristics of Limerick Generating Station with those of, similar facilities is given in Section 3.0.
1 The containment design is similar to that for Shoreham Nuclear Power Station wherein the "over-under" vapor suppression containment concept l
is used. The drywell is a reinforced concrete, steel-lined frustum )
of a cone; and the wetwell or vapor suppression chamber is a right I circular cylinder of steel-lined reinforced concrete which is located directly beneath the drywell. A secondary confinement, the reactor j building, is a reinforced-concrete building completely enclosing the l
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i two reactor containment structures. The reactor building provides an .
1 additional barrier to the release of airborne radioactive materials.
The waste gases from the main steam condenser and other sources will be processed in a cryogenic system for removal and hold-up of the liquified radioactive noble gases. Normally the liquid or liquid entrained radioactive materials will be collected and processed in the liquid waste disposal system. The gaseous and liquid radwaste systems design objective is to reduce activity such that the annual average concentrations for routine discharges are less than 1% of 10 CFR 20 limits.
The ACRS Subcommittee for the Limerick Generating Station made a site inspection on November 10, 1970. At that time, foundation excavation and other site work were underway. A request for a construction exemption in accordance with paragraph 50.12 (proposed) of 10 CFR ,
Part 50 was submitted in November 1970 by Philadelphia Electric Company. The request was not granted at that time because of the existence of unresolved design problems and a deficient quality i
assurance program. The latter program has been improved and found )
satis f actory. Action to process the request for a construction q l
exemption will commence in the near future. I i
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An ACRS subconusittee meeting on Limerick was held March 31, 1971 and a full /.CRS meeting was held on May 6,1971. The letter meeting involved comparison of site related features for Limerick and Newbold Island.
The following discussion in this section highlights the general scope of our review and the results achieved to date. A chronology {
l of significant events is provided for information in Table 1.0. J 1.2 ' Major Areas of Review _
Since the nuclear steam supply system and principal protective features for Limerick, Peach Bottom Units 2 and 3, and Newbold Island are -
generally similar, a comparison of the features of these plants served ,
I as a starting point for the review. Design features and problem areas j 1
unique to Limerick were identified and evaluated. Emphasis was placed i upont (1) site-related items, (2) calculation of internal containment pressure following a postulated loss-of-coolant accident, (3) provisions l
for supply of adequate cooling water, (4) fission product control and release following a IDCA, (5) radweste control and discharge, and (6) areas of concern in previous reviews which still required additional information.
1.3 Summary Our review and evaluation of the Limerick application is complete except for a few matters that have been discussed with the applicant. Some areas have been identified for which supplemental information must be
W provided by the applicant prior to the issuance of the construction permit while other areas can be resolved during construction of the facilities. These areas are enumerated below.
1.3.1 Areas Not Completely Resolved In the areas listed below, the applicant has not submitted all of the.
information or made all of the changes in design which we will require for resolution. A brief statement of the problem accompanies each entry while greater detail can be found at the referenced section.
The " Status Key" is defined as follows:
Letter Status U Applicant's position is unknown at this time.
S Applicant's position is known. See referenced section.
I The area has not been completely reviewed.
C A generic problem.
Section Status Key Subject 2.6 I Environmental Monitoring: Applicant will be required to submit descriptive material on all aspects of reoperation monitoring and plans for operations 1 monitoring in accordance with commit-ment in Supplement 5 to PSAR.
3.4.1 S Acceptable Limits: Applicant will be 4.2.1 required to document use of limits for NSSS Components loading criteria, f
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Section Status Key Su,b ject 4.2.4 U Class I (neismic) Design of HSL Piping:
Applicant will be required to document his response to nav requirement for Class I MSL piping design.
6.1 U LPCI Systems App;icant will be required to document his response to the change in LPCI system.
6.1 U HPCI Systems Applicant will be required to document his response to the change in HPCI system.
7.2.2 I On-site Power: Arplicant will be required to submi: additional justi-fication for noncenformance with Safety Guide No. 6.
7.3 I Instrument Cable : installation: Applicant will be required to provide additional explanation of criteria for minimum physical separatica and barriers.
8.2.2 I Caseous Radioactivity: Applicant will j be required to provide additional informa- ,
tion on expected raseous release rate j (Mech. vac. pump, HPCI turbine, ventila-tion exhaust, containment purge, other.)
8.4 U Flooding of ESW5 oumps: Applicant will be required to describe design features that prevent flooding of both ESWS pumps in event of a passive failure of the pump discharge line.
9.6 S Control Room Exposures: Applicant will be required to resolve differences in calculating exposures and/or provide adequate protection to personnel. ;
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1.3.2 Areas of Continuina Review _
The following areas have been reviewed but will require documentation following further discussion with the applicant during the construction period. No significant problems are anticipated. The status key is used as before.
Section Status Key Subject
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2.3 I Dilution and Dispersion: Applicant will be required to provide dilution and dispersion factors at eleven intake points in Schuylkill River downstream from site.
2.3 U Flood Protection: Applicant will be required to document details of flood protection for Schuylkill River intake structures, pumphouse, and pipes.
2.3.8 U Supply of Cooling Water: Applicant will be required to provide final con-cept for the supply of emergency cooling water for station. 3 3.3, 3.4 C Nuclear Fuel Integrity: Applicant will be required to document performance of fuel with high burnup and power, and with postulated flow blockage.
3.4 G, S Inservice Vibration Monitoring: Appli- ;
cant will be required to describe con- 1 firmatory test program.
3.4.1 G Cuide Tube Collapse: Applicant will be required to adopt solution to prob- J les as developed in future (follow Fermi II).
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Section Status Key Subject 3.4.2 S Seismic Design: Ceneral Electric Co.
must provide procedures for determining dynamic loading.
4.2.5 U Safety Valves: Applicant will be required to document reasoning and analysis supporting reduction of safety valves from four to two in number.
4.3 U Cast Parts: Applicant will be required to document acceptance of criteris.
4.4.2 I Fracture Toughness Criteria: Applicant will be required to document at OL stage the supporting data and the heatup and cooldown limits for reactor vessel.
Predict NDT temperature of ferritic piping systems'in drywell' by using DRT data.
5.1.1 S, I Cate Valves on Drywell Downcomers:
Applicant will be required to provide -
the specified test and equipment per-formance capability to assure proper operation in wetvell post-LOCA environ-ment. Document in-service test and surveillance planned.
5.1.1 b, S Dryvell Deck Leak Test: Applicant will 5.2.4 be required to determine detailed require.ments for testing drywell deck.
5.1.1.2 I Heat Load on Filters: Applicant will be required to submit data on maximum heat load and temperatures for charcoal and HEPA filters of recite system and SBCTS for post-L0CA situation.
1 6.1 C ECCS Evaluation: Applicant will be I required to utilize AEC Reg. review and !
evaluation of GE-ECCS capabilities to i modify, if appropriate, design and/or i operation. ]
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l Section Status Key Subj ect 6.2 S. I Hydrogen control, Post-LOCA: Applicant will be required to continue to provide design details and description of opera-tional capabilities to assure effective-ness of system.
8.2.2 I Caseous Radvaste System: Applicant will be required to provide additional information as the design, development, and function test of this system pro-gresses.
8.3 I Spent Fuel Cask Movement: Applicant will be required to document the manner for administrative 1y controlling the movement of the spent fuel caske 11.0 $ Quality Assuraness Applicant will be required to concinue observation and review of Bechtel on-site assignments and performance.
12.0, 2.5 S Site Environmental Monitoring Program:
2.6 Applicant will be required to include this program in the Technical Specifi-cations.
13.0 d C Anticipated trar.sients with f ailure to scram (ATWS): Applicant will be required !
to adopt staff recommendations following the review of CE ' Topical Report.
13.0 e C Common mode failure: Topical Report NEDO-10819 will be reviewed; apply results to design of plant.
1.3.3 Comparison of Limerick _and Newbold Island Throughout the review of Limerick, we have made a deliberate effort to adjust the design, where appropriate, to provide safety features equal in effectiveness and quantity to those provided for Newbold Island. The
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spplicant has been advised of the July 1971 ACRS recommendations for modification of the HPCI and LPCI systems (see section 6.1) and for designing the MSL piping to Class I (seismic) standards (see section 4.2.4). We expecc that the applicant will accept these recommendations to give Limerick, in general, the same functional capability as Newbold Island. Comparison of the Limerick 'and Newbold Island capabilities is provided in Tables 3.1, 5.1, 5.3, and 6.1 of this report.
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Chronology
- 1. February 26, 1970 Submittal of Preliminary $sfety Analysis and License Application
- 2. May 13, 1970 Initial meeting with applicant to discuss review schedule and areas for review
- 3. June 10,1970 Meeting with applicant to discuss site related matters, radioactive waste control and disposal, and radiation shielding.
- 4. June 19, 1970 Request for Environmental Report from applicant
- 5. June 25,1970 Request for additional infonnation from <
applicant
- 6. June 30-July 1,1970 Heeting with applicant to discuss reactor fuel and vessel internals, primary coolant system, containment, CSCS, and safety analysis
- 7. August 6, 1970 Request for additional information from applicant
- 8. August 19, 1970 Meeting with applicant to discuss struc-tural analysis including seismic design and forces due to LOCA, winds, ton.adoes, missiles, and thermal stresses
- 9. September 15, 1970 Request for additional information from applicant
- 10. September 23, 1970 Meeting with applicant to discuss pre-liminary technical specifications conduct of operation, emergency planning, and ,
instrumentation and cont _01 l l
- 11. October 6, 1970 Submittal of Supplement No. I to PSAR )
- 12. October 10, 1970 Request for additional information from applicant
- 13. October 13, 1970 Request for additional information from applicant
- 14. October 30, 197G Submittal of Environmental Report by applicant
'. ) . November 10, 1970 ACRS Subcomunittee site visit
- 16. November 24, 1970 Submittal of Supplement No. 2 to PSAR
- 17. December 2, 1970 Request for additional information from applicant
- 18. December 7, 1970 Submittal of a request for construction exemption
- 19. December 11, 1970 Submittal of Supplement No. 3 to PSAR
- 20. January 14-15, 1971 Meeting with applicant to discuss respense to varied questions as presented in Supplements 1, 2, and 3
- 21. February 8, 1971 Request for additional information from applicant l
- 22. February 24, 1971 Submittal of Supplement No. 4 to PSAR
- 23. March 2, 1971 Meeting with applicant to discuss analytical model for calculation of post-LOCA containment pressure transient
- 24. March 11, 1971 Request for additional information from applicant
- 25. March 12, 1971 Request for additional information from applicant .
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- 26. March 30, 1971 Meeting with applicant to discuss agenda items prior to ACRS Subcomunittee Meeting
- 27. March 31, 1971 Meeting of applicant and DRL staf f with the ACRS subcosutittee for Limerick Generating Station.
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- 28. April'7, 1971 Submittal of information for anti-trust review by Department of Justice
- 29. April 21, 1971 Submittal of "Special Report to ACRS" by DRL
- 30. April 30,1971 Submittal of Supplement No. 5 to PSAR.
- 32. May 26-27, 1971 Meeting with applicant to discuss questions on a variety of matters
- 33. June 11, 1971 Submittal of Supplement No. 6 to PSAR 34 . June 15, 1971 Submittal of Amendment No. 7 to PSAR
- 35. June 17, 1971 Request for additional information from applicent (update of financial status and data)
- 36. June 18, 1971 Meeting with applicant to discuss seismic design, selected structural design require-ments, and status of review i l
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I 2.0 SITE CHARACTERISTICS
- 2.1 Geography and Demography ne 587-acre site for the Limerick Generating Sta: ion (LGS) is located on the Schuylkill River 1.7 miles southeast of Pottstown, Pennsylvania and about midway between Philadelphia and Raading. Philadelphia is about 20.7 miles away.
The site is divided into three segments. The principal portion, which will have the major operating equipment and facilities, is on the east bank of the river. This is separated from the second segment, on which cooling water intakes are to be located, by the triple tracks of the Reading Railroad's main line. The third portion lies on the west bank of the Schuylkill River adjacent to the tracks of the Penn Central Railroad and provides additional exclusion area.
De topography of the area is hilly, with elevations within five miles of the site ranging from 110 feet above mean sea level (MSL) at the Schuylkill River to 560 feet HSL. The plant grade will be at 218 feet NS L. The track of the Reading Railroad is at 143 feet MSL, about five hundred feet from the plant. Two small parallel streams run southwest through the site to the river.
- Additional information is contained in Special Report to the ACRS,
! April 21,1971, " Comparison of Sites for Limerick Generating Station and Newbold Island Generating Station."
i The minimum exclusion distance available is about 2500 feet (760 meters) from the center of each reactor. There will be no private residences located -inside the exclusion area. Several minor roads traverse the site, but will not be available for public use within the exclusion !
area. A small, silt-formed island in the' Schuylkill River is within the exclusion area; the island is uninhabited and is owned by the State of Pennsylvania. On some of the land inside the exclusion area, farming may be permitted and public recreation areas may be provided under the control of Philadelphia Electric Company. An information center for the public will be built on the property.
1 The city of. Pottstown had a 1960 census population of 26,000 and a predicted population of 50,000 for the year 2000. The Population Center Distance (PCD) is 1.7 miles, the distance to the city limits of Pottstown. The Low Population Zone distance (1PZ) is 1.28 miles.
PECo. estimated the 1958 population was 500 and 4800 persons inside a one and two miles radius circle, respectively, from the site. Thus, we consider the LPZ distance of 1.7 miles acceptable.
With respect to residents, the cumulative population distribution in 1960 at the Limerick site was approximately the same as chose of Zion and Indian Point Nuclear Power Stations. Plots of cumulative population distribution for these facilities and several others including Newbold Island appear in Figures 1 and 2. There are two
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major industrial facilities with a total of 850 employees within the PCD (1.7 miles). This number of persons in considerably less than that within the same distance from the Midland Nuclear Pacility. With respect to population, the Limerick site is within the envelope of sites previously considered acceptable. The applicant indicated that nearby industrial activities, the two railroad lines (one through the exclusion area), and a stone quarry will operate without detrimental effect on the nuclear facility. Our study and that of the applicant have confirmed that no military installa-tions are located within five miles of the site. A private swimming club located about 1.3 miles from the site will be operated during the warner months of the y?ar. The maximum attendance at this pool runs around 800 people, but on the average will be less. The pool usage will be recognized in preparation of emergency plans, according to the applicant. 2.2 Meteorology The applicant has presented meteorological data from Weather Bureau records for Philadelphia and Reading airports (each about 25 miles ] from the site) and from PECo. meteorological studies at their Peach l Bottom site (about 50 miles away). No data on atmospheric diffusion I at the site is now available. The applict.nt has relied on extrapola- 1
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tion of Peach Bottom data, converting from Brookhaven stability l l 2
nr l' categories to a form usable in classifying the wind data into appropriate Pasquill stability categories. The converted data were submitted as an amendment to the Peach Botton PSAR and have been found to ' give, in the l most stable period, a Pasquill stability of Type F with a wind velocity of 0.7 meters per second (5 percentile meteorological diffusion equivalent). Type F and 0.5 meter per second wind conditions hare haen used for con-servative meteorological calculations when determining exposure doses from airborne radiation. The meteorological situation at Limerick should be similar to that at Peach . Bottom since the general climatology and topography are similar and they are only 40 miles acert. Our consultants at National Oceanic and Atmospheric Administration (NOAA), formerly ESSA, have reviewed the Peach Bottom data and found it to be accept-able; however, the final onsite determination of actual conditions must await the installation of towers and instruments. l There are several meteorological towers
- for the site: Tower No. I will be 280 feet in height erected at ground elevation of 250 feet HSL on the north edge of the site; Tower No. 2 will be located on the opposite side of the Schuylkill River at a ground elevation of
*The towers have not been erected as of June 7, 1971 due to delay in obtaining FAA approval. At the site for the highe- tower, a tele-phone pole was erected and mounted with minimal instrumentation.
-______.----__------._.s---
~125 feet MSL and height of 310 feet. Aerovanes and temperature detectors will be provided on the towers. A satellite tower will be located at ground elevation 140 feet MSL on the near shore of the Schuylkill River. An aerovane will record wind velocities and direc-tions about 40 feet above the river level. In view of our opinion that releases should be assumed to occur at ground level, the tower instruments should give data which can be used in evaluating ground level and river valley flow. Our evaluation of the planned installa-tion reveals that the applicant's data collection should produce satisfactory results.
2.3 Hydrology The Schuylkill River basin extends about 80 miles above its confluence J with the Delaware River. The total drainage area'is relatively small, j i being only 1900 square miles above the Limerick Station. The appli- l I cant has calculated a probable maximum flood (PMF) discharge which corresponds to a discharge stage elevation of 159 feet MSL at the site. The plant grade is at 218 feet MSL. All safety-related plant facilities, except Schuylkill River water supply structures, dre well above the Probable Maximum Flood (PMF) stage. The applicant has indicated that cooling water intake structures, lines, and the pump ; house which are located below FMF stage will be protected against l flooding. With formal documentation of this latter commitment, there p L I
1 l l l l l i 1 will be adequate substantiation of flood protection of safety-related structures and equipment. The total requirement for normal cooling water for the facility is ! i stated as 74 efs, including 54 efs consumptive use (cooling tower windage and evaporation). Downstream consumptive requirements for Schuylkill River water are 467 efs. The minimum instantaneous flow of s record for the Schuylkill River at Pottstown was 87 cfs in 1930. If dilution water requirements, about 33 cfs, are included, the applicant is correct in' his evaluation of the inadequacy of the Schuylkill River as a normal supply of cooling water. To provide another source of cooling water, Philadelphia Electric Company is completfog agreements with the Delaware River Basin Commission and other agencies to use Delaware River water. The . Delaware River water will be piped eight ! uds from a proposed Point Pleasant pumping station (to be operatad by Bucks County) to the headwaters of the East Branch of the Perkiomen Creek where the water will be routed by open channel (gravity) flow through the East Branch for about thirty miles to a PECo. pumping station. Consumptive cooling water requirements for the LLserick l Station may then be extracted and piped eight miles to the site. About 20 efs "nonconsumptive" water requirements will be obtained from the Schuylkill for normal cooling. Nonconsumptive water is returned to the Schuylkill River as " blowdown" or " dilution water" from the cooling tower basin.
e e l l Shutdown of one unit because of a design basis' accident in one unit and normal shutdown in the other unit requires special attention for supply of cooling water. Emergency shutdown will require a ministan
- cooling water flow of 38 cfs. The Delaware diversion is not con-sidered a source since pipelines and pumping plants are not to be Class I structures. A detailed independent frequency analysis of the j
historical record of the Schuylkill River low flow capability by both I the applicant and the staff indicates a 100 year instantaneous low flow of about 80 cfs. However, the historical record contains the 1 effects of upstream mine pumpage and consumptive water use. Since future uses are not (cutrollable by the applicant, he has informally suggested two possible alternatives to assure a cooling water supply. 4 One alternative would be to obtain assurance from the Delvare River l Basin Commission of a minimum supply. The other would be to provide an onsite Class I storage basin and piping system for use in spanning a twelve-hours duration of low flow in the Schuylkill River. Either proposal is acceptable; the method adopted will be required to be l documented by the applicant. i All liquid radwastes are to be contained in Class I structures. However, there are seventeen major water intakes located along the Schuylkill between the site and the Delaware River confluence in Philadelphia. The capacities of municipal users range up to that of
l . 20 - l the 198 million gallons per day (e.g.d.) for the City of Philadelphia, and indicate the extensive use and reuse of Schuylkill River flow. To aid in evaluating the potential effects of a radioactive spill, regard-less of probability, the applicant has been~ requested to evaluate dispersion and dilution factors at various flow rates at eleven downstream intakes. The applicant will be required to provide this information in the FSAR. ; All the ground water aquifers in the area occur in joints, fractures and other secondary openings in the consolidated rock beneath the thin cover of residual soils. The applicant has performed pumping , 1 and percolation tests at the site, and has concluded that the residual soils are very impervious and that only wells generally in the direc-tion of the strike of the strata may be affected by conditions at the site. The applickat has concluded, after study of regional geology and the location of public ground water wells, that all such wells cannot be affected by conditions at the site. The applicant has found approximately 77 wells within one mile of the site. The staff kgrees with the applicant's conclusions that none of these wells I can be affected by conditions at the site because of their relative j i locations and depths, the location of the river, and local ground water gradients. 1 We concur in the analysis of the flood protection of safety related l l facilities, subject to documentation of the river inlets and pump structure designs. We also concur in the ground water analysis.
2.4 Geology. Seismology, and Soil Mechanics The site is in the Triassic Lowland section of the Piedmont Physio-graphic Province, a section characterized by gently rolling surface on an eroded low plateau. A shallow residual soil overlies Triassic sedimentary rock of the Newark group which is about 8000 feet thick above the Paleesoic and Precambrian basement rocks. The dominant structural feature of the region is the Regional Appalachian Orogenic belt, marked by a northeast-southwest orientation (roughly perpendicular to the Schuylkill at the Limerick Station) of the axis and lineation of the rock structure. The nearest faults to I the site are 9 miles to the southwest, a Paleozoic thrust fault inactive for 200 million years; and 9 miles to the northeast, a Triassic strike slip fault, inactive for 140 million years. , Measurements at the site include geologic, seismic refraction, shear wave velocity, uphole velocity surveys, and micromotion measurements. Laboratory analyses of site borings include compression, shock scope, and torsional shear tests. All tests indicate that plant structures, such as the Rasetor Building or Radwaste Building, will be founded on competent bedrock. In the area of the site, the river flood plain is less than 100 feet across and consists of a thick layer of alluvial deposits over broken fragments of underlying bedrock. If Class I seismic structures cannot be founded on competent bedrock, the I l l l ____________-_-__a
- v. ,
applicant states that they will be founded in selected granular fill compacted to 95% of the maximum dry soil density according to AASHO Standard'T180-60, Method D. In-place fills are to be tested for soil strength in accordance with AASHO T147-54. The applicant states that-all Class I structures will be founded on bedrock. The applicant has been advised that the Class I discharge piping from pumphouse to plant should receive special attention. This portion of 1 the intake system will' be placed in engineered fill at ' a 2:1 slope. Descriptive data on fractured zones which are uncovered during excavation at the site were requested. One fracture zone which was discovered can be grouted and ' stabilized satisf actorily, according to the applicant. Other fractures, if encountered, can be stabilized against differential settlement or consolidation by concrete " dental work" epanning the fracture zone. A rock fracture known as the Sanatoga fault passes 1300 feet west of the site and' has not been active for 140 million years. Other similar rock fractures are found in the Newark basin further from the site but are also inactive. The area of the site has not experienced major earthquake activity as far back as the early 18th century. Most such activity in southeastern Pennsylvania occurs in the Piedmont west of the fall zone, the physiographic boundary between the Piedmont and the Coastal Plain occurring about thirty miles southeast. The closest e
Intensity VII earthquake was about thirty-five miles from the site near Wilmington, Delaware, one hundred years ago. Subterranean caverns, natural and manmade, can be found in the general region; for instance, there are coal minea in the area around Wilkes-Barre, Pennsylvania, which is sixty-five mifes to the north of the site. ' Collapse or cave-in of underground caverne have occurred; records exist for such occurrence at Wilkes-Barre. However, the shocks or earth tremors associated with these esve-ins are local and not transmitted over great distances. Such caverna are not in evidence around the Lim rick site. Our seismic consultants at the United States Geological Survey and at the Seismology Division of the National Ocean Survey of the National Oceanic and Atmospheric Administration (formerly U.S. Coast and Geodetic Survey of the Environmental Science Services Administration) have concurred in the applicant's selection of an deceleration of 0.06g resulting from an intensity VI (HM) earthquake for representing of an earthquake likely to occur within the lifetime of the facility (40 years). The consultants also concurred in selection by the applicant of a value of 0.123 for the acceleration associated with an intensity VII earthquake which will represent ground motion from the maximum earthquake likely to effect the site. Both g-values are based on the founding of (Class I Seismic design) structures on L_-____________--. ._. -l
competent bedrock' as stated by the applicant (page C.2-4, PSAR). See Section 4.2.5 for further information. 2.5 Ecology outside of the cities and towns which are located predominately along 1 the'Schuylkill River, the region is generally devoted to farming. The i applicant will supply additional information on the type, acreage and-f yields of pertinent local crops and the species, takes, and seasons for whatever sport fishing occurs on the river. The data accumulated i will be used in the environmental monitoring program for the site and will form the basis for a more precise estimate of potential exposures through critical pathways to man at the operating license stage. 2.6 Environmer.tal Radiation Monitoring The applicant has submitted a general outline of his environmental monitoring program and has indicated that a preoperational environ-mental program will be conducted. Although PEco will sample all of the important media and analyze for certain nuclides, the applicant has not documented the details of his program, such as frequency and location of sampling. In view of the need for more information, the applicant has been asked to provide additional details on the preoperational monitoring program. We expect these matters to be resolved prior to issuance of a construction permit. O
1
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I 2.7 Air Traffic Air traffic in the vicinity of the site has been analyzed; two small airfields for small aircraf t are nearby. . One field, owned by PECo., is about one mile to the northeast of the plant site. A few small air-
~
craf t use this field, but the runway is oriented so that its extension does not pass over the site. About 15 flights per day is the average for the Pottstown Municipal Airport which lies 4.3 miles northwest of the plant site. This airport has no scheduled airline service, but serves a charter service, flying school, and some privately-owned aircraft. the plant does not lie in the approach pattern for this airport and the runwsy's axis is oriented away from the plant. Commercial aircraf t to and from the Philadelphia area fly patterns cuch that most flights are above four thousand feet in the area of Pottstown. Some local consercial traffic controlled by Philadelphia baparture Control will be flying below four thousand feet; a hig'h traf fic day (July 2,1970) produced 25 such local flights. There are no holding patterns within five miles of the site. Flight frequencies, flight patterns, and type of aircraf t flying in the vicinity of the site are not significantly different from that in the general region around Philadelphia. We conclude there is no significant concern for plant safety occasioned by aircraf t flights in the area around Limerick. H V
~_~ _ _ _____.____---__-_-__.-____--___:______-
l 2.8 Railroad and River Traf fic The Reading Railroad main line tracks run along the north bank of the Schuylkill River and traverse the site about five hundred feet from and below the plant. About fif teen "through" freight trains (mostly j in the late evening and early morning hours), one local freight train, i and fourteen two to three car passenger trains (between 6 a.m. and 11 p.m.) pass through the exclusion area each day. The applicant will enter into an agreement with the Readin3 Railroad to allow the applicant to restrict access to the railroad right-of-way within the exclusion area in the event of an emergency. The Schuylkill River, itself, is not navigable to large boats becausa of the dans located downstream from the site. (In the legal i sense, the river is navigable downstream of Phoenixvi1Am.) Ust of 1 l small pleasure craft,.in warmer weather, does occur, but this activity is relatively minor in the region of the site. 1 I The applicant plans to have emergency plans to assure control over all portions of the exclusion area including those portions of the Reading Railroad and the Schuylkill River located within the exclusion area. The applicant has stated that structures at or near the Reading Rail-road, e.g., intake structures and pumphouse, which have safety functions will be designed to withstand effects of an explosion or
)
missiles resulting from any conceivable railroad accident occurring in vicinity of the facility (See Section 10.5). 7
i 3.0 REACTOR DESIGN 3.1 ' General 3 The reactor design of the Limerick Generating Station is similar to that of other plants listed in the comparison shown in Table 3.1. The nuclear fuel, the reactor vessels, and the thermal-hydraulic charac ' teristics of these plants are all basically the same. Differences
. are small and insignificant; a few differences are discussed in the -
following subsections. 3.2 Nuclear Design The nuclear fuel and its arrangement in the Limerick core is essen-tially identical to previously approved plants. Such differences which do exist are slight. For instance, the reactivity coefficient for fuel temperature (Doppler coef ficient) at rated power differs an
~
insignificant amount from that for Browns Ferry; e.g., -1.18 x 10 A k/k per *7 for Limerick versus -1.3 x 10'$ A k/h per 'F for Browns Ferry. This small difference makes an insignificant effect upon expected fuel performance during the reactivity (or power) transient. As in all the current designs, the nuclear fuel will provide " negative j reactivity feedback" that is suf ficient, in combination with other plant systems, to prevent fuel damage caused by abnormal operational l transients. i f
~
__ ?
I I i ne nuclear reactors compared in Tables 3.1 will use reactivity l control systems in a manner identical to that for Limerick. Temporary (first fuel cycle) boron - stainless steel control curtains, movable cruciform-shaped control rods, a variable recirculation flow rate which automatically accommodates demend for increase or decrease in power (for specified control rod patterns) to permit " load following," and a standby liquid control system (a solution of water and sodium pentaborate) are included in the Limerick design as well as the other listed facilities. Mechanical and electrical devices as'sociated with limiting positive reactivity insertion into the core include control rod velocity limiters, specified control rod operating patterns, limited control rod drive speed capability, control rod worth mini-miser program and its ro'd 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 rod patterns administratively permitted will limit rod worths to less than 0.01 A k/k. ne nuclear fuel design is for 19,000 HWD/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: one rod at 1.10 w/%, six rods at 1.54 w/%, fourteen rods at 1.98 w/%,
=
c__ and 28 rods at 2.58 w/%. The higher enriched fuel rods are designed and made with large diameter end plug-shanka which can only fit in the proper location in the upper tie plate of the fuel assembly.
'In the event lower enriched rods are accidentally placed in locations for highly enriched rods, the result is a fuel assembly with less than the standard fissionable loading.
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 sub-mitted provides a suitable basis to expect safe performance of the y reactivity control mechanisms and the nuclear fuel under normal i l operating conditions of the Limerick facility. 3.3 Thermal and Hydraulic Design The thermal and hydraulic characteristics of the Limerick reactor core are identical to those for Newbold Island, Browns Terry, and Peach Bottom Units 2 and 3. The same rated and design power levels are planned for each of these high power density cores. Core cooling systems for the facilities are identical: two recirculation loops and twenty jet pumps for each reactor. During normal, steady-state operation the thermal hydraulic design of the core will assure a minimum critical
9 heat flux 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 kW/ liter. We have reviewed four General Electric Company topical reports which
. relate thermal and hydraulic characteristics with fuel performance.
s The information in these reports is applicable to Limerick and other BWR's currently under review. The reports are: NEDO 10173, " Current State of Knowledge of High Performance BWR Zircaloy-clad UO 2 Fuel;" NEDO 10174, " Consequences of a Postulated Flow Blockage Incident in a BWR;" NEDO 10179, "ELfects of Cladding Temperature and Material on ECCS Performance;" and NEDO 10208, " Effects of Peel Rod Failure on ECCS Performance." The conclusions of the evaluation of these topical reporte are summarized below:
- a. NEDO 10173: Tnere is no experimental evidence demonstrating CE production fuel performance at power generation limits and expo-sures as proposed; however, there is no evidence to indicate a threshold for sudden catastrophic clad failure as a consequence of power or exposure.
- b. NEDO 10174: Information submitted was inadequate to permit evaluation of the ability of the reactor to scram soon after i i
significant fuel failure. In the evaluation of the "no scram" f l case, there was insufficient information to permit substantiation !
- - _ - _ . - _ . . _ - - - . _ _ - M
l I x of the CE conclusions that indicated that complete flow blockage of a single fuel assembly would not result in: (1) an incident capable of initiating failures in adjacent assemblies, (2) local high pressure production, or (3) excessive offsite doses.
- c. NEDO 10179: Estimates of the peak Zr-clad temperatures cannot be made with certainty based only upon experiments because of incon-sistent stainless steel bundle spray test results and'1ack of data on Zircaloy bundles. There was no correlation found between tests lL on ' stainless steel and Zircaloy bundles. The tests did~show that
- flooding of the bundles was very effective in cooling stainless steel clad test rods, but no test data were available on Zircaloy I
clad fuel for correlation of these results. (ne study and analy-sis of the Emergency Core Cooling System (ECCS) for BWR's as des-cribed in Section 6.1 considered the results of tests described in - NEDO 10179.)
- d. NEDO 10208: From this report the mechanics for fuel rod failure and the nature of rod perforations are understood and it is con-cluded that the flooding-mode cooling of the core can limit the maximum clad temperatures to less than 2300*F for high power density cores. The experimental verification on the capability for spray cooling at higher power densities with bulging cladding was established.
Current ACRS and Regulatory investigation of ECCS capabilities directly relate to items e and d above since these matters are of a r j
I' t : generic nature. Item b relates to another Topical Report, NEDO 10349,
" Analysis of Anticipated Transients without Scram" (ATWS) through the "no scram" condition. Further study and possibly experimentation may be needed to resolve adequately the problems raised in the conclusions expressed above. A report on ATWS is being prepared for the Committee for our Newbold Island review.
3.4 Reactor Internals 3.4.1 Design The reactor internal structures are classified as Class I' (seismic) components and will be designed to withstand normal loads, anticipated transients and the Operational Basis Earthquake within the. stress limit criteria of Article 4, Section III of the ASME Boiler and Pressure Vessel Code. 4 i Under the loads calculated to result from the Design Basis Accident, j l the Design Basis Earthquake and the combination of these postulated 1 events, Limerick reactor internal structures will be designed to the requirements of the C. E. Nuclear Steam System (RSS) Loading Criteria, "J Appendix C of the PSAR. The NSS loading criteria contain a number of design stress, deformation, f atigue and buckling limits some of which may exceed applicable limits as specified in the nuclear component codes.' However, during recent discussions, representatives of GE have stated that only those limits of Appendix C which are consistent with l
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)
l the limits specified in the nuclear component codes will be employed; v on this basis we find the design criteria for the Limerick reactor
~
internals acceptable. Additional documentation on the limits to be used is anticipated in the near future. We intend to follow the review of the " guide tube collapse" problem on Fermi Il and Hatch to incorporate pertinent modifications in Limerick when appropriate. 3.4.2 Dynamic System Seismic. Operating end LOCA Analysis Seismic loading on the reactor internals will be determined by means of a normal mode-time history analysis. We find this procedure acceptable. Dynamic loading due to normal and upset operating conditions will be computed by means of quasi-d;ms.mic' methodo Aased on ene measured vibration response of similar reactor designs. We are presently reviewing a summary of BWR vibration test histories and correlations of vibration test data with design predictions submitted as Amendment 19 (Proprietary) to the Quad Cities docket and referenced in Supplement 5 to Limerick. A request for the additional information required for us to complete our review of Quad Cities Amendment 19 has been sent to the General Electric Company: their reply is expected soon. Final accept-ance of the procedures proposed for determining the dynamic loading due to normal and upset operating conditions for the Limerick reactor
1 6 internals vill be withheld until we complete our review of Quad Cities Amendment 19. We expect to complete this review early in the post-construction permit' period for Limerick. Design loadings for the postulated Loss-of-Coolant Accident (IDCA) 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 concerns regarding the validity of this static analysis the applicant has stated (PSAR Supplement 5) that the natural frequency of the BWR internal structures is more than ten times the calculated frequency of the LOCA loads thus assuring no significant dynamic amplification. On the basis of the information submitted by the applicant we find this analytical method acceptable.
3.4.3 Vibration Control - Vibration monitoring of reactor vessel internals during preoperational tests and during normal operation has been accomplished on a few BVRs. General Electric Company topical report " APED-5433, Vibration Analysis and Testing of Reactor Internals" April 1967 presents the results of one test program. Application of a (draf t) Safety Guide requires, as a minimum, a confirmatory-type vibration test to assure that no vibra-tion of internals and no incidence of loose parts result during the operation of the facility. The appliaant has stated, but not formally I documented, that confirmatory testing or monitoring would be accomplished. We will require documentation during the construction period of the test program. 1 i J 4 _ _ _ _ J
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l J 1 1 I j' 4.0 REACTOR COOLANT SYSTEM l 4.1 General The principal design parameters of the reactor coolant system are shown in Table 4-1 for Limerick, Peach Bottom and Newbold Island. Table 4-1 Limerick Peach Bottom Newbold Is. Design Thermal Power, NW 3440 3440 3440 Design Vessel Pressure (psig) 1265 1265 1265 Design Vessel Temp., 'F 575 575 575 Total Core Coolant
~
Flow x10~ gate (fullpower)
, Ib/hr 102.5 102.5 106.5 full Steam power)Flow Ratg x 10' , (ib/hr 14.82 14.033 14.156 Normal Operating Pressure (in done), psig' 1020 1020 1020 Feedwater Flow Rate, x 10-6,.Ib/hr 14.099 13.30 14.117 Feedwater Temperature, 'F 420 376.1 420 Reactor Coolant Recircula- I tion Line:
Design Pressure, peig Inlet leg 1148 1148 1148 Outlet leg 1326 1326 1326 Design Temp., 'F 526 526 526 Recirculation Flow Rate, spa 45,200 45,200 45,200 As may be observed in Table 4.1, the differences among the listed parameters for the three facilities are few in number. Newbold Island l
- _ - - _ .___-__-_______A
i 1 1 1
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I 1 4 l and Limerick both have higher feedwater flow rates and temperatures than those for Peach Bottom. The higher feedwater temperature at Limerick is obtained by using an additional (high pressure) feedwater heater. The higher flow rates of feedwater compensates for the higher steam flow rates of Newbold Island and Limerick. There are no safety implications not already discussed in this increase of feedwater temperature. [See ACRS Report for Duane Arnold Energy Center, November 21, 1969, page 18.] 4.2 Reactor Coolant Pressure Boundary 4.2.1 Desian The reactor coolant pressure boundary (RCPS) will be designed as a Class I (seismic) system. Components of the RCPS will be designed to withstand normal loads, anticipated transients and the Operational Basis Earthquake within the normal and upset stress limits of the applicable codes cited below. Application of AEC code classifications is discussed in Section 4.3. The reactor pressure vessel and RCPB piping will be designed to the emergency and faulted stress ilmits of the applicable code for the loads calculated to result from the Design Basis Accident, the Design Basis Earthquake and the combination of these postulated events. The remaining RCPB components, for which emergency or faulted stress limits L_____ - --
i are not included in the ASHE Code for Pumps and Valves, will be designed to the requirements of the General Electric Nuclear Steam System (NSS) Loading Criteria, Appendix C of the PSAR. The NSS loading criteria ; I contain a number of design stress, deformation, fatigue and buckling 1 I limits which may exceed applicable limits as specified in the nuclear l component codes. However, for those components within the reactor l 1 coolant pressure boundary, the applicant will use only those limits l 1 which are consistent with the limits specified in the nuclear component l 1 codes. l We find that the design, fabrication, and inspection criteria discussed above are consistent with those accepted for all recently reviewed plants i of this type and form an acceptable basis for the design of the reactor ! coolant pressure boundary. 4.2.2 Pipe Whip Criteria The applicant states that a system of restraints and supports for recirculation and other piping will be so designed that reaction forces (pipe whip) associated with a circumferential break or split will not prevent: (1) scram, (2) isolation of either reactor vessel or primary containment, (3) adequate core cooling, and (4) maintenance of primary containment integrity. We find this criteria acceptable.
.4.2.3 Main Steam Line Isolation Valve Leakage Leakage .through the closed main stesa line (MSL) isolation. valves following a postulated loss-of-coolant accident is an uncertainty in calculating potential LOCA doses. The Millstone Unit 1 Technical Specifications require repair whenever the leakage on any closed valve reaches a test value of 11.5 SCFH. Low leak rates through MSL isolation valves on the order of 1.0 SCFH during the period following a LOCA may produce doses in excess of Title 10 CFR Part 100 guide-line values. The applicant has in progress a study to select a seal system. Under consideration are two concepts: provide a third isolation valve in the MSL with either compressed air or water to seal the volume between isolation valves; provide a water-seal system described in Supplement 5, PSAR, as a positive sealing of the verti-cal legs of the MSL's inside containment, upstream of the inner isolation valves. The applicant has committed to providing a sealing system.
4.2.4 Seismic Design of Main Steam Line Piping: The main steam line (MSL) piping from the reactor vessel out to and including the outermost isolation valve will be designed, fabricated and inspected to requirements of Class I (seismic) and ( AEC) Quality Group A. The main steam lines from the outermost isolation valve up I to but not including '.he turbine stop valve are classed as ( AEC) Quality Group D. To provide adequate protection against failure of this latter segment of MSL piping, the applicant has been advised that we require Class I (seismic) design from the outernost isolation valve 1
l up to and including the main turbine stop valve. The applicant's response is anticipated in the near future. 4.2.5 Primary System Pressure Relief The objectives of the pressure relief system are to limit any over-pressure of the reactor coolant boundary (reactor vessel and recir-culation lines) that might occur from abnormal operational transients. Automatic depressurization for small breaks of the primary system enables low pressure coolant injection or core spray system opera-tion. Eleven relief valves are provided that discharge to the suppression pool and perform the following protection functions: (a) limit overpressure and prevent safety valve opening, (b) augment safety valve capability, and (c) depressurize the crimary system following small breaks to permit CSCS operation. There are two (formerly four) safety valves that discharge to the drywell interior and function to prevent overpressurization of the primary system. The relief valves are self-actuating at preset relieving pressures, but can be operated indirectly to permit remote manual or automatic opening at lower pressures. Safety valve capacity for this rystem is based on a pressure rise resulting from a main steam flow stoppage (turtite trip) at operating condition, turbine pressure 980 psig (103% rated), l i 1 _ _ _ _ _ _ _ _ _ \
1
]
no steam bypass of turbine, and reactor scram due to high pressure. The analysis. indicates that a design safety valve espacity of 10 percent rated flow, in conjunction with a design relief capacity i of 60 percent rated flow, is capable of maintaining an adequate pressure margin below the peak ASME Code allowable pressure of i 1375 psig. An analysis of the relief valve capacity for a postu-lated severe pressure transient shows their adequacy to limit pressure rise. We have questioned the applicant on the reasoning and analysis behind the reduction from four to two safety valves. We expect a response in the near future. 4.3 Application of AEC Classification Groups The applicant has applied a system of code classification groups to those pressure containing components which are part of the resctor coolant pressure boundary and other fluid systems important to safety. These classifation groups generally correspond to the code classi-fication groups A, B, ~C and D developed by the regulatory staff. The codes and standards applicable to the components in each of the classification groups are identified in Table 4.3. We and the applicant are in general agreement on the application of the code classification groups for the reactor coolant pressure boundary and other fluid systems important to safety. For those l l l
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.;J i o systems, portions of systems, or components where the applicant's classification grouping differs from ours, the applicant has upgraded his classifications by provisions for a quality level substantially equivalent to that normally required by the applicable staff classi-fication code groups. Clarification of these upgraded classifications is documented in Amendment 6 Supplement 5 to the PSAR. For pressure-retaining cast parts of pumps and valves in systems which are classified as Group A or Group B, we find the non-destructive examination requirements in the ASME Pump and Valve Code for lines I over two inches up to and including four inches inadequate. To be acceptable, we require volumetric examination (radiography or ultra-sonic testing) of these pressure-retaining cast parts in lines over two inches up to and including four inches. Where size or configura-tion does not permit effective volumetric examination, surf ace exam-ination (magnetic particle, or liquid penetrant testing) may be substituted. Examination procedures and acceptance standards should conform with those specified in the ASME Code for Pumps and Valves. The applicant has been informed of this requirement. Applicant acceptance of this criterion is anticipated. We find that the system quality group classification as specified by the applicant and supplemented by provisions for upgrading quality
.I l
L f l levels, and additional nondestructive examination requirement discussed above are acceptable for this facility. 4.4 Reactor Vessel Material Surveillance Program and Fracture Toughness Criteria 4.4.1 ' Surveillance Program The material surveillance program is consistent with previously accepted programs. The requirements for specimens of reactor vessel
' belt-line" material were discussed with the applicant and the need for observing current trends and anticipating requirements of the pending AEC Fracture Toughness Criteria was pointed out. The appli-cant has been advised of the requirements already set forth as part of ASTM E-180-70. We conclude the applicant has an acceptable surveillance program.
4.4.2 F,racture Toughness Criteria Fracture control will be in accordance with the requirements of the ASME Boiler and Pressure Vessel Code, Section III which requires that the vessel be pressurized only above a temperature equal to the sum of the Nil Ductility Transition (NDT) 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 (Cv) impact test. However, recent fracture toughness test data indicate that the current ASME Code rules are not always suf ficiently conservative, i _____2__ - _ _ .
.- 4 3 ..-
I and may not guarantee adequate fracture roughness of ferritic mate-rials. While the Charpy V-notch tests continue to be useful in measur-ing the upper ' shelf fracture energy. value, the Cv specimens, generally, I do not predict correctly the NDT temperature. The latter, therefore, 8 must be obtained from other tests, such as the DWT test. Quite often, ] I also, considerable difficulty exists in defining, from the Cv test j
. . l curves, the transition temperature region in which fracture toughness ]
-1 of ferritic materials increases rapidly with temperature. In j addition, this transition temperature region shif ts to higher tempera-tures when the thickness of the specimen tested is increased (sise effect). )
In a meeting with the applicant we have described the proposed AEC fracture toughness criteria and advised the applicant that adequate fracture toughness data will be required to establish appropriate heatup and cooldown limits for this plant at the time of the operating license review. The applicant has stated (Supplement No. 2) that it is possible for many of the ferritic piping systems inside the drywell to be at a temperature as low as 40'F when the system pressure is in excess of 25 percent of the normal reactor operating pressure. The applicant has proposed to assure adequate fracture toughness for these systems by maintaining the lowest service metal temperature at least 60*F
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. 1 1
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above the NDT temperature as measured by the Charpy V-notch specimens. In view of the possible inadequacy of Charpy V-notch specimens to predict correctly the NDT temperature, we have advised the applicant that the NDT temperature should be verified by other tests, such as the DWT test. l 4.5 Inservice Inspection l The applicant is providing access to the reactor coolant pressure boundary in compliance with Section XI, " Rules for Inservice Inspection of Reactor Coolant Systems," ASME Boiler and Pressure i Vessel Code. For engineered safety systems beyond the limits of the reactor coolant pressure boundary, that are not presently .
),
covered in Section XI of the ASME Code, the applicant is providing j access to pipe welds, and pump and valve bodies in these systems. l i This committment has been made in answer to question 4.12 and is i provided in Supplement 2 of the PS AR. We conclude that the appli- ; cant is making adequate access provisions for the inservice j inspection of the Limerick Station. 4.6 Reactor Coolant System Sensitized Stainless Steel The applicant has indicated that no furnace-sensitized stainless steel components will be used within the pressure containing boundary or load-bearing members vital to the structural integrity of the ) l l k l l \ l J l l u ___ _ __ __ ,
l 1 I I reactor vessel and core. No furnace-sensitized stainless steel , components are to be used in piping. Cast material with less than I l 5% minimum ferrite will be used for hard-surfaced austenitic stainless steel' discs, seat rings, valve bodies with integral seats, and pump j bearing and seal components which may be subject to temperatures l that would sensitize wrought material. The applicant will not use any stainless steel for which deliberate additions of nitrogen has been allowed to enhance the strength of the steel. Additional pre- l cautions used will include rapid cooldown from solution heat treatment temperature and the use of low carbon stainless steel electrodes during welding. Ferritic noszlas and pipe ends will be buttered with low carbon stainless steel applied by the weld deposition technique; this procedure eliminates use of separate wrought austenitic stain-f less steel safe-ends which could become furnace-setsitized. We ! conclude that planning to avoid sensitization of austenitic stainless steel during the fabrication period is acceptable. l 4.7 Electroslag Welding The electroslag welding process will not be used within the reactor coolant pressure boundary. The applicant indicated that if their plans should be changed, the required data (process specifications, l variables, quality control procedures, heat affected zone, and weld l data) would be provided for evaluation. _ _ _ _ _ _ _ _ _ _ _ . - . _ _ _ _ _ _ _ _ *9=
J 4.8 Foreign Procurement To date no foreign manuf acturer has been engaged to design or fabricate I any component within the reactor coolant pressure boundary. , i 4.9 1eak Detection
.i Detection of leaks from the reactor coolant pressure boundary within f the drywell will be possible using airborne particulate sampling equip-ment similar to that now used at Dresden 2 for leak detection. Noble gases and iodine detection equipment will supplement the particulate detection capability. Other equipment in the leak detection f system will permit the measurement of drywell pressure, temperature, and sump liquid level. The combination of these measurements will enable the operator to reliably detect abnormal leakage in the dry well.
Sensitivity for leak detection on the order of a few cc/ min will result from the use of both continuous and periodic sampling of ; radioactive airborne particulate. Routine use of the equipment will provide an operational tool for maintaining controllable leaks such as from valve stem packings. Determination of the exact location of a leak will require detailed inspection of the containment internals. Leakage detection for vital fluid-carrying systems external to the j primary containment has been adequately provided for. Temperature, d 1 _ _ _ _ _ . 'l
pressure, and flow sensors with associated instrumentation and alarms are to be provided beyond the limits of the reactor coolant pressure boundary for the following systems (measured parameters in parenthesis):
- a. Main steam lines (high flow rate; low pressure at turbine' inlet).
- b. Main steam lines compartment (high radiation and temperature in vicinity of MSL).
- c. Reactor core isolation cooling system (RCIC equipment room temperature; RCIC turbine high steam flow, stesa line low pressure, turbine exhaust high pressure).
- d. High pressure coolant injection system (HPCI equipment room high temperature, turbine high steam flow, low steam flow, turbine exhaust high pressure).
- e. Reactor water cleanup system (high temperature in equipment rooms,
_ cleanup system high flow). In addition to the systems mentioned, the Reactor Building ventilation exhaust is monitored for high radiation which would signal a breach of h the nuclear system process barrier. Airborne radiation' sampling and I 1 monitoring is provided in the compartments housing the RER System and the Reactor Water Cleanup System. Finally, the RCIC and HPCI turbines ) and core spray pump areas will be esspled for airborne particulate i l activity. We find the leakage detection systems acceptable for the Limerick Generating Station.
n 5.0 CONTAINMENT AND STRUCTURAL DESIGN 5.1 Containment Design and Comparison with similar Plants The containasnt systems include the primary containment using the r pressure suppression concept and the secondary confinement which includes the reactor building, its recirculating (atmospheric ventila-tion) system, and the standby gas treatment system (SBGTS). The con-tainment configuration is similar to that for the Shoreham nuclear power station. The drywell is a frustumi.of a cone whose steel-lined reinforced concrete walls - are six feet thick. , The vapor suppression chamber is a steel-lined right circular cylinder of six feet thick
.l reinforced concrete located directly beneath the drywell. The drywell :
and wetwell are separated by a 3-1/2 ft thick reinforced concrete floor penetrated by eighty-five 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 Limerick Generating Station with those of Newbold Island and Peach Bottom is presented in Table 5.1. Both primary and secondary con-l tainments will meet, among others, the criteria for Class I seismic j design. 5.1.1 Primary Containment The vapor suppression concept for the reduction of pressure inside the primary containment following a LOCA has been used in the Limerick design as in other BWR facilities. There is a departure in the l i k;
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traditional BWR structural design in that the drywell and wetwell configuration is of the " oyer-under" type; i.e., the drywell is con-structed'above the wetwell and together they form a continuous, single
. structure for the primary cemtainment. De codes utilized in design of the primary containment are ACI 318-63, " Building Code Requirements for Reinforced Concrete" and the ASME Boiler and Pressure Vessel Code, Section III, Subsection B. W e 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 1/4-inch thick steel liner plate is designed to function only as a leak tight membrane.
Containment liner material is A-285 steel, Grade A, Firebox Quality, i for'1/4 inch plate and- ASTM A-516, Grade 60, for sections of thick- J ness greater than 1/4 inch. Connecting the drywell and wetwell are eighty-five straight pipe vents, forty-four feet long and two feet in diameter. The vente project a short distance above the reinforced-concrete drywell floor and extend into the wetwell so as to be submerged by about eleven feet of wet-well water. Each vent has a protective cap sounted about ten inches above its entrance to keep foreign objects from entering the vent and also to improve the fluid flow properties of postulated accident atmosphere. The vents are structurally tied together by beams to provide resistance against forces which would be developed during
the postulated LOCA. Installed on twelve of the vents are vacuum relief valves which are monitored in the control room for their position (open or normally closed). The check valves (nine inches diameter) are designed in accordance with the ASME Boiler and Pressure Vessel Code, Section III B, and are remotely testable. They function to relieve wetwell atmospheric pressure should it exceed the drywell atmospheric pressure during the long-term cooling phase of the loss-of-coolant accident. The relief pressure is set to keep the pressure dif ferential below 3 paid. To eliminate the chance of one of these relief valves being open during the LOCA, a nonna11y-closed, air operated gate valve will be installed in series with the relief valve and opened manually during the long-term cooldown, post-LOCA. Further discussion of details of operation will be required at the operating license stage. A removable steel pressure head caps the top of the trencated cone of the drywell. The attachment for the head is anchored in the vessel shell concrete and is welded to the drywell steel liner for leaktight integrity. Leak test of the head seal gasket is possible. Design temperatures are: drywell = 340*F, suppression chamber = 205'F. Design pressures are: internal = 55 psig, external = 5 psi. Deck design dif ferential pressure = 30 psig (55 psig-25 psig). Margins: 15% for containment internal pressure and 30% for deck AP. f 1 . _ - - _ - _ _ _ _ _ _ _ -
l Numerous areas of the primary containment design have been reviewed. Significant 'results, including areas which have not been resolved to our satisfaction, are discussed below:
- a. Post-LOCA Containment Pressure: An extensive review of the method of calculating the post-LOCA peak drywell containment pressure and determination of the margin to be applied for the design pressure has been made. Using a refined vent flow model, the General Electric i i
Company has recalculated the post-LOCA peak dryvell pressure to be 48 psig. The applicant has accepted our suggested margin of 15% to give a design pressure of 55 pois. 'the newly calculated peak drywell floor differentia 1' pressure is 23 psig. The new design value of 30 psig for the deck differential pressure results from the use of a 30% margin which we suggested to the applicant. As dis-cussed in Section 9 of this report, the LOCA containment pressure transient calculation has been documented by the General Electric l Company in the Topical Report NEDO-10320, "The General Electric i Pressure Suppression Containment Analytical Model," April 1971, and Supplement 1 to NEDO-10320, May 1971. Additional information supporting and extending application of the model has been pre-seated in Supplement 5 to the Limerick PSAR. We find the calculated pressures and margins acceptable.
- b. Drywall Deck Design: The leakage rates through the drywell floor 1
slab during r, postulated LOCA were discussed above. The applicant l
1
. a i
1 has indicated willingness to test for l'eakage of 'the drywell { floor. The method of testing (whether at low or high drywell pressure for qualitative or quantitative results) will be developed prior to the issuance of an operating license as a tech-nical specification requirement. The drywell deck will have a steel liner on its upper surface of the same quality as that on the containment walls. The structural details of the floor to wall joints and the method of floor to wall liner details were acceptable,
- c. Turbine Missiles: In the event of a turbine disc breakup, missile protection for the primary containment, control room, and the fuel pool is provided by structural barriers (reinforced concrete walls and roofs) and geometric arrangement of the facilities.
- d. Hydrogen Control. Post-LOCA: The applicant has accepted the AEC Safety Guide criteria for designing a system to control and keep the concentration of hydrogen below the lower flammable limit in the primary containment following a LOCA. See Section 6.2 for further discussion.
- e. Inerting: The applicant has indicated that the primary containment will be inerted with nitrogen gas during plant operation. See Section 6.2 for further discussion.
- f. Seismic Instrumentation: The applicant has indicated that strong
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)
motion seismographs and accelerometers will be installed on the j primary containment in accordance with our suggested criteria except where the physical configuration of the structure prohibits conformity with the recommendations on vertical alignment of instruments.
- g. Sacrificial Shield: The sacrificial shield design pressure will be based upon the calculated pressure buildup within the annular space between the reactor vessel and the shield. Shield wall blockouts will be installed to prevent movement of the recirculation piping or feedwater piping such that the effective area for water flow to the shield wall - reactor vessel annulus is reduced. The applicant has shown that the design of the shield wall annulus is dimensioned to prevent significant direct impingement of jet forces on the wall following a safe-end failure of the recirculation lines. Failure of nozzle to safe-end welds on smaller reactor coolant lines might result in significant jet impingement. Hence, the applicant analyzed the twelve-inch feedvater line break with jet forces and static pressure combined l l
to give a resultant stress within the current design value of J 70 paid.
- h. Primary Containment Penetrations: Penetrations of the primary containment are in accordance with current design criteria. The applicant has designed an instrument line isolation system in i
9
. 54 - .
l keeping with the intent of our Safety Guide No.' 11. Inside the l l containment a-1/4 inch orifice will be installed in instrument 'i lines as close to the instrument tapoff as possible. Outside,' ! and close-coupled to the containment, ~ a combination manual bypass - valve and improved automatically reopening azeess flow check valve will be installed. Position indication of the latter on one of 1 two local panels and a remote alarm in the control room will signal the operator of valve closure. ') l
- 1. Itain Steam Isolation Valves:- Isolation valves for the main steam lines are presently two per line, one inside and one outside the primary containment. The applicant has committed to the concept of sealing against MSLIV leakage during a post-LOCA situation either I by installing a third isolation valve outside the primary containment with a capability for pressurizing the inner volumes between the -
valves or,by some other equivalent system. Ihe layout of MSL piping has been made with sufficient space for the third valve. See Section 4.2 for more details. j . Containment Vacuum or External Overpressure: The applicant has eliminated external vacuum relief valves for the containment because his study showed that the maximum cooldown rate of the post-LOCA drywell atmosphere by containment spray produced no more than 2.5 poi pressure differential. The applicant will pro-vide a design overpressure capability of 5 psig. i-
+
j i
, k. Drywell Floor' Flooding: . To preclude a 'remo:ely possible buildup of . water on' the' drywell floor during a.LOCA to such an extent that the downconers may be choked, thereby causing higher than predicted peak pressures, the applicant will lower the height .,
of a. select number of vents so that these few could act as drains. Any resultant peak pressure increase will be kept within the design margin, according to the applicant. 5.1.2 Secondary Confinement The secondary confinement structure or Raactor Building will be designed and built to limit the release of airborne radioactive mate-- 1 rials and will provide for controlled release of building atmosphere i oo that offsite doses from the postulated design basis accidents i will be well below the 10 CFR Part 100 guideline valves. The Reactor. Building will enclose.both reactor units; however, each reactor unit i' ) L with its primary containment is physically separated from the other by a dividing wall. Basically, there will be three compartments: two snelosing the primary containment of each of two reactor units and one for the volume above the level of the refueling floor. The , Reactor Building, including the dividing wall below the level of the refueling floor, is constructed with reinforced concrete. With rein-forced walls and roof, the structure will have greater protective capability than the standard BWR secondary confinement design that uses structural steel members and steel panels above the level of l l l { L _ -_-_ . +
g__ - ; o the refueling floor. The Reactor Building will be designed to withstand internal pressures of up to 7 inches of water, forces from tornadoes or. earthquakes, blast effects from a railroad accident, and , missiles from tornadoes or rotating equipment to the extent that safety feature equipment inside would remain operational. The original design objective was to provide a leak-tightness that would limit in-leakage to 100% of the building volume per day at 1/4-inch water (vacuum) while operating the Standby Gas Treatment System. The applicant has further increased the capability of the Reactor Building 1.y changing the design objective in-leakage to 50% of the building volume per day. The following sections present information on ancillary features and equipment which contribute to the leakage control capability of the Reactor Building. 5.1.2.1 Reactor Building Recirculation System A Clasa I seismic design Raactor Building Recirculation System will be provided to assure reduction in t'ae activity of the contained atmo-i sphere in the event of the postulated loss-of-coolant accident or ] i the refueling accident. Within three to five seconds af ter receipt of an appropriate signal (low reactor water level, drywell high pressure, high radiation in the exhaust ventilation duct, or manual actuation) the Reactor Building is isolated from the outside atmosphere. The Recirculation System and the Standby Cas Treatment System (SBCTS) both automatically start on isolation of the Reactor Building. I ________-_-___-____-______-Q
i - f. I The recirculation of Reactor Building air, following an isolation signal, utilizes the normal ventilation system ductwork which has been sealed to prevent outleakage from the building. The normal ventilation fans are shut down and one of the two recirculation fans is started. The fan suction draws air from the areas above and below the refueling floor through a filter assembly and discharges throughout the building. The flow rate in the main duet to the fan is about 60,000 CFM. A small fraction, about 1000 CFM, of filtered air is exhausted downstream from the Recirculation System filters to the Standby Gas Treatment System (SBGTS) in order to maintain a slight negative pressure (1/4" water vacuua) in the reactor building and to discharge some radioactivity from the Reactor Building. The effect of the affluent radioactivity is discussed in Section 9.0 of this report. The Reactor Building recirculation equipment includes redundant full-capacity fans and filter assemblies. The filters are full-flow, and include pre-filters, HEPA filters, and charcoal filters. Tests will be performed to demonstrate applicant stated removal efficiencies of 99.9% for inorganic iodines, 85% for organic iodines, and 99.97% for 0.3 micron particulate. Our evaluation of the Reactor Building recirculation filters will allow efficiencies of 90% for elemental iodine and 70% for organic iodine providing that the filters are of a quality commensurate with the l 1 l j j u j _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ l
I I .
" standard" (not " super-filter") SBGTS filters. An imposed limitation on filter efficiency results from our assumptica that it of the iodines cannot be removed. The applicant is preparing calculations to determine the heat load and temperature on the filters during a LOCA. The charcoal bed filters will have water sprays to prevent overheating. Bypass of the recirculation system filters is controlled by operating the duct system at a positive pressure with respect to the reactor building atmosphere and by improved filter installation design.
Mixing of the Reactor Building atmosphere following isolation is assured since the Recirculation System supply and exhaust ductwork involves the entire Reactor Building. Mixing within the Recirculation System itself is accomplished to assure the SBGTS extracts only a small portion of any activity in the system; such mixing is accomplished by the suction ductwork design. We intend to follow this system closely to verify operation in conformity with design. 5.1.2.2 Standby cas Treatment System The Standby Cas Treatment System (SBCTS) provides controlled filtration and discharge of the secondary confinement air via the reactor building f vent. This Class 1 seismic system uses HEPA and deep-bed charcoal filters (" super-filters") to assure maximum removal of iodines and particulate. The charcoal bed depth is about.15 inches (normal depth is about two inches). Efficiencies for removal will be greater than 99.9% for inorganic iodines, 90% for organic iodines, and 99.97% 1 l h m_. _____:.____m__ a
- l
)
~i for 0.3 micron particulate, according to the applicant. For our accident i l- calculations, we' allow only 90% filter (removal)' efficiencies for both i
elemental and organic iodines. Also there.' is imposed a limitation on filter efficiency. resulting from the assumption that 1% of the iodines F cannot be removed.' The applicant is . performing calculations to determine the heat 1oad and temperature of the filters during a' LOCA. W ere are water
~
sprays for the charcoal ~ beds to prevent overheating. 5.2 Structural Desian
- 1he applicant has provided definitions and identification of the structures, aquipment, and systems that will be designed to Class I or Class II seismic criteria. Criteria for structural design not included in currently published codes or standards have been specified 4
for our review. Our consultants on seismic design are Newmark and Hall and specialists in the U. S. Geological Survey. As indicated in Section 2.4, the values of .12g and .06g are acceptable for the DBE , and OBE conditions for . structural design. I 5.2.1 class I (Seismie) Structures For Class I (seiscic) concrete structures, the factorO load approach is used for design purposes. For Class I (seismic) steel structures working strasses are used for normal operating loading combinations, while under a combination of nor. mal loads, design accident, and extreme environmental conditions increased stress limits of 0.9Fy for bending, 0.85Fy for axial tension and 0.5Fy for shear are used. These design approaches are acceptable.
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1 l' , The applicant's design of Class II structures will Se such that failure of such structures will not degrade the integrity of any Class I structure. According to the applicant, sore less essential portions of Class I structures may be Class II (seismic) if not related to loss of function and their failure would not render i the Class I structure inoperative. We find this criterion acceptable. The reactor vessel support is a typical steel-skirt to steel-ring-girder to concrete pedestal design. This design of the support is similar to previously reviewed supports and is acceptable. 5.2.2 Environmental Effects All Class I seismic structures, systems, equipment needed for safe shutdown, the primary containment, and essential hest removal systems will be protected from tornado effects. The design basis tornado with winds of 300 mph rotational velocity and 60 mph transnational velocity and a pressure drop of 3 psi' within 3 seconds will be used as design criteria. Tornado missiles assumed for design are similar to those previously accepted. As indicated earlier, the Reactor Building 1 i will be built to withstand forces of earthquakes or tornadoes, including i associated missiles to the extent that safety feature equipment will remain operational. 5.2.3 Containment Structural Design Analysis l l To carry tangential shears due to earthquake, tangential, diagonal reinforcing will be placed in the shell of the primary containment. l
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_ - _ _ _ _ _ _ _ _ _ __ -__-_ A
1 i 1 l: ' l t r l The two basic codes used in the design are the ACI 318-63 code and the ASME Boiler and Pressure Vessel Code, Section III, Subsection B. , The basic analytical method used for axisymmetric loads is the finite element method. Nonaxisymmetric loads are analyzed using a
, computer program developed by S. Ghosh and E. L. Wilson.* Principal stresses and strains are calculated from the program, and then converted to rebar stresses using ACI 505. The ASME code is used to establish a
{ basis for designing the liner, wherever applicable, and also to design the drywell steel head. The primary containment design analysis is acceptsble. The secondary containment will be designed by the same methods and to the same criteria as other Class I structures mentioned earlier in this report. 5.2.4 Testing and Surveillance During construction, rebar user tests will be performed on full section bars. Arc welding of rebar is not permitted, but if it becomes neces-sary to arc weld, it will be done in accordance with AWS D12.1 require-ments. Cadweld splice sampling for tensile tests is acceptable.
*Ghosh, S. and Wilson, E. , Dynamic Stress Analysis of Axisymmetric Structures Under Arbitrary Loading, University of California, Berkeley, Earthquake Engineering Rasearch Center, Report EERC 69-10, September 1969.
Liner sesus' inaccessible af ter construction will be provided with a leak-chase system. A minimum of 4% will be radiographer, or a minimum of 10% will be tested by magnetic particle inspection where radiography is not possible. Initial structural integrity tests of the primary containment will be conducted at 115% of the design conditions. A full design pressure test of the primary containment can be performed at any time during plant life when not actually in operation. As noted in Section 5.1.1(b) the requirements for leak testing of the drywell deck remains for evaluation prior to issuance of an operating license. Allowable leakage rates through the drywell floor slab were discussed, but the applicant has not committed to establishing a limit since he concludes that only a large drywell floor slab break (bypass) area (about 1.75 sq. f t.) would use up the design margin of 7 psi for drywell containment pressure. This area and associated leakage would be easily discovered and corrected, according to the applicant, during the structural and pressure integrity test. We consider this acceptable.
'Ihe reactor building (secondary containment) leak rate will be tested by isolating the building, operating the SBCTS, and measuring to assure en in-leakage rate no greater than 50% of the building free volume per day. The criteria for test of the primary and secondary containments as well as associated penetrations are acceptable, except that the
'Y,
I i details of the drywell deck test remain to be evaluated prior to issuance of an operating license. Likewise, surveillance requirements will be evaluated at the time of the PSAR review. 5.3 Seismic Design our consultants have reviewed the seismic design criteria and the results of analysis presented by the applicant for foundations, seismic response, desping factors, and seismic analysis of structures, piping, reactor internals, equipment, and critical items of control and instru-mentation. Our consultants indicate that the design and analysis procedures are in general in accord with the state of the art and incorporate an acceptable range of safety margin for the hazards considered. All Class I systems, components, and equipment outside of the reactor coolant pressure boundary will be designed to sustain normal loads, anticipated transients and the Operational Basis Earthquake within the appropriate code allowable stress limits, and the Design Basis Earthquake within stress limits which are comparable to those associated with the emergency operating condition category (within the yield strength of the meterial for membrane stresses). We consider that these stress criteria provide an adequate margin of safety for Class I systems and components which may be subjected to seismie loadings. I 9
,pyn
l I 5.3.1 Seismic Input The seismic design response spectra submitted by the applicant produce ] i a magnification factor of 3.5 between the period range of 0.15 to 0.5 I l acconds. Proposed structure and equipment damping factors are in accord-ance with those recosmaanded by our consultant. The response spectra ! I derived from the time histories to be used for design are adjusted in suplitude and frequency to envelope the response spectra specified for the site. We and our seismic consultants conclude that the seismic input criteria proposed by the applicant provide an acceptable basis for seismic design. 5.3.2 Seismic System Dynamic Analyses Hodal response spectrum multi-degree-of-freedom and normal mode-time history methods are used for all Category I structures, systems, and components. Governing response parameters will be combined by the square root of the sum of the squares to obtain the modal maximums when the modal response spectrum method is used. The absolute sum of responses is used for closely spaced frequencies. Floor spectra inputs to be used for design and test verification of structure, systems, and components are generated from the normal mode-time history method. A vertical seismic-system dynamic analysis is being employed in lieu of a constant vertical load factor for all structures, systems, and components with natural frequencies greater thas 30 Hz. We and our , i
seismic consultants conclude that the seismic-systes dynamic methods and procedures proposed by the applicant provide an acceptable basis for the seismic design. - 5.4 Comparison Between Limerick and Newbold Island Containment Concepts Consideration of the similarities and differences between the Limerick and Newbold Island Nuclear Generating Stations has centered on the capabilities of each containment system for retaining the leaked fission products. The ACRS in a letter dated September 10, 1969, to Public Service Electric and Gas Company, had specified certain requirements for Newbold Island containment of radioactivity. Subsequently, Public Service endeavored to fulfill these requirements. We have encoursged the Philadelphia Electric Company to analyze and adopt these ACRS requirements in the design of Limerick. The Special Report to the ACRS , dated April 21, 1971, titled " Comparison of Sites for Limerick Genergt-ing Station and Newbold Island Nuclear Generating Station" presented cite-related matters such as population distribution with distance and time, foundation design, seismology, flooding and meteorology considera-tions. Included in that report was the following listing (Table 5.3) which presents a comparison of the " extra protective features" for Limerick and Newbold Island. Reference to sections in this report provide for more detail. Following our review of the applicant's acceptance of the concepto advanced by the ACRS for Newbold Island, we j
}
find the Limerick containment design to be acceptable. I
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T_ABLE 5. 3 COMPARISON OF LIMERICK AND NEWBOLD ISLAND Extra Protective Features Limerick Newbold Island
- 1. Confinement of radioactivity in the Reactor Building:
- a. Internal Pressure (design) 7" water (positive) 2 psig
- b. Inleakage at .25" water 50% per day 10% per day
- c. Construction Reinforced concrete Reinforced concrete throughout throughout
- 2. Recirculation and Filtration
- a. Reactor Building Recircula- 60,000 CFM mixing 100,000 CFM mixing tion System and filtration and filtration prior to a metered prior to a metered 1000 CFM release to release to the SBGTS. Filtration dilution path and involves high effi- environs .
ciency HEPA and charcoal filters. (Section 5.1.2.1) .
- b. Standby Cas Treatment High efficency HEPA Incorporated into System (SBCTS) and charcoal filters the system for the for cleanup of the Reactor Building 1000 CFM Reactor above.
Building effluent prior to dilution and release, post-IDCA or during purging. (Section 5.1.2.2)
- 3. Main Steam Line Two MSL isolation Three MSL isolation valves. (Section 4. valves with air 2.3). A third valve pressurization with air pressure between valves to or equivalent system eliminate any will be installed. leakage through valves.
- 4. Main Steam Line Tunnel No pressure retain- Enclosed tunnel ing enclosure. with design pressure Opens to Turbine of about 2 psig Building. enhances leak detec-tion and filtration of steam from small leaks or breaks.
ni
- 6 7. -
Extra Protective Features , Limerick Newbold Island
- 5. Main Steam Line Flow Reduction Flow restrictor in Turbine bypass lines following MSL Rupture outside HSL to reduce flow reduced in diameter containment rate (standard). to 10" to reduce flow through lines during first few seconds of accident.
- 6. Caseous Radwaste Control (1) Off-gas treatment (1) Similar system includes catalytic contemplated.
H -0 recombiner, 2 2 cryogenic trest-ment to remove noble gases, hold-up for decay to give less than 1% MFC at discharge. (Section 8.2.2) (2) Use of " clean steam" (2) Same. 1 sealing for turbine gland seals and valve stems. (Section 8.2) I
- 7. Liquid Wastes Deuineralizer, Same.
evaporator, and ] holdup. Discharge at less than 1% NPC. (Section 8.2.1)
- 8. Reactor Vessel Integrity Designed to ASME ASKE Code plus Boiler and Pressure change of nozzle Code, Section III, velds for con-Class A, Reactor toured shape and Vessel.- Will follow use of two forged improvements devel- bottom head plates.
oped for Newbold Reduced size by i Island vessel. large nozzles. (Section 4.2). l [ u _ _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _ _ . _ . - _ _ _ _ - 3
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, 6.0 ENGINEERED SAFETY FEATURES 6.1 Easraeacy Core Coolina Systems (ECCS)
{ The core standby cooling systems consist of the High Pressure Coolant Injection System (HPCIS), the Automatic Depressurization System (ADS), the Core Spray System (CSS), and the Low Pressure Coolant Injection System (LPCIS) which is one mode of operation of the Residual Heat
- c Removal Systes (RNRS). The HPCIS and ADS are high pressure systems while the CSS and LPCIS operate at relatively low pressures. The vertous systems are initiated by a high drywell pressure signal or a low reactor vessel water level signal (except for the ADS which requires coincidence of the two signals.) Table 6.1 presents a comparison of the designed ECCS capabilities for Limerick, Newbold Island, and Peach Bot ton Units 2 & 3.'
The ECCS for Liastick are functionally the same as those for other BWR l
, facilities for which construction permits have recently been issued. k The ECCS are designed to limit the peak fuel cladding temperature in the event of accidental loss-of-coolant from the reactor coolant system.
The core cooling capabilities of the ECCS span a broad spectrum of reactor coolant system break sizes from a small area which does not Q rapidly depressurize the reactor vessel to the largest area of 4.205 square feet which results in a rapid depressurization of the reactor vessel. This lares break area is an effective area obtained by
h TABLE 6.1 COMPARISON OP ECCS CAPABILITIES l Limerick Newbold Island Peach Bottom !
)
HPCI No. of pumps 1 1 1 ; capacity, spm 5,000 (100%) 5,000** 5,000 1 Head, pside 1,120 to 150 1,120 to 150 1,120 to 150 Backup ADS + CS and Same Same i LPCI LPCI No. of Pumps 4 4 4 3 Capacity (ea.) gpa 10,000 (22.3 10,000 10,000 ' efs)
% of Required Total, (ea.) 33 1/3 33 1/3 33 1/3 Head, poid* 20 20 20 Backup CS CS CS Core Spray No. of Loops 2 2 2 No. of Pumps 4 4 4 Capacity (ea.) gpu 3,125 3,125 3,125
% of Required Total.
(ea.) 50 50 50 Head, psid* 122 122 122 Bee.kup LPCI LPCI LPCI Aukusatic Dep'lessuri-zation System No. of Relief Valves 5 5 5
% of Raquired Total (ea.) 25 25 25 Capacity (ea.)
lb/hr 800,000 800,000 800,000 Head, psid* 1,100 1,100 1,100 Backup Remote - Same Same Manual Relief Valveo
*psid = pounds per square inch differential between reactor vessel and priraary containment.
** Normal run of HPCI line moved from reactor vessel to core spray line to provide additional margin for cooling cladding early in the heat-up transient associated with LOCA.
',)
2
I i ! summing the areas of the completely severed suction line to the recirculation pump 3.667 ft ) and the effective. area of ten jet pump nossles ' (0.538 f t ) . The ECCS also functions to limit the peak fuel . cladding temperatures in the event of a main steam line break inside the drywell. , 1 Sufficient redundancy and reliability have been provided for sensors and i associated controls and instrumentation of the ECCS that no failure of a single initiating sensor or control device will prevent starting one , of the cooling systems. The controle and instrumentation can be tested and calibrated to conditions representative of an accident. The ECCS equipment and piping will meet the requirements of a Class I seismic design. A loss of offsite power will not prevent ECCS operation. Pro-visions have been made to provide capability for testing the sequential operability and functional performance of each system. The ECCS pumps which will provide cooling water to the various systems will be designed and installed to eliminate dependence on containment pressure to achieve adequate net pump suction head. Core cooling water supply for the pumps is provided from the suppression pool and the con-densate storage tank. The applicant has stated that a break in any one of the suction lines leading from the suppression pool to the ECCS pump will not lead to draining the suppression pool since the ECCS pump roome
.g
-_____-______--___m
s
)
are water-tight enclosures. Hence, a ptsap room flooded by water from a break will equilibrate with the suppression pool after only a small drop in the level of the pool. . i
)
As discussed in Section 3.3 of this report, Geostal Electric Company ! topical reports that are under review at this ties contain information . on.the expected temperature response of the fuel whee cooled by " sprays" (Core Spray System) and/or " bottom flooding" (Iow Pressure Coolant In-jection System) during the loss-of-coolant accident. The results of I ECCS performance review previously indicated that the designeo ECCS l i capabilities were sufficiently conservative to assure that the fuel cladding never exceeds 2300*F. The results obtained by General Electric in calculating the effect of the Emergency Core Cooling Systems acting either singly or in combinations, with heat transfer correlations for varied flow and fluid conditions in the core, indicated the peak clad temperatures did not exceed 2300*F. The ACRS questioned the con-servativeness of the (IDCA) core fuel rod heat transfer correlations being used by General Electric for (single mode) core spray cooling that occurs during the period of " lower plenua flashing (swell)." We questioned other influsatial assumptions such as the time needed to ruvet the fuel assembly channel wall (channel quench time), the lack of steam quality as a variable in the heat transfer coefficient correla-i tion developed by General Electric, values of emissivity for fuel clad q i. l l l l II l 1
and channel box, and the sensitivity of fuel cladding heat transfer to parametric variations (heat transfer coefficients, e:uissivity, wetting tino). Using the response to these and other questions as well, the AEC regulatory staff completed its study of the BWR ECCS capabilities and has developed an AEC policy
- for review and accept-ance of the design of ECC systems. Ve have requested a reevaluation of the ECCS for Limerick based upon postulated accident conditions and assumptions presented in the AEC policy statement. The applicant's revaluation will be reviewed for acceptability prior to issuance of a construction permit. The currently accepted peak clad temperature is 2300* F.
For long-term cooling of the core following a IACA, a continuous heat removal capability is provided. The ECCS will remove the decay and residual heat from the core under DBA conditions using suppression pool water. If the latter is unavailable for some reason, RHR ser-vice water is available through cross connections. Under low river flow conditions, as discussed in Section 2.3, the applicant will pro-vide assurance of an adequate supply of water for the RRR Service Water System and the Emergency Service Water System (ESWS), that provide the cooling water for ( *..e RHR System and emergency equipment, respectively. The RHR Service Water System, the ESWS, and their structural enclosures,
*" Interim Acceptance driteria for ECCS for Light-Water Power Reactors,"
U. S . A.E.C. , June 19, 1971 i
! 1 1- l L l
. i 1
1 i are to be designed to Class I seismic standards with adequate redundancy of equipment and pipelines to assure adierence to the
" single failure criterion."
Our review o'f the Newbold Island CP application resulted in changes to the LPCI and HPCI systems proposed for Newbold Island. We have I reconnended sintilar changes for Liastick and anticipate that the ' applicant's response to these changes will be docenented in the near future. The changes are discussed below: I LPCI: The changed LPCI system will provide four separate and inde-pendent paths to deliver cooling water inside the core shroud of the { reactor vessel. This change eliminates a " single failure" problem associated with the selection and operation of the LPCI injection valve to the (unbroken) recirculation loop. Also, direct introduction of cooling water inside the core shroud will provide a reduction in peak cladding temperature. HPCI: The changed RPCI system will introduce the cooling water into the core spray sparger lines instead of the feedvater lines. Also, the I HPCI steam turbine will be modified for operation se higher capacity to give an increase in HPCI cooling water flow rate. This change extends the small break regime, for which no rod perforations occur, 2 from 0.2 ft to 2.0 ft . i l
t l 6 '. 2 Hydronen Generation in Primary' Containment Followina a IDCA 6.2.1 Hydronen Control System To eliminate the possibility of hydrogen buildup during the period following a loss-of-coolant accident, the containment atmosphere will be processed through a catalytic recombination system to remove hydrogen and oxygen. The recombiner trains, located outside the primary con-tainment, will pass containment atmosphere through a halogen profilter (to remove iodine and its compounds), a catalytic recombiner, and a cooler-condenser to remove water vapor and to cool the remaining gases for return to the containment. The ABC design guidance, Safety Guide ! No. 7, will be used for the design of this system. We find the description of this system adequate for this stage of the review, but will require further information on design details and operation to assure its capability for keeping hydrogen concentrations in the post-IACA containment belot, the flammable limit. This information can be I reviewed during the construction of the facility. 6.2.2 Containment inertina As an operational technique to reduce flammaable gas concentrations, I the containment will be inerted with nitrogen. Information to be pre-sented in the PSAR will further amplify the description of this technique. W. ,
6.3 Corrosion Cracking of Pipe Metal Potential for cerrosion cracking of pipe metal increases wherever lines contain oxygen or other gases in a stagnent volume. To avoid this condition, the applicant will keep the oxygen content in feedwater 1 to the reactor below 200 ppb by de-seration in the condenser. The project design personnel will perform a continuous review of piping layout, ele- i vacion, and plan diagrams to try to eliminate high points where gas could collect. In appropriate piping runs, piping will be sloped towards' ( the reactor and remaining areas of concern such as thermal sleeves and i sparger connection boxes will be designed with a gas escape passage. Sufficient width and root radius will be provided to avoid gas collection where crevices cannot be avoided. We find this approach acceptable. 1 s __ - _ - _ - - - - - - - - J
7.0 INSTRUME!frATION, CONTROL AND ELECTRIC POWER SYSTDS 7.1 Instrumentation and Control Systems The instrumentation and control systems have been evaluated against the Commission's General Design Criteria (CDC) dated February 20, 1971 and the Proposed IEEE Criteria for Nuclear Power Plant Protection Systems (IEEE-279) dated August 28, 1968. A comparative review was made with l i the Edwin I. Hatch Nucicar Plant, Unit 1. The reactor protection system i and the instrumentation which initiates and controls the engineered safety features were found to be functionally the same as those proposed for Hatch. The evaluation is limited to the BWR generic problem areas l and to those areas for which new information has been obtained. Specifically, these areas are: i i
- a. BWR Generic Problem Areas
- b. Post Accident Monitoring Equipment
- c. Environmental Testing, and
- d. Engineered Safety Features Testabili.ty.
7.1.1 BWR Generic Problems he following BWR generic problems were evaluated: a. Auto-Relief System Interlock, (b) Rod Block Monitor, and (c) Tiov Reference Scram. i
)
I he criteria for these systems were found to be identical to those of l the Hatch, Brunswick, and Duane Arnold plants, which are acceptable. 3
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Two other BWR generic problems under review by the staff are cosmon mode failures and anticipated transients without scram. We do not anticipate completing our review of common mode failures in time for the Cosunittee's consideration of the Limerick application.. In the matter of anticipated transients without scram we expect to provide our assessment of this BWR generic problem in our supplemental report applicable to both Newbold-Island and Limerick. 7.1.2' Post Accident Monitoring Instrtamentati_o_n We have reviewed the applicant'.s description of instrumentation for monitoring and recording key operational variables during and af ter a loss of coolant accident. We find that instrumentation provided for this function is acceptable. 7.1.3 Environmental Testing
- The applicant has identified electrical equipment located in the primary containment which is required to operate during and subsequent to an accident . The qualification test procedures and test conditions to assure that the components will perform as required were adequately identified.
Instrumentation, controls, and emergency power ecuipment that are required to be functional during or af ter a design basis earthquake will be analyzed or vibration tested to assure that the design require-f ments will be met. All equipment supplied by Geceral Electric ths.t is il y
- ' i l
required to be functional ~ will be tested. All equipment supplied by. Bechtel, where practicable, will be teste'd. We conclude that the criteria for environmental testing is acceptable.
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l 7.1.4 Engineared Safety Feature Testing Periodic testing of the engineered safety features presented in the PSAR requires the same degree of on-line testability as is required for the reactor trip system. We conclude that the applicant's commit-ment to this requirement is satisfactory for the construction permit stage of ~ the review. 7.2 Elect rical Power Systems 7.2.1 Offsite Power Limerick Units 1 and 2 will be interconnected to the Pennsylvania-New Jersey-Maryland system through 220 kV and 500 kV transudssion sys-tems. Power from Unit 1 feeds a 200 kV substation and power from Unit 2 feeds a 500 kV substation. The substations are approximately 3000 feet apart and interconnected by a 500-220 kV bus cie transformer and trans-udssion line. The transmission line is on its own independent set of towe rs . Each substation is arranged in a breaker-and-a-half confi-guration. Five transmission lines emanate from the station, all on individual sets of towers. Two lines leave the 220 kV substation on separate rights-of-way and three lines leave the 500 kV substation, one i I I J
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on one right-of-way and two on another right-of-4tay. Sufficient distance exists betwen any two rights-of-way that assures any one event will not ) l physically affect any other.one.' Two startup auxiliary power transformers are ' provided, one located at the 220 kV substation and the other located
'at the 500 kV substation. Each transformer is ' capable of supplying power to either unit.
The applicant has completed transient stability studies that have simu-lated the loss of each of the Limerick generators and the loss of the I largest generator on the 500 kV grid. The results have shown that loss j l of offsite power would not occur under these conditions.
]
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l Our review of the preliminary design indicates that GDC 17 can be met. j We conclude that the offsite power system will be acceptable. l 7.2.2 onsite Power - The engineered safety feature and safe shutdown loads are divided between
] . i four 4 kV emergency buses for each unit such that the operation of any I l
three for a unit will supply minimum safety requirements for that unit. Each bus in both units can be supplied power from the f 00 kV startup ! l auxiliary transformer or the 220 kV startup auxiliary transformer. ! These supplies have independent and separate auxiliary transformers for I reducing the. voltage. A third supply to each bus is from a diesel l I
\
1 l l l
lf L. generator. A total of four diesel generators are provided for both units and each is assigned to two emergency buses, one bus for each unit. Each diesel generator is started automatically by an emergency core cooling system initiating signal from either unit or by loss of offsite power. The accident loads are automatically sequenced on each diesel generator. The diesel generators will have a continuous rating greater than the electrical safety loads. Each diesel generator is housed individually in a reinforced concrete Class I seismic structure above the design flood level. Each diesel generator will also be a self-sustaining entity with its own independent lube oil, fuel oil, jacket water, lube oil cooling, air starting and control systems. Four diesel fuel oil storage tanks are provided, one for each diesel unit. Each tank has suf ficient fuel for operating its associated diesel unit for seven days. The 115 voit a-c systems provided for safety are arranged with two
.l physically separate reactor protection system buses for each unit.
The 125 and 250 volt d-c power supplies for each unit consist of two independent 125/250 volt, 3-wire systems. Each system has two 125 volt batteries, each with its own charger and distribution panel. The battery ) l chargers are capable of keeping the batteries fully charged and supplying ] t
- . _ _ _ - - _________ _ __ _ __ _ a
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the d-c system loads simultaneously. The two 125 volt batteries for each system are located in a separate room. There are four rooms on the sta-
< tion, each have separate and independent supply and. exhaust ventilation systems with two fire dampers. The batteries will be sized to supply essential loads for a period of greater than four hours on loss of its battery chargers during any plant operating or incident condition.
1 This d-c power supply arrangement provides for adequate separation and independence of redundant supplies for both units. Our review of the preliminary design of onsite power supply indicates that. Safety Guide No. 9 and GDC 17 can be met. We find that the pre-liminary design is not in accord with Safety Guide No. 6. The applicant has proposed the use of automatic switching of loads from one emergency bus to another bus whenever loss of an emergency power supply occurs;
.e.g.: loss of one diesel generator would cause its electrical loads to be automatically transferred to another bus for power. We do not con-sider this technique, the use of a " swing bus", to be in conformance with Safety Guide No. 6. Additional information on the matter is ,
anticipated from the applicant and we plan to be able to discuss this with the Comittee at its August meeting. 7.3 Design Criteria for Cable Installation and Identification The applicant's criteria were found to be acceptable in all but one i area: Criteria for minimum physical separation and barriers between l 1 S'l
82 - i redundant manual switches and relays located on boards. racks, and panels are presently under discussion with the applicant. Additional information will be provided and we expe'et to be able to discuss 'our conclusion with the Coosaittee at its August meeting. 6 l e o_----____----__________ _ _J'
8.0 AUXILIARY SYSTEMS 8.1 General We have reviewed the following auxiliary systems and have concluded that the design concepts are acceptable for their intended purpose.
- a. New fuel storage.
- b. Tools and servicing equipment. i
- c. Fuel pool cooling and cleanup.
- d. Service water.
- a. Rasidual heat removal service water (See discussion at Section 8.6) .
- f. Reactor building cooling water,
- s. Turbine building cooling water.
- h. Chilled water.
'i. Fire protection.
- j. Engineered safeguards heating and ventilating.
- k. Plant heating, ventilation, and air conditioning. ,
- 1. Makeup water treatment,
- m. Instrument and service air.
- n. Domestic and sanitary water.
- o. Plant equipment and floor drainage,
- p. Process sampling. ;
i
- q. Communication.
- r. Station lighting,
- s. Plant auxiliary boilers.
l l l 1 l ________-___-__m
l Significant aspects of the review of other auxiliary systems are i discussed in the following sections. 8.2 Radioactive Waste Systems The applicant's design objective for gaseous and liquid radwaste dis- , charge is to reduce the radioactivity in eff'luents such that the annual I average concentrations are less than 1% of 10 CFR Part 20 limits for each system. The liquid and solid radioactive waste disposal systems are similar to those provided for other BWR's. However, the gaseous radioactive waste control and disposal systes represents a significant departure from earlier plants since a cryogenic sysytes will be used l to provide auch longer holdup (and decay of the radioactive gases) prior to discharge. Another improvement will be the use of a non-radioactive steam source for sealing the main turbine gland seals. The redwaste and off-gas buildings will be designed to Class I seismic crite ria. These buildings are largely below plant grade; henca, any spills of liquid will be contained not only by the Class I concrete structure, but also by the low permeability bedrock. 8.2.1 Liquid Radwaste System There are separate subsystems for the collection and processing of the drainage or " drains" from plant equipment, floors, laundry, and those areas involving use of chemicals. The subsystems can be interconnected l 1
i
,.j.
f through existing cross-connections to provide flexibility in processink. q 1 The waste evaporators are available for processing liquid with high radioactivity or chemica1 impurities; the distillate is recycled to ,
- - the condensate storage tank for re-use while the evaporator concentrates j are processed through the Solid Radwaste System.
The system has the capability to process and re-use the majority of liquid wastes within the plant. The majority of liquid waste comes from equipment and floor drains which generally are of higher purity j (lower conductivity) than the laundry or chemical drains (high con-ductivity). The equipment and floor drains are normally filtered, domineralised,' end returned to the primary coolant via the condensate storage tank. - The chemical and laundry drains are normally discharged to the environs after evaporation and filtration, respectively. Liquid wastes being discharged to the environs are handled on a batch basis . The liquid batches are held for a period of time to allow com-plete mixing, sampling, analyzing,- and processing orior to the transfer to the condensate storage tank, solid radwaste storage, or to the environs . Discharge of any processed l'iquid is accompanied by dilution in the plant circulating water discharge pipe.
-2 The expected annual release of radioactivity in liquids will be 1.5 x 10 curies per year for the normal mode (chemical and laundry drains only).
If the floor drains are discharged because of high chemical content,
the annual average radioactivity concentration in the plant effluent will still be below 1% of HPC since the floor drains may be decontam-insted in the waste evaporator, if required. The applicant's study of operating experience of boiling water reactors such as the KRB reactor in Germany and the Dresden I reactor provided the basis for his consideration of trititse. A Public Health Service (PHS) publication on Dresden I indicates that the average radweste release rate for tritius in liquid for thia 200 Me(e) reactor was 0.05 microcuries per
. second and that the calculated ratio of (radwaste) tritium release race to tritium production rate in the fuel was 0.001.II} Tritium in the coolan. e is produced f rom neutron interactions with deuterium and from diffusion or leakage of fission product tritium through the fuel cladding. The amount of fission product tritium in the fuel, while greater in quantity than the neutron interaction sources, does not contribute significantly to the total quantity in the coolant because of the fuel cladding integrity and the low rate of tritium diffusion through the Zirca11oy-2 cladding. This latter conclusion was stated by both the PHS in their report and D. G. Jacobs in a monograph ( } on tritium sources. The applicant initially used this information to corelude that about 5.9 pCi/second production rate for II)" Radiological Surveillance Studies at a Boiling Water Nuclear Power Reactor", U. S. Department of Health, Education, and Welfare Public Health Service - Environmental Health Section, pages 85 and 86.
( ' Ibid, page 13. ( " Sources of Trititsn. . .", D. G. Jacobs, USAEC/ TID publication,1968, page 22. _ _ _ _ _ _ . - _ _ _ _ 1
O , d l l each reactor under maximum fuel leakage conditions would be the maximum. 1 However, subsequent operating experience from the KRB station indicated ) that numerous fuel failures had little affect on the tritium release rate. The applicant accordingly reduced the tritium formation rate to , l that resulting from neutron-deuterium interaction 0.3 microcuries per secced, which is believed to be the only major source. Observation and measurement of radioactivity levels and production rate for tritium will be followed during reactor operation to assure that technical specifi-cation limits, developed at the FSAR stage, are not exceeded. Adequate dilution water is available to reduce concentrations, onsite, to levels as low as practicable through controlled liquid radwaste discharge. 8.2.2 Caseous Radwaste System l The gaseous radwaste system has the main functions of recombining hydrogen ! and oxygen, condensing the steam, providing hold-up for radioactive decay , i of short-lived isotopes and removing krypton and zenon from the process s tream and providing for its s torage. To accomplish these functions, a catalytic recombiner subsystem, a cryogenic treatsent subsystem, and a storage subsystem have been provided. The sources of the radioactive gases that will be processed are the off-gases from the normal operation of the main steam condenser (fission product noble gases, activation gases, radiolytically formed hydrogen and oxygen, and air inleakage to the condenser) . The expected flow rate is about 263 SCm per unit. 6
l 1 l The recombinar for each unit includes a preheater, recombiner, and aftercondenser. Hydrogen will be injected into the off-gas flow up-stream of the recombiner to assure that osygen is completely removed from the flow. Without oxygen there will be me flammable mixtures. Of i concern is the possibility of a failure of the hydrogen control valve in a closed position thereby leading to a possible flammable mixture of hydrogen with oxygen. Hasearch on recombiners has been supported by the applicant. The design and testing of this subsystem will be followed closely during the post-CP period. , i
% d The cryogenic treatment subsystem provides for removal of krypton and xenon. Retreatment to reduce oxygen, CO2 , and water vapor, and a single distillation column to remove liquified noble gases are provided.
Liquified krypton and xenon are collected in the column reboiler. The remaining purified gassa are metered, monitored for radioactivity, and released with dilution to the atmosphere via the turbine building vent. The ultimate storage of the krypton and zenon in gas holdup tanks follows batchwise removal of the liquified gases from the subsystem. Storage is under ambient temperature and pressure. The off-gas treatment system is not an " engineered safety feature". It does include redundancy for operational aspects, but not primarily for safety. The supplier guarantees a 99.99% removal ef ficiency for krypton 1 l i i i 1
and xenon. The buildup of N-13 in the liquified nitrogert coolant will have to be accounted for in exposure calculations; however, the appli-cant is now considering use of a delay device wherein sufficient time for radioactive decay of N-13 and other short-lived isotopes may be achieved. The applicant has indicated that the catalytic recombiners of the gaseous treatment system will trap halogens so that no significant release is expected. Experience with the KRB BWR gaseous vaste treatment system indicates that this plating-out' of iodines on metals and on heater elements in the recombiner is to be anticipated. The particulate content of off-gases will be diminished by the scrubbing and filtering in after-condensers, filters, water-sealed compressors, water separators retreatment sections, and distillation coltans. The performance of these components in accomplishing the reduction of airborne particulate can be anticipated and will be verified. Since the applicant will operate the gaseous radwaste system contin-uously and has provided for a " clean steam" seal of the turbine glanda, using auxiliary steam, we consider the concept for control, collection, and processing of gaseous radwaste a significant contribution to l t l l
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g.' 'l l I l 1 4 fulfilling the requirement to release radioactivity at a level as low l as practicable. The applicant has agreed to provide information on the detailed design and development of the system as it progresses. The applicant has not attempted to include in the of f-gas stream other sources such as off-gas from the mechanical vacuum pump (10 times per year for four hours during startup) HPCI turbine condenser testing, and the gases resulting from ventilation of contaminated (low level) areas (equipment rooms, chem lab). The applicant indicated informally that the containment atmosphere may be recycled and stored to conserve nitrogen inerting gas; in such an event, clean-up through the SBGTS is adequate. In the other cases, the applicant maintains that the few occurrences and/or low level activities expected do not justi,fy attempting to process their very high volumetric flows in the of f-gas treatment system. Since there will be less radioactivity released in the off-gas discharge, these other sources of gaseous radioactivity become relatively more significant. Accordingly, we have asked the applicant to _ estimate contributions from all sources in order that we may evaluate their conformance to numerical criteria of 10 CFR 50 for "as low as practicable" releases. 8.2.3 Solid Radwaste System The solid radweste system is housed in the radwaste building and is designed for processing vet waste from water cleanup systems or
O radvaste processing, liquid concentrate from the vaste evaporator, and dry wastes such as filters, rage, clothing, and equipment parts.
. Radioactive materials contained in such solids will be packaged for i
shipment to an unauthorized disposal site. 8.3 Spent Fuel Storate Each reactor unit has a spent fuel storage pool capable of accommodating ' safely 150 percent of the full core load of fuel assemblies. Each pool is lined with stainless steel. No inlets, outlets, or drains are per-mitted that might allow the pool to be drained below approximately 10 ft above the top of the active fuel. Lines extending below this level are equipped with syphon breakers, check valves, and other anti-drainage devices. The spent fuel pools and the spent fuel s torage racks are designed as Class I seismic structures. A combination of physical arrangement and administrative procedures for spent fuel cask handling vill ensure that the cask cannot fall into the spent fuel storage pool or effect other equipment required for a safe plant shutdown. The reliance on administrative precedures to achieve this design goal will be described further in the FSAR. 8.4 Emergency Service Water System The Emergency Service Water System (ESWS) is designed to provide cooling l vater from the Schuylkill River intake structure to emergency equipment 1 l
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? under loss of offsite power conditions or loss-of-coolant accident. The system and related structures are of Class I seismic design, operable at river flood stage and at low river flow rates. The applicant has been asked to describe measures taken to prevent flooding o.f both ESWS pumps in the event of passive f ailure of a discharge pioeline. His documentary of protection is anticipated in an amendment to the PSAR that is to be submitted in the near future. 8.5 Main Control Room Ventilation System A ventilation and air conditioning system is provided inside a Class I seismic structure. A radiation monitor will provide the initial signal, on high radiation in the fresh air intake duct, that will result in the div'e rsion of the air flow automatically to one of two charcoal filter systems. Should smoke be detected in the ducts, a purge with 100 percent outside air may be manually-actuated. - The applicant's analysis of the exposure of control room personnel to radiation during the course of a postulated loss-of-coolant accident gave exposures within the design basis criteria of 5 rem whole-body and 30 ren thyroid exposure for personnel manning the control room over the first thirty days following the design basis accident. Our calculations discussed in Section 9.6 resulted in exposures in excess of the design criteria of 5 rem whole-body and 30 ren thyroid exposure. We will require the applicant to provide the capability for operation of the control room during the course of the loss-of-coolant accident. Resolution of this problem will l be presented orally to the committee. l t
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n 8.6 Residual Heat Removal Service Water Systen This system must provide a reliable source of cooling water for heat l-removal from the RHR System under post-accident conditions and also supply a source of water if post-accident flooding of the core or primary containment is required. The system is designed to Class I neismic criteria and is operable under conditions of high or low flows of the Schuylkill River. The system is operable when offsite power is not available since the electrical load can be supplied by the emergency diesel generators. The RHR Service Water Systes has a cross-connection to the RHR System for core flooding or containment flooding if required following a 14CA. An automatic throttling valve is provided in the piping outlet of an RHR heat exchanger to maintain the RHRSW at a pressure higher than the RHR System thereby eliminating the possibility of leaking radioactively contaminated cooling water from the shell to the river. The two independent closed loops, each with full pump and heat exchanger capacity, provide protection against the single f ailure cri te rion. We conclude that the RHR Service Water System is acceptable to meet emergency requirements for safe plant operation.
1 1 i 9.0 ACCIDENT ANALYSIS 9.1 General A number of heign basis accidents have been evaluated in the . course ' of our review of the Limerick Station. Tables 9.1 and 9.2 present a sumusary of assumptions and associated exposure doses for the postu-laced accidents. Pertinent details of each desip basis accident are discussed in the following subsecticos. The applicant initially did not use our assumptions to calculate the design " base" case accident exposures. However, following a request for such calculations, the applicant prepared data on exposures using AEC assumptions for each postulated accident, asespt that he assuand a filter efficiency of 99% for the recirculation system filter and for the SBGTS filter instead of using our lower efficiencies for these filters. Our calculations accout for the reactor building recirculation system, mixing and cleanup of reactor building atmosphere, and use of the
$8GTS wherever appropriate to the accident. As noted earlier, the fil-ter efficiencies we assumed are conservative. Likewise, meteorological values have been chosen conservatively since onsite meteorological in-formation has not yet been collected. Type F and 0.5 meter per second (stable) wind conditions were utilised for the " worst case" conditions; i.e., during the early hours following the accident. All of the 1
1 1 1 i f i d 1 . _ - - - _
. A
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- 95 accidents result in calculated doses well below the 10 CFR Part 100 guidelines. The accidents and assumptions used are described below.
9.2 Loss-of-Coolant Accident _ The assumption of the complete severance of the suction line to the primary water recirculation pump is the standard accident for this and other BWR's. The calculation of mass flow rate for the blowdown of the reactor vessel is based on an effective break area of about 4.82-sq. ft. which results from the sumsting of ef fective flow areas for the following fluid sources: recirculation line (3.667 sq. ft.), twenty jet-pump nossles (1.076 sq. f t.), and the reactor water cleanup line (0.087 sq. f t.). The applicant had used the General Electric Company's standard proce-dure for calculating reactor coolant blowdown and the subsequent pri-mary containment pressure transients. Peak calculated drywell pressure and peak calculated drywell floor differential pressure depend upon many assumptions for the thermal and hydrodynamic flow phenomena associated with reactor vessel blowdown and the suppression of pressure , I by steam condensation in the wetwell pool. We questioned the conser-l vatism of some of the assumptions used in the ensuing calculations in I several areas of which the most significant on pressure was the analy-tical model for the vent flow phenomenon. Hence, the applicant and in l l l 4
-1
1 turn, General Electric Company, were asked to justify or refine their analytical model used to calculate the primary containment pressure transient following a loss 4f-coolant accident.. In response, the General Electric Company refined their analytical model for calcula-tion of the post-LOCA containment pressure transient. The most signi-ficant change in the analytical model is the use of a mixture density term for the vent flow model which accounts for the mixture of water, gas (air), and steam which flows through the vents. This refinement of the calculational model has been doctamented by the applicant's forsal response to our questions. Also, the General Electric Company has pub-lished Topical Report NEDO-10320 "The GE Pressure Suppression Contain-ment Analytical Model", April 1971, and NEDO-10320, Supplement 1, May 1971. These topical reports will be followed by a proprietary document to be published later by GE. Our evaluatio'n of released airborne radioactivity and the resultant exposure doses was based upon our assumptions listed in Table 9.1. The assumptions include conservative values representing the capabilities of the Reactor Building, its Recirculation System, and the Standby Gas Treatment System. With these assumptions, the exposure doses were calculated and are shown in Table 9.2. These dose levels are well within the limits specified in 10 CFP. Part 100. The applicant has also calculated exposures resulting from the LOCA using design " base" i L.
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1 i l case assumptions as well as another set of assumptions which were a
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modified version of our assumptions. The expo'aures calculated by the applicant are also well within the limits specified by 10 CFR Part 100, i as expected, since their assumptions are less conservative than ours. i 9.3 Fuel Handling Accident (Refueling Accident) This accident uses the standard model: a fuel assembly is dropped by the refueling equipment onto the top of the reactor core from the maxi-mum height allowed by the equipment. The dissipation of the potential ' energy of the dropped fuel assembly results in damaging 111 fuel rods. Exposure dose calculations appear in Table 9.2 and are within the limits of 10 CFR Part 100. The applicant calculated the doses from this accident af ter modifying our aestamptions: 49 rods ruptured in-stead of 111 rods, and 99% filter efficiencies instead of our lower more conservative values. The Reactor Building Recirculation System and SBCTS mixing and cleanup capabilities have been included in our i calculations since the applicant has stated, following our query, that ; there will be adequate time to isolate the Reactor Building and to divert the normal ventilation flow to the Reactor Building Recircula-tion System filters before an unfiltered release of contaminated air can occur. Other assumptions are presented in Table 9.1. 9.4 Control Rod Drop Accident The standard model for this and other BWR's was used by the applicant to formulate the accident conditions. Based upon the postulated worst I case (reactor critical, hot standby), the analysis indicates that 330
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- fuel rods failed. Isolation of. the main steam lines to prohibit re-lease of contaminated steam will occur within the naximum allowed time of 10.5 seconds following the rod drop accident. Exposure doses in
' Table 9.2 were based upon a prescribed set of assunctions some of which' appear in Table 9.1'. The applicant ' initially did not use all of our assumptions for exposure dose calculations, but subsequently did calculate.the. exposures based on our assumptions. The applicant has stated, following our query, that radiation detection in the main steam line would initiate shutdown and isolation of the rechanical vacuum pump before the arrival of activity at the pump (should this pump' re-1 main open, the exposure dose calculation must be calculated using dif ferent assumptions). The exposure doses presented in Table 9.2 for this accident are well within the limits specified in 10 CFR Part 100.
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9.5 1M,in Steam Line Break The standard model was initially used by the applicant. The amount of l radioactivity released was calculated from the activity of noble gases l and individual iodine radioisotopes 'that pass through the main steam line during the maximum allowable closure time for the MSL isolation valves. The applicant assumed the noble gases and radioisotopes of W iodine discharged at a 0.1 Curie /second rate af ter a 30-minute holdup. The iodine concentration in the primary coolant will be set in the Technical Specifications such that any release to the atmosphere from l l l
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a break in any live steam line will not result in a thyroid dose at the site boundary in excess of 30 reas. A primary coolant' concentra-
' tion of I-131 of 0.06 pCi/cc would result, after 30 minutes holdup, in 1
a release of 0.1 curies per second to the off-gas treatment system. 1 i 9.6 control Room Exposure Doses During Accidents Using a design basis of 5 Ram whole body and 30 Rom thyroid exposure i doses as the criteria, the applicant has described measures taken to keep exposures within these limits. The exposure analysis perfonned accounts for radiation exposure throughout the course of a postulated LOCA. The applicant's analysis gave 0.44 rem (whole body) integrated l dose in the control room for 30 days continuous occupancy with a pri-mary containment leak rate of 0.635% of the volume per day. Assuming ' shift operation with normal rotation of personnel, their entrance and exit to the control room, no credit for breathing apparatus, and other accident conditions, the applicant obtained an exposure dose for, 30 days to a t.ontrol room operator of 1.7 Rams (whole body) and 0.24 Rem (thyroid). If personnel are exposed to unfiltered outside air due to a control room air purge operation for the period of one hour, the applicant's calculation results in a whole body dose of 0.16 Ram and
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s thyroid dose of 3.2 x 10 Ram. Our calculations of the control room exposures. differ from the appli-cant's. We used a meteorological dilution effect (wake factor = 0.5) l 1 I __. _-_-___ _ _ a
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) , 2 I for a building area of 2100 m and wind velocity of 0.5 meters per l second in our calculations. Personnel (control room) occupancy factors were taken as 1.0 for the first 8 hours, 0.0 for 8 to 24 hours, 0.33 for 1 to 4 days, and 0.20 for 4 to 30 days. With these assumptions, scaling the LOCA LPZ doses by considering the differences in atmospheric dilution, and allowing an additional reduction factor of 10 for halogen removal in the charcoal (intake) filter, we calculate that control room personnel exposure to filtered outside air gives doses in excess of the design criteria. Resolution of this difference in the results of calcu-lacion and steps to protect against the accident environment must be accomplished prior to issuance of the construction permit. The applicant is aware of the problem and action to resolve the difficulty will be reporteo orally to the committee. 9.7 Main Steam Line Isolation Valve Leakage Assuming one isolation valve fails to close in one main steam line following a loss-of-coolant accident, leakage through the other (closed) 1 valve may result in exposure doses which approach 10 CFR Part 100 l guideline values. The low flow rates which can cause exposure prob-lems cannot be eliminated by current steam line isolation valve design and testing concepts. Hence, the applicant has agreed to provide a method of sealing the MSL isolation valves. See Section 1 l 1 4.2.3 of this report for details. I l
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9.8 Instrument Line or Process Line Break Recent concern with the consequences of a breat. outside primary contain-ment of an "unisolatable" line 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 'I 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 rems. Such a limit in Technical Specifications combined with the applicant's agree-ment to conform to Safety Guide No. 11 result in an acceptable solution to the instrument line or process line break probica. 1 I l l l l l l l l 1
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- 10.0 ' CONDUCT OF OPERATIONS 10.1 Tecimical Qualifications _
The Philadelphia Electric Company is responsible for the design, ' construction, and operation of the Limerick Generating Station. The utility has acquired experience in the design, construction, and operation of numerous fossil fuel power plants, hydroplanes, diesel-engine and gas-turbine units, and other nuclear power plants. Peach bottom Atomic Power Station Unit No. 1,at gas-cooled nuclear facility 1 constructed by the applicant, continues to provide nuclear experience for Philadelphia Electric Company personnel. Peach Bottom Units 2 and 3 are boiling water reactors, each similar to those proposed for Limerick, that are now under construction by the applicant to meet
- scheduled operat'i ng dates of 1972 and 1973, respectively.
The applicant maintains an Engineering and Research Department staffed by several hundred engineers. As of June 1,1971, fif ty-three engineers i in Philadelphia Electric Company were assigned full or partial respon-sibility for the Limerick project. This staff includes personnel who have acquired training and experience in many phases of nuclear projects undertaken by the utility and other organizations. This technical staff is responsible for the review and approval of design features of the plant. This staff will prepare and conduct, with consultant aid, a quality assurance program and will follow field construction l of the plant until its completion (see Section 11 for details), l ! 5
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1 l 1 The applicant' has obtained the services of an Engineer-Constructor - 1 (Bechte1' Corporation) who provides enginee ring and construction I services and integrates the items supplied by the General Electric - - Company with the items supplied by others for the remainder of the , l plant. The General Electric Company will supply the nuclear steam supply system, the' nuclear fuel, and technical personnel for guidance ! and support of the applicant during start-up operations. Other sub-contractors and consultants have been engaged to provide expertise in fields of a specialized nature such as meteorology, hydrology, seismo-logy, geology, and environmental radiation monitoring. 10.2 Organization of Plant Management The responsibility for plant operations is assigned to the Station Operating Department, that has direct responsibility for operating all of Philadelphia Electric's large power generating facilities. The i Superintendent at the ' plant has prime responsibility for safe and reli-able plant' operation. Organized under the Superintendent are two groups: the Operating Group and the Staf f Group. As their names infer, the Operating Group personnel are in attendance at the station seven days a week, 24 hours each day on a shift basis. A normal shif t will consist l of eight men appropriately trained and experienced. The Staff Group provides support in technical, maintenance, and administrative duties. In defining plant personnel qualifications, the proposed " Standard i for Selection and Training of Personnel for Nuclear Power Plants q
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(ANS-3)" was used, but not rigidly followed. Organizational experience in operating Peach Botton Unit I with a plant organization einilar to the one proposed for Limerick provided additional guidance. We have compared the Limerick personnel qualifications with the ANS-3 require-monts and conclude that the proposed personnel qualifications for Liastick are acceptable.
- The Limerick superintendent, assistant superintendent, and other key personomi vill be assigned at least three years prior to initial fuel loading so that they say participate in the design and construction phases and the formal training program. Other personnel to receive
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training will be assigned at least twenty months prior to initial fuel loading. The proposed plant personnel training program will follow j J the pattern previously used by Philadelphia glectric Company. A comprehensive and continuing training program and schedule has been
- adequately described.
10.3 Operating Procedures l l' Written operating procedures will be developed for the plant. Deviations from approved procedures will be permitted only in accordance with the procedures established t.nder the Administrative Section of the Technical Specifications. All approved procedures will be avail-able in the plant control room and will incinde at least: l I d t 'g ; _ g;
O 108 - l Standard Operating Procedures i Maintenance Procedures Blocking Procedures Health Physics Procedures Emergency Procedures Refueling Procedures Technical Specifications 10.4 Startup, Reoperation, and Power Tests Plans for testing have been delineated. The several different cate-l gories of tests are discussed below: ' 10.4.1 Construction and startup tests < Bechtel Corporation and/or General Electric Company personnel or their subcontractors under technical direction of Bechtel or GE will perform these tests, most of which require formalized procedures, reports, and approvals. Philadelphia Electric Company will provide " surveillance" over these tests. Examples of these tests ares containment final leak rate test, system hydrostatic test, calibration and setting initial j trip set points on instruments, and relief /aafety valve adjustment. 10.4.2 Preoperational tests Included within this category are the construction and startup tests I just discussed and also tests of plant systems such as plant electrical i 1 Y
a 4
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systems, instrunent air systems, and makeup water supply systems.
, Such testing occurs before ~ completion of construction and provides opportunity for training plant operators. The applicant has listed.
the major systems and indicated' that. detailed test procedures for each must be prepared by General Electric or Bechtel for review and approval by the utility's management. 10.4.3 Startup and power ~ test program Some systems cannot be checked out during the earlier phases of the test program, but can be tested properly only with the . reactor at power. Nuclear characteristics of the reactivity control system, thermal and hydraulic characteristics of the primary system, and neutron monitoring capabilities will be checked with the reactor critical. The applicant's plant staff will perform these tests under the technical direction of a General Electric Company startup crew. Specific, approved startup procedures will restrict testing to the objective of proving that the plant is capable of operating safely and satisfactorily up to rated power. We shall review the startup tests and planned startup personnel staffing at the operating license 1 stage in greater detail. j 1 10.5 Emergency Planning and Plant Security . 10.5.1 Emergency plans The applicant has used the guidance in Appendix E, " Emergency Plans I o for Production and Utilization Facilities" 10 CTR Part 50 in presenting I l l.
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! l a description of emergency planning for the ticility. The plans l l have as "their primary objective the protection d the health and safety of the general public and of station personnel." Emergency procedures implementing the plans will be written for specific t emergency situations dealing with a spectrum of accidents, related both to radiation and non-radiation incidents. Security measures ! and actions associated with the prevention and control of civil dis- ! l
. orders and acts of sabotage are treated as non-radiation emergencies.
The applicant's emergency planning conforma satisfactorily with the j guidance and criteria provided in Appendix E,10 CFR Part 50. i 1 10.5.2 Plant security The applicant has included aspects of plant security or security l measures in the facility Emergency Plans. Security measures will l include perimeter fencing, single point access under control of a i security guard on 24-hour duty, use of badges for personnel and locking of outside doors on buildings. The applicant will prepare detailed procedures that will be integrated into existing company emergency plans for plant security. Adequate description of security I planning has been presented by the ' applicant. Additional review will l be accomplished at the operating license stage of the review process. l e _ - _ _ _ _ _ _ _ _ _ _ _ i
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l 10.6 Independent Safety Review of Operator Actions All work in the plant will need to be approved by either the Station Superintendent' or properly designated authority. here will be written operating procedures spproved for plant use. A Plant Operation Review
, Committee and the Operating and Safety Review Committee are organiza-tions at the plant and at the utility headquarters, respectively, to
- provide control over the initiation of in-house tests, changes in operation procedures, advice on any unusual or unreviewed safety .
question pursuant to 10 CFR 50.59, and review of any violations of the Technical Specifications. One member 'of the Operating and Safety Review Committee will be required to be a qualified consultant indepen-dent of'PECo. . The other members of this committee will not. be required to be members of the Limerick Generating Station staff and will be qualified per Section 4.6 of the proposed " Standard for Selection and Training of Personnel for Nuclear Power . Plants." Eence, provision has been made to oversee objectively, on a scheduled basis and as needed, the performance of the plant personnel. We conclude that the. plans for objective review are acceptable. 1 j; i 1 l i i l- - al
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.1 3 11.0 ' QUALITY ' ASSURANCE' The Philadelphia Electric. Company described their Quality Assurance ;
Program in ~ Appendix D of the Limerick PSAR. Our review resulted in requests _ for additional information that have been doctanented. This-added 'information covered a broad band of topics within the QA
- Program and' generally responded adequately to our questions. -One document that was not submitted for review was the so-called " quality assurance plan" which was defined as "a working document to define ~ in greater detail the responsibilities of Phils'delphia Electric, . General- i Electric-APED, and Bechtel Corporation and identifies the guidelines .
to be used in the Philadelphia Electric third level quality assurance auditing." ' This plan, other aspects initially documented in the PSAR, and the criteria of Appendix B,10 CFR 50, as implemented by the applicant were inspected by Region 1, Division of Campliance in December 19)0. Compliance found that the applicant's implementation of its Quality Assurance Program was unsatisf actory. During the management aseting af ter- the inspection, the applicant concurred in the Compliance findings, made commitments for corrective action, and requested a re-inspection. The applicant has engaged MPR Asecciates to advise them in the organization, writing, and implementing the details of their QA Program. d A
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- Since the December .19/0 CompJ.iance inspection, .the applicant has
, reorganized the QA personnel and prepared additional documentation of procedures for QA/QC actions. The QA reinspection by' Region I, Division of Compliance,. in April 1971 resulted in an overall finding of satisfactory response to the guidelines set - forth in the above-mentioned documents. With the possible exception disctased in the next paragraph.
the applicant's current quality assuranc program is acceptable. The applicant has presented information specifically describing the organization of the architect-engineer-constructor (Bechtel Corp.) for performing on-site quality assuraece. The Bechtel Project Field Engineer, Lead Field Engineers, Lead Quality Centrol Engineer, Discipline Quality Control Engineers have quality sssurance functions and responsi bilities. These persons also.have, to a varying degree, functions and i responsibilities related to performance of tasks on which they also exercise QA supervision or surveillance. There is an independent Bechtel audit / surveillance group, however, which mitigates the possible conflict of interest. Further study ano observation of the workings of this
. Bechtel organizational concept will be accomplished to assure adherence to the principles of Appendix B, Criterion No. I - Organizations,10 CFR Part 50. '
i The quality assurance requirements for Class 1 (seismic) structures, ' systems, and components are specifically stated in Supplement No. I to l 4 _ _ _ _ _ _ _ k
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the Preliminary. Safety Analysis Report. We believe that these quality assurancy provisions, that were implemented for all items designated
. as seismic Class I for design, comply.with the requirements of Appendix B,
' Quality Assurance Criteria for Nuclear Power Plants" of 10 CFR 50.
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12.0 YECiti41 CAL 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 had noted the lack of:
+ a section on Administrative Controls; a listing of important and/or i
uncommon words or phrases with their definitions; a plan for. including i details on the plant " Inservice Inspection Program"; a plan for describing the concept and procedures for a Site Environmental Monitoring Program; and a section which included limiting conditions for operation and surveillance requirements for core physica or reactivity effects. The applicant has responded adequately to pro-vide for plans in each of these areas of the Technical Specifications except for the Site' Environmental Monitoring Program. The applicant indicated that he does "not believe that such a program is properly l part of the Technical Specifications." We believe, on the contrary, that the program should be included as it has in the case of our recently-approved reactor facility Technical Specifications. The Technical Specifications will be reviewed in depth at the operating license stage of the review and will contain requirements for an 1 environmental monitoring program that is acceptable to us. i l l 1 I l l l
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13.0 ACRS CONCERNS Our review has considered concerns of the ACRS as expressed in its previous reviews of boiling water reactors. The PSAR indicatee where work on these concerns has been accomplished or is in pro- f j 1 gress. The following list presents what appears to be uncompleted work, unresolved problem areas, and completed action of significance. {
- a. Fuel rod intearity The staff has received and reviewed four General Electric Company topical reports that address four areas of concern j l
with the integrity of fuel rods supplied by the General l Electric Company. These topical reports were discussed briefly in Section 3.3. Our future plans to assure fuel rod integrity include the following:
- 1. Surveillance of G.E. production fuel to determine performance of the fuel with long exposures and high power generation rates. (NEDO-10173 follow-on).
- 2. Obtaining adequate supporting evidence to substantiate the manufacturer's claim that complete flow blockage of i
one fuel assembly combined with a failure to " scram" would ' not cause failure propagation to adjacent assemblies, local -- high pressure, or unacceptable calculated radiation doses 4 offaite. (NEDO-10174 follow-on). l l 1 ( > I 1 1 y
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3.- Evaluation of further test results of staf aless-steel and Zirconium clad rods to determine effects es peak cladding temperature during the spray mode (Core Spray) of th'e ECCS. (NEDO-10179 follow-on).
- 4. Evaluation of additional information on the models used to predict the ef festiveness of core spray cooling of high power density cores (NEDO-10208 follow-on).
- b. Hydrogen generation Analytical and experimental work has been performed to determine the extent of hydrogen generation and to provide a means for con-trolling hydrogen generation. Results of these studies were sub-mitted in Amendment 23 to the Dresden 2 SAR (Docket No. 50-237).
We recommend that appropriate means should be provided for mixing, controlling and sampling combustible gases in the containment following a LOCA without recourse to release of radioactive gaseous waste (purging). Additional information on the hydrogen control system proposed for Limerick is provided in Section 5.2.5 of this report. {
- c. Primary containment inertin2 Section 5.2.6 sets forth the requirements for inerting during reactor operation.
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- d. Anticipated transients with failure to scram pTWS)
This probicm has been under study by the General Electric Company and has been documented in their Topical Report NEDO-10349, " Analysis of Anticipated Transients Without Scram" dated March 1971. Ve have prepared a review of this topical report that will be provided to the ACR$ prior to the August meeting.
- c. Common mode failure study The General Electric Company has studied this problem and submitted a topical report, NI:DO-10819, "An Analysis of Functional Common Mode Failures in General Electric BWR Protection Systems."
Review of this report by the staff has not been completed. Results of the review will be applied during the construction period.
- f. Main steam line isolation valve leakage This problem is discussed in Sections 4.2.3 and 9.7 of this report,
- g. In-service monitoring of the reactor vessel - vibration and loose parts detection Section 3.4 of this report covers these problems in some detail.
The applicant has orally committed to performing a confirmatory-type vibration test. Additional details will be required during
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l the construction period to describe the scope of the testing. Formal documentation of the commitment also is needed. I h.- Adequacy of'onsite emergency radiation instrumentation'and adequacy .) of of fsite instrumentation and procedures during and following an_ ' accident We have reviewed the applicant's concept for providing onsite . radiation instrumentation for emergency use. Operational concepts and plans for offsite radiation manitoring in an emergency are being made by the applicant and the Commonwealth of Pennsylvania. We find the applicant's concept for emergency radiation monitoring ! I on-site and the actions being taken for emergency off-site monitoring are adequate; additional review will be made at the FSAR stage. Our review of procedures for offsite control of people reveals that P.E.Co. is taking reasonable steps to obtain cooperation and com-mitments of railroad, private industry, government (local and state) of ficials in the ef fort to provide satisfactory control in the event of an emergency. Emergency planning is discussed in Section 10.5 of this report.
- i. Adequacy of operational procedures that affect public safety The entire listing and description of " operational procedures" have not been developed for Limerick at this stage in the licensing process. The applicant has provided a listing of the documents that will contain the approved procedures for plant operation.
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An acceptable program for developing the procedures is evident. l Further discussion appears. in Section 10. J. Adequacy and independence of- operator safety review
' Section 10'.6 of this report presents information on this' topic.
Establishment of approved operating procedures, proper super-vision, and existence of two safety review bodies (one at the facility e.nd another at the utility headquarters) should assure adequate as well as independent operator and operational safety review for Limerick.
- k. Effectiveness of leakage detection inside primary containment Section 4.9 of this report containa information on this topic.
We conclude that the' system available should be adequate to detect the leakage of primary system coolant. It appears that the precise location of any leakage will be found only af ter detailed inspection of the primary containment.
- 1. Adequacy of operational staff Section 10.2 discusses this area. We find that the applicant, at this stage, is planning adequately for plant operation. Further review at the operating license stage will be accomplished to assure that suf ficient experienced and/or trained personnel are available.
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[. ; , L' , 121 - n, iffect-of loss of offsite power I- . . l Adequate capability to supply the plant emergency electrical
' loads'in event of the loss of offsite electrical power has been provided in cl.e design. See Section 7.2 of this report.
. n. Automatic depressurization system (ADS) .
. Redundancy in equiptaent and provision of an interlock between the ADS and the low pressure ECCS equipment (Core Spray pumps and LPCIS pumps) have eliminated concerns about the capability of ADS to reduce the reactor vessel pressure safely. The' ADS is adequately designed to fulfill require:nents.
- c. Testing frequencies of safety systems and adequacy of preoperational tests of vital systems and functions Each section of the PSAR contains entries' on test and inspection.
The specific details for frequency and scope of testing will be incorporated in the Technical Specifications that will be submitted at the operating license stage. The preliminary plans for pre-operational testing' of plant systems and functions are acceptable. Further review of this area vill be accomplished at the operating license stage. i 6 e m..
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- p. I:outine discharle of radioactivity _to t_he envJ ronment
- The appilcant will incorporate an of f-nas treatment system described in Section 8.2.2 to reduce the discharge of ' airborne radioactive gases to an insignificant quantity, well below the limits of 10 CFR Part 20. Procedures and equipment for routine discharge of liquid radioactive wastes -are acceptable. The applicant will meet the "as low as practicable ' radioactivity, discharge criterion. Continued review in this area vill be accomplished to assure fulfillment of design objectives and application of current regulatory criteria.
- q. Fire protection and fire prevention The applicant is providing for the prevention and protection of equipment and components against effect or onset of fire. Pro-visions in design of equipment and components, available fire quenching or deluge systens, and emergency operational procedures are incorporated into the design and intended plant operation.
- r. Capability of the biological (or sacrificial) shield to withstand forces associated with a safe-end rupture Section 5.1.1, subparagraph g. , contains information on this matter. The applicant haa presented information that indicates that the possibility of jet forces causing failure of the sacrificial shield is remote. The shield ie designed
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to withstand hydrostatic pressure in the annulus rel; ion cosined-with Jet forces from a 12" feedwater line break. This design is comparable to that of recently licensed BWR's. l
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- 14.0 CuNJORMAhC6 TO GdNERAL DESIGil CRITEHI A
- Based upon our evaluation ~ of the preliminary design and the design.
criteria' for the Limerick' Generating Station, Units 1 and 2,' we conclude that the applicant plans to provide a final facility design that will meet the intent of the latest AEC General Design Criteria. 9 , , f f L
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