ML20236B938
| ML20236B938 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon, 05000000 |
| Issue date: | 11/18/1969 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20236A877 | List:
|
| References | |
| FOIA-87-214 NUDOCS 8707290290 | |
| Download: ML20236B938 (50) | |
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30 30 H0VE November 18, 1969 CIDENSE AUTHORITY FILE CW l
SAFETY EVALUATION BY THE DIVISION OF REACTOR LICENSING UNITED STATES ATOMIC ENERGY COMMISSION IN THE MATTER OF PACIFIC GAS AND ELECTRIC COMPANY i
DIABLO CANYON NUCLEAR POWER PLANT UNIT 2 DOCKET NO. 50-323 t
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fa 8707290290 870721 PDR FOIA CONNDRB7-214 PDR b
lABLE OF CONTENTS PAGE
1.0 INTRODUCTION
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l 2.0 SITE 3
3.0 NUCLEAR STEAM SYSTEM DESIGN 9
l 4.0 CONTAINMENT 11 5.0 ENGINEERED SAFETY FEATURES 16 6.0 DESIGN OF CLASS I STRUCTURES FOR SEISMIC AND ACCIDENT LOADINGS 21 7.0 INSERVICE INSPECTABILITY AND SURVEILLANCE 24 8.0 ELECTRIC POWER SYSTEMS 26 i
9.0 RADIOACTIVE WASTE CONTROL 28 10.0 ACCIDENT ANALYSIS 29 11.0 QUALITY ASSURANCE 34 12.0 RESEARCH AND DEVELOPMENT PROGRAMS 36 i
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13.0 TECHNICAL QUALIFICATIONS 40 14.0 CONFORMANCE TO THE GENERAL DESIGN CRITERIA 41 15.0 REPORT OF THE ADVISORY COMMITTEE ON j
REACTOR SAFEGUARDS 41
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16.0 COMMON DEFENSE AND SECURITY 44 l
17.0 CONCLUSION
S 44 l
l APPENDICES i
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Chronology of the PG&E Application Review 47 B
Report of the Advisory Committee on Reactor l
Safeguards 48
TABLE OF CONTENTS (cont 'd)
PAGE C-1 Report of the U. S. Weather Bureau (Unit 1) 52 C-2 Report of the U. S. Weather Bureau (Unit 1) 53 l
C-3 Report of the Environmental Science Services Administration (Unit 2) 54 C-4 Report of the Environmental Science Services Administration (Unit 2) 56 D-1 Report of the U. S. Geological Survey (Unit 1) 57 D-2 Report of the U. S. Geological Survey (Unit 2) 60 E
Report of_the U. S. Coast and Geodetic Survey 63 F-1 Report of N. M. Newmark and W. J. Hall (Unit 1) 66 l
F-2 Report of N. M. Newmark and W. J. Hall (Unit 2) 78 G-1 Report of U. S. Fish and Wildlife Ser-vice (Unit 1) 84
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l G-2 Report of U. G. Fish and Wildlife Ser-vice (Unit 1) 89 G-3 Report of U. S. Fish and Wildlife Ser-vice (Unit 2) 91 11
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1.0 INTRODUCTION
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The Pacific Gas and Electric Company (PG&E), by application dated j
July 1, 1968, and subsequent amendments, requested a license to construct and operate a second pressurized water reactor at its Diablo Canyon site which is located in San Luis Obispo County, Califernia.
i The proposed reactor is designed to operate at 3250 MW(t) with an expected ultimate capability of producing 3580 MW(t).
The applicant has designed the major components including the containment structure and I
emergency core cooling system for a power level of 3580 MW(t), and has used 1
this power level in analyzing postulated accidents in conformance to the l
guidelines of 10 CFR Part 100. We have evaluated the containment and l
emergency cooling systems for 3580 MW(t); however, the thermal and hydraulic l
characteristics of the reactor core were evaluated at 3250 MW(t).
Before i
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operation at any power level above 3250 MW(t) is authorized by the Commission, the Commission mus t perform a safety evaluation to assure that the facility can be operated safely at the higher power level.
Several facilities and items of equipment will be shared between l
Diablo Canyon Unit 1, for which a permit to construct has already been issued, and the proposed Unit 2.
Our review of these shared facilities and items l
indicates that the independence of either Unit with regard to safety would not be compromised by a failure of the shared items. All shared items impor-l tant to public health and safety are suitably backed-up by redundant equipment.
Our technical safety review of the proposed plant has been based on the applicant's Preliminary Safety Analysis Report (PSAR) and the six sub-sequent amendments, all of which are contained in the application.
The
2 technical evaluation of the preliminary design of the proposed plant was accomplished by the Division of Reactor Licensing with assistance from the vari cus consultants, as reques ted. Within Reactor Licensing, the Reactor Projects group was responsible for the review, for coordinating j
parts of the review involving personnel with various special technical disciplines from the Reactor Technology and Reactor Operations groups within the Division, and for obtaining technical consulting on specific aspects of the review from groups outside the Division of Reactor Licensing.
In the course of the nyiew of the material submitted, a number of meetings were held with representatives of the applicant to discuss the proposed plant. As a consequence, additional information was requested of the applicant which was provided in certain of the amendments. A chronology of the review process is attached as Appendix A to this report. Appendices C through G include reports from our consultants on meteorology, cooling l
water dilution studies, geology, seismology, structural design, and j
radiological monitoring, respectively.
We also have attached to this safety evaluation for Unit No. 2 the reports of our consultants who advised us during the safety review of the Diablo Canyon Unit No. I facility.
The Commission's Advisory Committee on Reactor Safeguards (ACRS), in consideration of this project, has met with both the applicant and the staff. A copy of its report to the Commission on Diablo Canyon Unit 2 is attached as Appendix B.
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The evaluation of the proposed Diablo Canyon Unit 2 at' the cons truction permit stage is only the first stage of the continuing review of the design, l
construction, and operation of the plant. Prior to issuance of an operating license, we will review the final design to determine if all the Commission's safety requirements have been' met.
The facility would then be operated in l
accordance with the terms of the operating license and the Commission's regulations, under continued surveillance of the Commission's regulatory staff.
The issues to be considered, and on which the findings must be made by an Atomic Safety and Licensing Board before the requested license may be issued, are set forth in the Notice of Hearing issued by the Commission in this proceeding and published in the Federal Register.
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2.0 SITE
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2.1 Description The site for the proposed nuclear plant is adjacent to the Pacific Ocean in San Luis Obispo County, California. The site consists of
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approximately 750 acres near the mouth of Diablo Canyon Creek. The 585-acre portion south of the creek is leased to the company for a term of 99 years I
with an option to renew for an additional 99 years. The 165 acres north 1
of the creek is owned by PG&E in fee.
The minimum exclusion area distance
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i from the reactor to the nearest site boundary on land will be one-half mile.
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The nearest residence from the site is approximately 1-3/4 miles and the area out to a distance of about 6 miles is sparsely populated.
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4 six-mile distance was selected by the applicant as the low population zone i
radius. The neares t population center is San Luis Obispo.
Its nearest boundary from the site is ten miles, which is considered to be the popula-tion center distance.
2.2 Me teorology i
For evaluation of accidents the applicant has chosen diffusion parameters which correspond to a Pasquill Category F meteorological condition with one meter per second wind speed for short term releases, and meteorological con-ditions corresponding to a distribution of Pasquill Category C, D, and F for long term releases.
The Environmental Science Services Administration (ESSA) has evaluated the meteorological assumptions and has concluded that the assumptions chosen by the applicant are conservative. This conclusion is based on a review and analysis of the meteorological data taken by the applicant at the site.
The data, presented in Amendment 6 in response to the questions raised by ESSA in its August 11, 1969 comments, show that more than j
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95% of the time the actual meteorological conditions at the site are more favorable for atmospheric dispersion in the on-shore direction than those assumed for accident evaluations. The ESSA reports are attached in Appendix C to this report. The applicant's meteorological program includes meteorological measurements from a 250-foot tower near the plant location, from a 100-foot tower at the top of a 914-foot hill on the site, and at four other locations.
We conclude that the meteorological program is adequate to provide a basis for the development of a gaseous radioactive release limit and to confirm the conservatism of diffusion parameters used in the analysis of potential accidental releases.
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2.3 Geology and Hydrology The geologic features of the plant site were presented in the Preliminary Safety Analysis Report.
The U. S. Geological Survey (USGS) has reviewed the application and other available literature, and has examined the exploratory trenches at the site.
It has concluded that the applicant's analysis appears to be carefully derived and to present an adequate appraisal of those aspects of the geology which would be pertinent to an engineering evaluation of the site, The report of the U. S. Geological Survey (USGS) is attached
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as Appendix D of this evaluation.
Our s tructural consult:ints, N. M. Newmark, W. J. Hall, and A. J. Hendron, Jr., have recocrnended that analyses be performed of the stability of slopes that might present a hazard to the Unit.
Their report is attached as Appendices F-1 and F-2.
We concur in this recommendation.
Based upon these reports, with the understanding that stability analyses will be carried out, we have concluded that the geology of the site presents no j
unusual engineering problems for the construction of this nuclear facility.
The hydrological characteristics of the site were previously reviewed I
by the USGS for Unit 1.
It was concluded that the reactor location would not i
l be affected by floods of Diablo Canyon Creek, the only developed drainage nearby, except perhaps for part of the switchyard. It was noted that there were no reports of ground water developmentsin the vicinity of the site l
l and concluded that it does not appear that the reactor would affect fresh water resources of the site. These conclusions remain applicable for the Unit 2 facility.
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6 2.4 Seismology The seismic history of the Diablo Canyon area was studied by the applicant and its consultants and by the U. S. Coast and Geodetic Survey for the Unit 1 site. The results of this study apply to the adjacent Unit 2 site.
In summary, the applicant concluded that the following four possible types of earthquakes would result in maximum accelerations at the site:
Earthquake A: Magnitude 8-1/2 along the San Andreas Fault 48 miles from the site, resulting in a ground acceleration of 0.10 g at the site.
Earthquake B: Magnitude 7-1/4 along the Nacimiento Fault 20 miles from the site, resulting in a ground acceleration of 0.12 g at the site.
Earthquake C: Magnitude 7-1/2 along the off-shore extension of the Santa Ynez Fault 50 miles from the site, resulting in a ground acceleration of 0.05 g at the site.
Earthquake D: Magnitude 6-3/4 af tershock at the site associated with Earthquake A which results in a ground acceleration of 0.20 g at the site.
i For design pruposes, the applicant proposes to use both an envelope of the B and D response spectra and the B and D spectra separately.
The operational basis earthquake (OBE) would encompass Earthquake B with a l
horizontal acceleration of 0.15 g and Earthquake D with a horizontal acclera-tion of 0.20 g.
The use of two earthquake response spectra for the OBE I
was selected by the applicant because the frequencies of ground motion for the two earthquakes are different as a consequence of unequal attenuation du e to earthquake location. The applicant further states the design will I
be reviewed using 0.30 g and 0.40 g for the design basis earthquake for this j
site. The applicant has recommended, and we agree, that the same acceleration I
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values be' employed for the Unit 2 design.
2.5 Oceanography Condenser cooling water and auxiliary salt water cooling for the plant will be taken from the Pacific Ocean and returned through an out-fall via Diablo Cove. All of the Class I structures and equipment are located 60 or more feet above mean sea level with the exception of the '
suxiliary salt water intake structure.
The peak tsunami wave height, coincident with peak storm and high tide run-up, is approximately_18 feet above the mean lower low water (MLLW). The applicant will provide protection for this e.,quipment to an elevation of 30 feet above MLLW.
Tam maximum draw-down due to a tsunami coincident with low tide is -
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nine feet below MLLW.
The bottom elevation of the intake structure is 28.9 feet below MLLW and the auxiliary salt water pumps are designed to operate with the water level down to 17.4 feet below MLLW. Therefore, the ocean down to this level provides a reservoir for the auxiliary salt water system during a tsunami downsurge.
1 Liquid wastes containing small amounts of radioactivity will be diluted in the plant circulating water and discharged to the ocean at concentrations i
well within the limits of 10 CFR Part 20.
Wastes will be discharged on a batch basis and will be monitored and controlled to assure compliance with 10 CFR Part 20.
The applicant has also performed rhodamine dye tests to estimate the dilution capacity of the coastal ocean waters. The connents of our consultant, USGS, on the applicability of these tests are presented in Appendix D-2.
As a further check, the applicant's environmental monitor-ing program includes sampling of ocean water, sediments, and marine life-r 7
8 in the plant vicinity during plant ope rat ion.
2.6 Environmen t al__Monit oring The applicant proposes a preoperational environmental monitoring program which will begin about two years before operation of Unit 1.
Background radiation will be measured,- and samples will be taken of air particulate, bovine thyroids, milk and vegetables.
Sampling of marine j
life will include abalone, clams, rockfish, salmon, mussels, barnacles and
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bull kelp (seaweed). Bottom sediments and seawater samples will also be included. Af ter completion of the two-year program, it will be reviewed, modified where experience indicates that changes should be made, and j
l continued during plant operation.
The U. S. Fish and Wildlife Service, Department of the Interior, I
has reviewed the proposed environmental monitoring program and concluded that is is essentially the same as the program for Unit 1, and that comments contained in its letters of June 23, 1967, and January 3,1968, apply to the Unit 2 program.
In its letter of January 3,1968, F&W recommended that the environmental monitoring program proposed for Unit 1 include sampling of scavater and sediments.
These have been included in the proposed program for Unit 2.
Comments of the Fish and Wildlife Service are attached as Appendices G-1, G-2, and G-3.
2.7 Conclusions On the basis of our review we conclude the site ic acceptable for the proposed nuclear facility.
9 3.0 NUCLEAR STEAM SYSTEM DESIGN The nuclear steam supply system consists of a light-water-moderated pressurized water reactor (PWR) which transfers reactor heat to four steam generators. It is similar in design to other PWR systems that have been granted construction permits and operating licenses. There have been evolutionary changes in the design of Diablo Canyon Unit 2 since issuance of the construction permit for Unit 1 and the applicant states that these changes will also be made in the design of the previous unit.
The major design modifications in the nuclear steam supply system involve the emergency core cooling systems which are discussed in Section 5.1 of this report.
One other major design change is the change of control rod absorber material.
The control rods for Unit 2 will contain boron carbide neutron ab-sorbing material rather than silver-indium-cadmium as originally proposed for Unit 1.
The nuclear control characteristics of the reactor are not affected since the material substitution does not significantly change the rod worths. We presented questions to the applicant concerning the potential for control rod swelling and the applicant's response is contained in Amend-ment No. 5.
We have concluded that the change in absorber material is i
acceptable.
For Diablo Canyon Unit 2, we have further reviewed the applicant 's l
l proposed instrumentation to assure that the power distribution is adequately
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f controlled. The applicant has proposed that the four external flux monitors l
will detect abnormal power patterns. The in-core instrumentation for determining the power distribution in the Diablo Canyon core, as presently a
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1 10 proposed, consists of six traveling flux probes which may traverse any of 58 thimble locations, and 65 thermocouple located in guide tubes at the exit flow ends of the fuel assemblies. The in-core flux probes are not designed to operate in the core at full power for more than a few months.
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The applicant believes that the planned test programs (primarily at the Ginna and Indian Point 2 plants) will adequately demons trate prior to i
operation of Diablo Unit 2 the capability of the external long ion chambers to measure core power dis tributions at any time during core life.
It remains to be demonstrated that the proposed external monitors can detect l
incore power distributions and anomalies with adequate sensitivity,to assure that no loss of safety margins can occur.
In the event this cannot be shown, fixed in-core monitors will be required to assist the operator l
in positioning the part length rods for proper power distribution control.
In this regard, the applicant has agreed that the in-core monitors will be used with frequent flux scans, or other monitors could be semi-l permanently installed in the existing in-core thimbles for this purpose.
i We will review this aspent again at the operating license stage of our review.
On the basis of our review of the information provided by the appli-cant and on our previous reviews of similar PWR systems, we have concluded that the proposed nuclear steam supply system, including the nuclear design, the thermal and hydraulic design, the mechanical design, the plans for the t
instrumentation and control systems design and the secondary system design j
is acceptable.
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l 11 4.0 CONTAINMENT 4.1 Description l
The proposed containment vessel is a reinforced concrete structure with a steel liner which encloses the reactor and reactor coolant system.
The internal concrete structures will be designed to provide missile protection for the primary coolant systems in the event of failures of valves, including valve stems and bonnets, instrument thimbles, closure l
bolts, and complete control rod drive mechanisms. The design will aisc
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include provision to withstand forces associated with a double-ended rup-ture of a main coolant pipe.
Further detail on the containment design is presented in following s ec ti ons.
1 4.2 Containment Loading l
Factored loads for the design of the containment structure have been proposed which combine dead loads, live loads, pressure loads, temperature loads and earthquake loads (or wind loads if greater than the earthquake loads). The containment will be designed such that the most restrictive I
load combination for each particular region of the containment results in l
average stresses not greater than the yield point.
The factored loads indicated that the containment will have the capacity to withstand loadings, as follows:
(a) at least 50 percent greater than those calculated for the los s-of-coolant accident,
12 (b) at least 25 percent greater than those calculated for the loss-of-coolant accident coincident with the operational basis earthquake, (c) at least as great as those calculated for the loss-of-coolant accident coincident with an earthquake twice the magnitude of the operational basis earthquake with no loss of function.
The design pressure for the containment is 47 psig.
The applicant l
I has performed a parametric study of the containment pressure as a function of time using various assumptions for the course of the accident.
The engineered safety features assumed to be operable in these studies were one of two containment spray systems and three of five containment fan-cooler units. The design basis case assumes a primary system break size of about three square feet which is calculated to result in transfer of the maximum available net energy f rom the core internals and vessel to the coolant and in the highest peak containment pressure. This design basi'.
-l accident results in a peak containment pressure of 39.6 psig.
Analyses were also performed by the applicant te demonstrate the adequacy of the containment design pressure by including additional energy transferred j
I' from stored energy in the components and from reactor decay heat, and by includ ing metal-water reactions. Inclusion of a metal-water reaction which involves about 32 percent of the Zircaloy within 1000 seconds causes a peak l
containment pressure of 45,7 psig, which is less than the design pressure.
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Earthquake loadings will be computed on the basis that the vertical acceleration values are two-thirds of the horizontal ground acceleration values with the effects of the h orizontal and vertical loadings combined j
l on the assumption that they act simultaneously.
The operational basis earthquake and the design basis earthquake are defined in terms of accelera-
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i tion in Section 2 of this report.
Our consultants have reviewed the seismic aspects of the containment design and their conclusions, with which we concur, are included in Appendix F to this report.
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4.3 Structural Design Details i
The containment reinforcing consists of hoop bars and helical bars J
l extending the full height of the wall at the exterior face, and continuing
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i over the dome to meet the reinforcing on the other side.
The diagonal re-I I
inforcing at the inside face of the wall reaches an elevation of approximately I
83'-0".
This arrangement permits the main reinforcing at the exterior face to be placed with no need for end anchorage points in zones of biaxial ten-sion in the structure.
The applicant's criteria for splice stagger and user testing of rebar are adequate.
The ASTM A615 GR60 reinforcing bars will be limited in carbon and manganese content te assure better fusion welding and bending properties, although cadweld spidces will be used where possible.
The applicant has agreed to utilize cadweld splice testing criteria which conform to our requirements.
A hinge design at the base of the containment wall structure permits the wall to rotate outward at the base while still preventing radial expan-sion of the wall at the base. We and our dynamic design consultants have reviewed the hinge concept and design criteria and find them to be acceptable.
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i The 18.5 foot diameter equipment hatch and the 9.0 foot diameter personnel hatch are framed by 2-1/2 inch thick A-36 structural steel rings which transfer the loads around the openings.
The ASNE Boiler and Pressure Vessel Code Section III will be followed in design, fabrica-l tion, and heat treating of the rings. The applicant has stated that proper consideration will be given during design to the f act that the ring will not be planar. Connection of reinforcing to the ring will be through I
a rod with a Cadweld sleeve on the rebar end and a threaded connection at the steel ring. We find the desigt. criteria for these rings acceptable.
The steel liner for the containment will conform to f.STM A516, l
Grade 70, and will be fabricated in accordance with the ASME Boiler and Pressure Vessel Code, Sectivn VIII and IX. Since the design criteria for the l
liner result in a factor of safety against buckling of 1.17, we find the i
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design approach acceptable.
1 The dynamic analysis methods and assumptions used in evaluating the l
design of the containment have been reviewed by us and our dynamic design consultants and found to be acceptable.
.4 Testing and Surveillance All liner plate welds will be covered by channels, strength tested at 54 psig for fifteen minutes, then leak tested at 47 psig.
The acceptable leak rate will be no greater than 0.1% of the containment free volume per day.
The wall and dome liner plate welds will be radiographer for the first ten l
1.5 feet made by each welder for each position, and then at least ten per-cent of the weli. thereafter will be radiographer.
The containment structural proof tests will be conducted at 35,
- 0, 47, 50, and 54 psig pressure.
The liner strains will be neasured by means of electri-e 1 etrain gauges. Radial and longitudinal growth of the cylinder will be maasured through visual targets on the structure. Radial growth at the base will be monitored by linear motion transducers. Visual checks on specified areas will be carried out to verify crack patterns and sizes.
If major maintenance or modifications of the containment building are made, it will be possible to pressure test it to 54 psig. The applicant has stated that to perform this on Unit 2, an evaluat.1 would have to be made of the necessity to close down Unit 1 and evacuate personnel during the test, but if a 54 psig test should be required in the future, it would be possible i
to conduct it.
Periodic testing of the containment as proposed by the applicant will con-sist of pressurizing the double-walled penetrations and weld seam channel.s without an integrated leakage test. We will require the periodic performance of an integrated leak test at a presr2re and frequency to be developed at the operating license review stage.
Based on our review we conclude that the applicant's proposed testing and surveillance prograns are accepttble for the construction permit stage of re-view. Requirements for periodic integrated leakage testing of the containment will be considered further during the operating license stage of our review.
4.5 conclusion We conclude that the Diablo Canyon Nuclear Unit 2 containment and Class I stzuctures can be adequately designed, constructed, tested and utilized under the criteria presented by the applicant, and are acceptable.
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i 5... Cr.I'1EEPCJ SAT"JTY TEATt'i'IS 3.-
Ererrency Core _ Coolinr Systens
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The ererrencj co-
'olina svt.ter (r.CCS) for thf = -lant !s on i~aroved 1
version of those previouri f revieued for other 4-loo 'Jesttnckmuse ""s, It consists of (a) part of tbc reactor charrine nvnte, (b) n hiah nressure i
1 coolant iniection and recirculation euhsyster (HPS), (c) tun lou pressure coolant I
l injection and recirculation subsysters (LDS), and (d) an accumulator auhryster.
The mininur ECCS capability (assured in desirn basis accident evaluations) is 3 accumulators, one lov pressure ounp, one hich pressure nur, and one centrifugal charrinc punp, all deliverino at rated ca acity snd povered by the diesel generators.
For lonr term coolant recirculation, the ninfrun ECCS capability can be provided even in the event of a failure of any single active or passive component.
All of the ECC subsysters can accomplish their functions when operatinF on nornal or everrency (oncite) pouer.
If one of three diesels fails tn start followina a LOCA and los9-o f-o f f site cover, the -int u~ ECCS vill be providinr coolant (assurinr no further comnonent failures) within 30 seconds following a sa fety inj ection sicnal.
The ap: 11 cant presented ECCS performance analyses based on corputer codes developed by Westinghouse. For larpe and intermediate sized breaks the per-formance of the rinimur ECCS is identical to that of ether four loon,1000
' t.se cla s s "'.?F 's. The calculated peak terperatutes are below the rente for accelerated Zircaloy-cater reactions (i.e., less than 2200* F) and the total I
l clad-water reaction for any break size is much less than one percent o f the l
l total fuel clad nass, Furthurnore, only a samil percent ( < l.3%) of the I
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l 17 total clad mass ever exceeds'1800* r and then only for a brief time (<f 30 sec.).
We have reviewed Arronne National Laboratory as well ar Westinchoune regarding poter tial therr.a1 shock shatterine o' oxidized Zircaloy claddine, and conclude that the core heat transfer peometry should not be significantly I
altered by therral shock upon quenchinr because the time at elevated temperatures-would be so short.
On the basis of our review, ve conclude that the Diablo Canyon Unit 2 ECCS will (a) limit the peak clad temperature to less than the reitinr, temperature, (b) limit the fuel cled-rater reaction to less than one percent of the toal clad mass, (c) tereinate the clad ternerature transient before the geometry necessary for core cooline is lost and before the clad is so embrittled as to fail upon quenching, (d) reduce the core turperature and then maintain core and coolant temperature levels in the sub-co.oled condition until accident recovery operations can be accomplished, and (e) provide rapid boron iniection to minirize the power transient for a steam Idne break and to preclude a return to power from zero power for the ster.m line break equivalent to opening of a safety valve. The ECCS will arovide this protection for all pipe breaks up to and including the double-ended ruptitre of the largest reactor coolant pipe.
5.2 containment Spray Syster The applicant has provided a containment spray systen containinc redundant active components whose function is to spray cool water into the containment atmosphere in the event of a loss-of-coolant accident in order to renove l
I heat. The system is designed to Class I seismic requirements with the vessel meeting ASME Section III requirements and the piping meeting I!SAS B31.1 l
18 requirements. All' of the active system components are located outside of containment; the spray headers inside containment are protected from missiles originating within the shield.
The system includes the caeability for adding NaOH solution to the spray in order to effect iodine removal from the containment atmosphere.
The elemental iodine removal characteristics of NaSH are well established and its compatibility characteristics with various materials have been extensively investigated. The applicant is presently modifying the NaOH '
addition systen in order to provide the operator with additional Information concerning its status during operation. This modification sdll be reviewed 1
at the operating license stage.
Based on our review of the containment spray system, we conclude that 3
the design is acceptable.
5.3 containment Isolation Systen The containment isolation syster is intended to close the vr.rious piping systems which penetrate the containment to ereclude the escape of radioactivity i
to the outside environment in the event of a loss-ef-coolant accident. The isolation system provides a minimum of two barriers between the outside environ-ment and the containment atmosphere, the reactor coolant system, or any closed systems inside the contain ent which may be vulnerable to the accident forces.
All valves and equipment which are intended to be isolation barriers are protected against potential missiles and water jets and are designed to Class I seismic require.ments.
For those lines which must be isolated immediately following an accident, automatic trip signals to the valves are provided; no reliance is placed on manual operation. Other limes which must remain in service subsequent to an accident for safety reasons are provided with at least one ramote-manual valve.
19 Isolation valves are automatically tripped to their closed position by one of two separate containment isolation strnals. The najority of the valves which are in the "non-essentini" process lines are tripped in conjunction uith autoratic actuation of the safety injection myster. These are pipelines whose isolation vill not increase the potential for damage of containment equip-ment. The renaining valves, located in " essential" process lines, are tripoed upon actuation of the containment spray system. The essential lines (i.e.,
those providing cooling water and seal water flow to the reactor coolant pumps) are not interrupted unless absolutely necessary. Both the automatic isolation valves and renote-manual valves are operable from the control room area. Position indication for each valve is also provided on the control room panels.
The isolation valves are designed to close upon loss of control power or air. Also, the instrumentation and control circuits are redundant in that a sinrlc f ailure vill not nrevent containment isolation. In addition, provisions are included to perrit periodic testing of the leak tightness and functioning of the isolation valves.
We have revieued the design of the containment isolation systen and the applicable design criteria and have found then to be acceptable.
I 5.4 Hydrogen Control In Amendnent 3 to the PSAP., the applicant discussed the potential for accumulation of hydrogen in the containment following a loss-of-coolant accident and a proposed method for limiting the concentration to preclude an explosion hazard ' The various radiolytic and chenical mechanisms for hydrogen generation were considered and the sopiteant concluded that J
20 the concentration of hydrogen would reach the lower flarcability limit j
in the containment in approximately 130 days following the accident. At this time, the applicant would propose to vent the containment nases to the e
atmosphere through filters and a charcoal bed.
The applicant states that by controlling the venting operation over a 30-day period, offsite doses from I-131 and Kr-85 could be maintained below the limits of 10 CFF 20, if averaged over a one-year period.
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Using parameters for hydrogen production which are sornewhat more conservative than those assumed by the applicant, we calculated that the lower flammability limit would be attained in less than 40 days.
For this case, the purging operation could result in activity concentration levels offsite which exceed 10 CFR 20 limits; however, additional capability I
for filtering of containment ef fluent, including charcoal beds, vould f
reduce the I-131 concentration levels. We have calculated that, as suming 90% filter efficiency for iodine removal, doses at the site boundary would be about 0.8 rem whole body and about 8.5 ren to the thyroid if the entire contents of the containment were vented over a 30-day period.
I We are continuing our evaluation of the potential for hydrogen accumulation in large water-cooled power reactors and of systems for post-
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accident hydrogen control. On the basis of our review to date, we have
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concluded that the venting approach is technically feasible and will limit off site doses. However, continued ef fort should be pursued in developing means in addition to containment venting, such as flame and catalytic recombiners, hot surface combustion or the use of chemical additives in 1
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the spray solution for scavenging hydrogen. We vill continue to work with the applicant on this problem area during construction.
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21 3.0 DESIGN OF CLASS I STRUCTURES FOR SEISFIC AND ACCISENT LOADINGS 6.1 General Those structures, syctems and components of the nuclear plant which are important to nuclear safety, i.e.,
failure of which might cause or increase the severity of a loss-of-coolant accident or result in the release of excessive amounts of radioactivity, are termed Class I.
Those components which are not essential to the safe shetdown of the reactor and failure of which would not result in the release of substantial amounts of radioactivity are considered Class II. Those structures and components not related to reactor operation are Class III. The subsecuent discussion on design criteria contained in thin section relates to the Class I structures and components.
6.2 Earthquake Response Spectra and Damping The magnitude of ground acceleration for earthquakes at this site, including the revie" of this aspect by the 1!. S. Pear t and Geode tic Survey for Unit 1, was discussed previously in Section 2 The resnonse spectra for the 0.15 g nround acceleration earthquake ("far away") and that for the 0.20 g cround acceleration earthquake ("close-by") were presented in the application. Together these response snectra represent the operational basis earthe,uake. A larger 0.40 c_ r.round acceleration earthquake response spectrum is also contained in the appif cation, uhich represents the design basis earthquake.
The dampinr values to be umed in the desian are presented in the PSAD.,
There values compare favorably nith values used in the desirn of other facilities.
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During detai'sd design, vibrational response characteristics will be calculated for the assumed earthquakes. We agree 'dth the ACRS that consideration should be civen to obtainine exaerirental verification of the anticipated behavior of corronente and instrumentation to the extent prac ticable.
6.3 Desirn Loadinf_ and 3 tress-Strain Linits This section pertains to the design loadine and limits for Class I structures, systems and components including the pressure vessel and piping, but excluding the containment structure.
(The desirn loadinr,and stress limitations for the containment structure vere de=cribed in Section 4.2).
The requirement for these iters is that they should be designed to withstandt (a) Load combinations includine nornal design leads and operational basis earthquake loads within norral verking stress or deflection 11 nits.
(b) Load combinations includin2 desien basis eart'. quake loads and applicable design basis accident loads, uit'..eut less of function I
o f the specific structure, syster, or conponent.
We have reviewed these requirements and re consider trese loading combinations and lirits to be both realistic and satisfactory. The proposed stress or deformation linits for specific corponents are dis-cussed in nore detail belou.
Reacto_r_ 'lessel Internals For norcal dasirn loads of ecchanical, hydraulic, and thernal origin, including anticipated plant transients and the operational basis earthquake, the reactor internals will be designed to the stress limit criteria of the ASFE Boiler and Pressure Vessel Code Section III.
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23 The reactor internals will also be designed to withstand the con-current loads resultine from the loss of-coolant accident and the design
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basis carthruake. The direct primary stresnes under this combined loadine i
will net exceed stresses corresponding to 20 percene of uniform strain at I
temperature. Allovable deflections will be limited to chout 50 percent of the design loss-of-function deflections for the specific components.
Preliminary calculated deflections are smaller than even these limits.
1 We find that the strees and deflection limits discussed above provide an adequate margin of safety.
Vessels, Pipint and Supports The AS:T Section III and B31.1 Code stress linits for vessels, piping and supports as proposed by the applicant are considered satis-l factory for type (a) load combinations.
l ror the type (b) load combination, the allowable extent of plastic I
I deformation can be larger than that associated with the AS'T Section III 1
stress limits. Further, the limite on strain defined in Appendix F of the pSAR will assure no loss of function. We conclude that the design provides an adequate margin of safety.
6.4 Conclusion On the basis of the information presented in the application and the analyses of the design by our consultants Newnark and Hall as attached as Appendix F, we have concluded that the proposed design criteria for Class I structures systers and components provide an adequate nargin of safety to withstand the selected combined loading conditions.
24 7.0 INSERVICE INSPECTABILITY AND SURVEILIANc3 In Amendment 1 to the PSAP, the applicant has indicated that the design of the reactor coolant system and its components mskes provision for virtually complete compliance with the draf t ASME Code for Inservice Inspection of Nuclear Reactor Coolant Systens dated October 1968. The applicant is currently formulating plans for inservice inspection and will submit them for our review prior to the application for an operating license. The plano are intended to reflect the requirements of the latest accepted revision of the inservice inspection code to the extent practical.
Visual and/or non-destructive inspection techniques can he applied on a conventional basis to all primary s>atem components (including the reactor coolant pump flywheel) with the exception of the reactor vessel because of high radiation levels and remote underwater inaccessibility.
In some areas, the inspection cust be limited to external or internal surfaces or cannot be performed in shield penetration areas because of accessibility difficulties. For the reactor vessel, however, volumetric inspections must await the availability of non-destructive testing techniques suitable for areas of high radiation.
In anticipation of the development of these techniques, the applicant will perform ultrasonic test mapping of selected areas of the vessel during fabrication.
On the basis of our review of the applicant's approach to inservice inspection, we conclude that an acceptable program will be developed.
We will review the applicant's program as it is developed during con-struction.
23 The applicane. has included -revisions 'or n survaill.ance cro~rar to ronitor the extent of radiatior dansee to "htch tha ?cactor vesse] tr exno se d.
The nrorrar rovidec 'or etrht c.nsu]ae ce-taiainr n -iriwim o f 351 npecinens for tensile, C':ar," Y-notch, and ree e enin~ inadi-- enetinr.
A tentative schedule for cansule renoval durina the a crationel Idfe o f the niant is provided. Tbc saecirens are located se that they experience hieber irradiation than the vessel vall thereby nroducine data representative of the vessel at a later tine in if fe.
3ased upon our review, we conclude that the applicant's pronosed prorran is acceptable.
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8.0 ELECTRIC POWER SYSTEMS _
Offsite Power Diablo Canyon Units 1 and 2 vill be interconnected to the PG&E systems through 500 kV and 230 kV circuits. Power from both Units 1 and 2 feed the 500 kV switchyard.
Three transmission lines emanate from the 500 kV switchyard on two separate rights-of-way. The 230 kV switchyard interconneeds the plant to the PG&E system by means of two circuits, each on a separate right-of-way.
Redundant independent sources of offsite power provide power to the 4160 volt buses which feed the engineered safety features upon loss of the j
normal unit supply. One source is derived from the standby-startup trans-former connected to the 230 kV switchyard.
The second source is made available by backfeeding through the main transformer from the 500 kV switchyard. This is made possible by a manually initiated motor-operated disconnecting ' ink in the generator's main leads. This disconnect link i
is operated from the control room and is interlocked to prevent opening l
l under load.
The applicant states that this source of power could be made l
available in approximately 30 seconds.
We conclude that sufficient redundant and independent sources of offaite power are provided to reasonably ensure that no single failure will cause the loss of offsite power to the engineered safety features.
Onsite Power The engineered safety feature loads are divided among three 4160 volt buses such that operation of any two will supply the minimum safety requirements. Three 2500 kW diesel generators are provided. Two are
27 exclusively assigned to two of the above mentioned buses, and the third diesel generator is capable of being connected to either the third bus of Unit 2 or to a similar bus in Unit 1.
The fuel supply is suf ficient to operate the diesels required for accident loads in one unit and safety shutdown loads in the second unit for a period of seven days. The applicant has made provision fer quick resupply from nearby sources.
Two d-c systems are provided for control and protection instrumentation, annunciation, and emergency lighting and lubricating systems.
One is a 250/125 volt three-wire ungrounded system with two 125 volt buses, and one 250 volt bus for turbine generator emergency motors. The other system is a separate 125 volt two-wire ungrounded system supplying a 125 volt bu..
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Each of these buses has its own battery and its own redundant battery chargers.
Sufficient battery charger capacity is provided to carry the normal continuous load and to recharge the batteries with any one charger out of service. The batteries are sized fcr two-hour operation of all vital loads.
Each battery is located in a separate room in a Seismic Class I area.
The losss of any one battery will not preclude the operation of the minimum required engineered safety features.
The 120 volt a-c vital instrument bus system supplies power for con-trol and protection system instrumentation and alarms. These loads are divided into four groups. Each group is served by a separate inverter supplied from the station batteries. A normally open standby connection between the inverter fed buses is available for use when an inverter-is out of service.
Based on our review, we conclude that all relevant aspects of the onsite power system comply with our criteria, and that the system design is acceptable.
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.1 9.0 RADI0 ACTIVE WASTE CONTROL Liquid wastes are collected during plant operation and pumped to the l
vaste holdup tanks for batch processing by the Waste Disposal System.
I Liquid processing will include filtration and evaporation as necessary.
Liquid releases to the condenser discharge tunnel will be in conformance with 10 CFR Part 20 limits and will be continuously monitored by a radiation i
monitor. High radiation from the monitors sd11 automatically close a dis-charge valve in the liquid waste disposal system. Prior to release of any batch of liquid, samples will be analyzed to determine the type and j
i amount of activity in the batch to ensure conformance with release limits j
l which will be established at the operating license review stage.
j Solid radioactive wastes from plant operation will be temporarily stored on site and shipped from the site in containers approved for that j
purpose.
Gaseous radioactive wastes from the chemical and volume control system, various cover gases and venta will be collected and compressed into Gas Decay Tanks. Af ter a suitable decay period, the contents of a tank will be sampled to determine its activity and released to the vent pipe. A radiation monitor is also provided in this discharge line and, should the radiation level become high during the release, will automatically close a valve in the discharge line from these tanks.
The limits for gaseous release in accordance with 10 CFR Part 20 vill be established at the operating license review stage.
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29 We conclude that the waste disposal system proposed by the applicant will effectively control the discharge of radioactive waste generated on the site in xcordance with AEC regulations and that it is acceptable.
10.0 ACCIDENT ANALYSIS 10.1 Operating Transients J
A nuinber of plant operating transients were considered by the appli-cant including rod withdrawal during startup and from power, boron dilution, loss-of-coolant flow, loss of electrical load, and loss of offsite AC power, in order to assess the safety rmrgins of the plant design. The criterion for detailed design of the reactor control and protection system is to be able to automatically take corrective action to copy with any of these transients. Preliminary analyses as presented in the application will be recalculated during detailed plant design to verify that transients are within the capabilities of the reactor control and protection systems.
Based on our evaluation of the information submitted and on the results of evaluations of other PWR designs at the operating license stage, we con-clude that anticipated transients can be terminated with adequate margin, j
10.2 Accident Evalu at ion Potential accidents which could result in radioactive releases to the environment have been analyzed by the applicant. We have evaluated these accidents and the engineered safety features provided to limit the potential e xposures. During the course of the detailed design the applicant will continue to reassess the potential consequences of the various accidents, 4
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and we will evaluate the final design at the operating review stage. We have evaluated the potential consequences of these accidents and all of
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i the resulting doses, with the exception of the postulated fuel handling accident, are well below the 10 CFR 100 guideline dose levels at the available exclusion zone radius (0.5 mile) and the low population zone
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radius (6 miles).
The more significant of these accidents are discussed in greater detail in the following sections.
10.2.1 Loss-of-Coolant Accident i
The design basis loss-of-coolant accident for Diablo Canyon Unit 2 is similar to that used for other PWR's in that a double-ended break in 2
the largest pipe (9.2 ft ) in the reactor coolant system was assumed. As discussed in Section 5.1, the emergency core cooling systems are designed to limit fuel cladding temperatures to well below melting temperature.
Although the basis for sizing the emergency core cooling system is to limit fission product release from the fuel, we nevertheless require that the containment and its associated engineered safety features shall be capable of limiting potential doses in conformance to 10 CFR Part 100 guidelines assuming sigM ficant releases of fission products from the i
fuel. We analyzed the consequences of this accident assuming that 100%
of the core noble gases, 25% of the core iodines and 1% of the core solids are released to the primary containment atmosphere. Based on our evalua-tion of the containment spray system, we assumed an iodine removal time constant of 3.7 per hour for the sodium hydroxide sprays. We also assumed that 10% of the iodine would be in a chemical form not removable by the spray system and that the containment leaked at the design basis rate
31 of 0.17. per day for the first day and one half : hat value for the remaining 29 days.
1 We assumed a ground release using Pasquill Type T meteorology and 1 m/see wind speed for the first 8-hour period af ter an accident. The same meteorological conditions with a uniform dispersion into a 22-1/2 sector were used for the 8 to 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period an:! the stability, wind speed, I
and wind direction were varied for the 1 day to 30 day period. We l
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l calculsted potential doses for the loss-of-coolant accident at the l
l exclusion distance for a two-hour period of 4 rem and 190 rem for the whole body and thyroid, respectively.
The corresponding 30-day potential l
doses at the low population zone boundary were less than 1 rem whole i
l body and 26 rem to the thyroid.
10.2.2 Fuel Handling Accident j
In our analysis of this accident we assumed that all of the fuel rods in a dropped fuel assembly (204) were damaged, that 107. of the iodines l
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and 207. of the noble gases contained in the rods were released to the i
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refueling water, that 107. of the iodines and all of the noble gases that j
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were released to the water were released to the building atmosphere, that the accident occurred 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> af ter shutdown of the reactor, and that
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the dropped fuel assembly had previously operated at a power density 807.
j higher than the average core fuel assembly.
Using meteorological assumptions i
l of Pasquill Type F with a wind speed of 1 m/sec, we calculated potential j
i doses at the exclusion radius of 4 rem whole body and 1575 rem to the l
i thyroid.
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32 The applicent has calculated a thyroid dose of 0.3 rem as a result of this accident using less conservative assumptions than those presented above. The differences between the applicant's calculated value and ours results from the different values assumed for iodine inventory in the rods which are released to the water; the fraction of iodines retained by the pool water; and the number of rods damaged as a result of the dropped fuel assembly.
We have discussed the above differences with the applicant and in Amendment 5 he indicated that he plans to conduct further experimental and analytical studies of the problem in order to justify his less conservative assumptions for the accident analysis.
The applicant has further stated that, in the event his study results are not accep:able, he will provide additional iodine fixing equipment in the Fuel Handling Building ventila-tion system. We have concluded that the inclusion of such equipment could provide the dose reduction factor necessary to make the consequences of the fuel handling accident well within 10 CFR 100 guidelines.
The applicant's results and the possible need for additional iodine fixing equipment will be evaluated during the operating license review.
10.2.3 Steam Line Break Accident l
In our evaluation of this accident we assume that a break occurs in l
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the steam piping between the steam generator and the turbine, resulting f
in the loss of all secondary water in one steam generator. The break is assumed to occur at zero power when the steam pressure is the maximum.
i 33 The resultant contraction of reactor-coolant system water is characteris-tic of.the beginning of a loss-of-coolant accident and, for large steam line breaks, causes actuation of the ECCS to limit the transient by boron injection.
For the calculation of potential doses, we assume that the break occurs while a primary to secondary leak of 10 gpm exis ts, the reactor coolant contains equilibrium activity corresponding to 17. fuel failures, and a continuous steam generator blowdown of 10 gpe is occurring.
It is further assumed that the entire iodine inventory in the secondary side of the steam generators is released.
The resultant calculated 2-hour thyroid dose at the exclusion radius is 39 rem, well within the guidelines of 10 CFR 100.
10.2.4 Steam Generator Tube Rupture We have evaluated the potential consequences of a postulated double-ended rupture of a steam generator tube. We assumed that reactor coolant, amounting to 80,000 lbs. an/. containing equilibrium activity due to 1%
failed fuel, would be released directly to the environment through the secondary safety valve over a five hour period. This release would result from a simultaneous loss-of-offsite power permitting no bypass to flow to the main condenser.
The leakage of reactor coolant is assumed to persist until the system pressure is reduced to below 1100 psia, permitting isola-tion of the faulty steam generator; in practice this should occur in less than one-half hour. No core damage is anticipated for this accident which
34 results in actuation of the safety injection systen should no operator action occur. On the basis of the above assumptions, we calculated a thyroid dose of 35 rem at the exclusion radius - well within the 10 CFR 100 guidelines.
11.0 QUALITY ASSURANCE In Amendments 4 and 5 to the PSAR, the applicant has described the Quality Assurance Program (QAP) which it proposes to utilize for the design and construction of the Diablo Canyon Unit 2 reactor plant.
The applicant states that its QAP is intended to be responsive to the recently published l
AEC quality assurance criteria.
PG&E performs its own architect-engineering. The only major contractor is Westinghouse, the supplier of the nuclear steam supply system. Within the applicant's engineering department a new group has been established separate from the design group with responsibility for a continuing review of the QAP and for reporting on its adequacy to PG&E management.
This new group is headed by the Director, Quality Engineering, who is directly responsible, as is the Project Engineer for Unit 2, to the Vice-President of Engineering.
This arrangement is intended to provide a quality assurance activity which is independent of the design and construction activities. The Director, Quality Engineering will have a staff of engineers and specialists who will au lit quality related activities during design and cons truction.
This organizational arrangement satisfies our requirements.
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In his description of the QAP, the applicant has presented its approach 'to each of the QA criteria specified -in 10 CFR 50 Appendix B I
(Proposed).
The program requires that planned and documented actions be applied to all quality related activities which involve the structures, 1
systems and components important to safety. All contractors, including Westinghouse, will be required to provide a QAP to the extent necessary as determined by their scope of work.
Surveillance of contractor's ef forts will be performed by PG&E.
Although many aspects of the applicant's QAP will require further definition, the applicant's commitments and its planned approach in each of the critical areas satisfy the criteria for the construction permit I
stage. During construction, we will follow the development of the details of the applicant's QAP to assure adequate implementation of j
1 current plans.
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16 12.0 RESEARCil AND DEVELOP!T4T PRO'i' LAMS The applicant has identified several aream in t'e deci~n o f tbc nlant j
requiring further rencarch and develovent ef fort. T eme arann include those in which additionni information is reantred for,lant o,eration and those which will provide added confir nation that the contemplated desinns are sufficiently conservative. "one o f the research and develonnent areas can be characterized as unique to Diablo Canyon Unit 2; rather, they are common to all Westinghouse PWR's and the results of these ef forts as they are perforced for preceding N's will be applicable to Unit 2 Each of the research and development areas is briefly surrarized below:
12.1 Power Distribution Control It is necessary to demonstrate the canability of out-of-core detectors to indicate axial and diaretral core nower distributer. to permit control of xenon oscillation and to warn of nisplaced control rods, to develon a control system for axial power shapine with part-lencth rods and for j
diametral power shaping if necessary with full len-th rods, and to verif" durinc start-up testing that the control syster satis'actorily provides no'aer distribution control and adequate safety r.arrins. ' e do not arree uith the applicant that the initial t'ro objectives have been adequately accor.plished.
We are not convinced that an adequate capability for determining, the nuclear status within larac cores such as Unit 2 has yet been demonstrated by out-o f-core detectors. The applicant has made provision for the use of incora detectors should they become necessary.
37 Westinghouse expects to fulfill these objectives durine startup testing and subsequent operation of the Ginna and Indian Point Unit 2 plants scheduled for 1969 and 1970 respectively.
Tinal verification of the ziequacy of these control systems in la'er units *111 be obtained durinc startun testine prior to full power operation.
12.2 Fuel Rod Burst Ef forts are underway to determine the extent of core Reor.etry distortion which may result from a loss-of-coolant accident. This verk is being performed by various industry and covernrent croupe with overall coordination by the Oak Eidge National Laborator'r.
\\ny naior uncertainties rerardinr the extent of core geonetry distortion and the capability for adequate ECCS parforr.ance should be resolved prior to operation o' niablo Unit 2
.3 Failed Fuel hbnitor The experimental evaluation of several detector tynes by Uestinchouse is underray at the Saxton reactor rith the obj ective of developina a unit which will provide prorot detection of failed fuel pins. Detailed evaluation of these detector t pes vill be conpleted in 1970 At the present tine, f
a nonitor in the letdo m stren". is included in the Unit 2 desian.
Determination of the adequacy of the various detector types for use in the riablo Canyon Units will be made during the o, erat n-licenne revien.
d 12.4 Burnable Poison Experimental and analytical efforts to develop burnable poison rods have been satisfactorily completed.
'/crificetion o' the.'esi~n aAecuacy will be availa')1e fror cornercial operation of the Cinna plant prior to I
operation of Diablo Unit 2 m_.m.
I 38 12.5 Fuel Development Fuci development efforts' are underway in-both the Saxton and Zorita irradiation programs to deternine performance rarsins available relative to peak pot:cr and the accommodation of overpe'er transients. Results from these efforts will becone available beginninp in late 1971 and extendinn
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through the end of 1973, prior to operation of Diablo Canyon Unit 2 12.6 In-Core Detector Evaluation of the lifetine of in-core detectors suitable for continuous monitoring of power distributions is proceeding in several reactor facilities such as San Onofre, Brookhaven and Western New York 7uclear Research Center.
Evaluations will be completed in 1971.
12.7 Rod Bundle DNB Experimental rod bundle departure from nucleate boiling data with non-uniform rod axial flux distributions including the effects of mixing vanes will be obtained at Colunbia University under the direction of Westinghouse.
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Preliminary results indicate that the mixing vane provides a greater DNB heat flux than predicted.
1 12.8 Full Length Emerpeney Cooling Heat Transfer Tests Experimental investigations are being conducted to determine more precisely the thermal behavior of a FWR core following a loss-of-coolant accident.
Of particular concern is the experimental verification of the performance capabilities of proposed emergency core cooling techniques. This work, initiated in 1968, is being done by Westinghouse under contract to the AEC.
12 9. Flashing Heat Transfer Experimental investigations of the heat transfer behavior in-the core during various phases of the loss-of-coolant accident are underway.. The
39 objective is to demonstrate that the heat transfer model assumed for the core during blowdown, uncovery, and reflooding is conservative. This work is being done by L'estino, house and the University of "ichigan. The current I
najor effort is reduction of test data.
12.10 Loss-of-Coolant Analysis An analytical study is underway to integrate the results of the experimental programs in order to obtain a more realistic core thermal design code. Preliminary results of this effort, performed by Westinghouse, show that for the Ginna reactor peak clad temperatures of 300* to 400' F less than prior predictions are obtained for large and intermediate size I
breaks. A similar analysis will be performed for Diablo Canyon Unit 2 and
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i presented in the FS AR for our review.
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1.11 Blowdown Forces A computer program has been developed to determine pressure, velocity, and force transients in order to calculate bloudott forces on reactor vessel internals. The program, BLODWN-2, has been applied to the Indian Point Unit 2 reactor uith the conclusion that the calculated performance meets the design criteria.
Specific analyses for niablo Canyon Unit 2 remain to be performed and will be presented for our review in the FSAR.
12.12 Reactor Vessel Thermal Shock Analysis Preliminary analyses have been conducted which indicate that thermal shock effects resulting from ECCS operation in the event of a 1.oss-of coolant accident would not be expected to cause reactor vessel failure, at least during the early years of vessel life.
Further evaluations will be performed L
40 based on analytical results and material property data being obtained from the Heavy Section Steel Technology Program at tak P.idre National Labo rato ry.
1 Based on our review of the.above research and development efforts, we l
have concluded that these programs should provide the required design information and assurance of design adequacy for the various areas of concern in the Diablo Canyon Unit 2 reactor plant prior to cperation.
The results of these programs as applied to Unit 2 vill be evaluated during the operating license review; and as further information is accumulated.
13.0 TECHNICAL QUALIFICATIONS The applicant has been active in the design, construction and operation of nuclear power plants for approximately eleven years. This experience includes the operation of the Humboldt Bay reactor and the design and construction of Diablo Canyon Unit 1.
On several occasions during this time, we have reviewed the structure, responsibility and operating philosophy of the applicant's engineering organization. The Nuclear Steam Systen Supplier, Westinghouse Electric Corporation, has designed and constructed a number of pressurized water reactors which have been approved by the Commission. We conclude, based on these reviews, that PG&E and its contractor t.re technically competent to design and construct Diablo Canyon Unit 2.
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l 41 14.0 CONFORMANCE TO THE GEN _ERAL DESIGN CRITERIA In November 1965, the Commission published its General Design Criteria for Nuclear Power Plant Construction Permits, and on July 11, 1967, published in the Federal Register its revised General Design Criteria taking into
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I account comments received on the initial criteria and further development J
of the criteria by the regulatory staff.
The applicant has cross-referenced the information as presented in the application with the criteria. We have evaluated the applicant and have concluded that the proposed unit conforms I
to the revised criteria. We will review the proposed unit at the I
operating license stage in light of the criteria as formulated at that I
time.
15.0 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFECCARDS The Advisory Committee on Reactor Safeguards, by letter dated October 16, 1969, reported on the Diablo Canyon Unit 2 Nuclear Facility.
A copy of this letter is attached as Appendix B.
The letter contains a number of comments and recommendations which are listed below and noted in appropriate sections of this evaluation.
These items will be resolved
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to our satisfaction and will be reviewed by the ACRS n'ior to the issuance of an operating license.
15.1 Boron Carbide Control Rods (See section 3.0)
The Committee noted that careful attention should be paid to the possibility of control rod swelling and related consequences.
Based on our review of the applicant's response to our concern, we have concluded that the use of boron carbide for the control rods is acceptable. We will examine this design aspect further during the operating license review and, if necessary, impose appropriate requirements in the Technical Specifications to provide assurance that incipient rod swelling is detected early.
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l 42 15.2 Fuel Handling Accident (See Section 10.2.2) l i
It was observed that the applicant intends to perform experimental and
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i analytical studies relative to the assumptions made in essuring the con-cequences of a fuel handling accident.
Additional fodine remeval equipment may be required. We will evaluate the study results during the operating license review.
15.3 In-Service Monitoring for vibra tion l
The Committee recommended development and implementation of means i
I for in-service monitoring for vibration or for the presence of loose parts in the reactor cociant system.
The applicant in Amendment 5 has indicated j
that he is evaluating the need for in-service monitoring. We will consider this aspect during the operating license review, i
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15.4 Common Failure Modes _
The Committee expressed concern about the possibility of common mode failures which could negate the action of a protection system or other l
engireered safety features. The applicant will perform further evaluations l
l of the effects of protective system failures. We will cc:,tinue our review l
of this design area during the operating license review.
15.5 Hydrogen Control (Section 5.4)
The Committee noted that the applicant has preposed a purge system to corr:rol the. potential buildup of hydrogen in the containment following a loss-of-coolant accident. We have concluded froc our studies to date that a purge-type control technique may not be considered acceptable as the primary
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means for hydrogen disposal. We have informed the applicant that an j
alternate technique should be developed. We will pursue this design area with the applicant during the construction phase.
15.6 Integrity of Vital Post-Accident Systems The Committe indicated that potential problems in vital cooling systems may arise during the post-accident period as a result of severe environments and degraded coolant. The applicant has stated he will r,"ure operability of vital systems by appropriate design and testing.
We "4?.1 evaluate the adequacy of the applicant's efforts during the operating license review.
15.7 Testability of Relay Circuitry The Committee noted that the applicant has proposed a partial con-tinuity check for the operability of the relay circuitry for engineered i
safety features. We have asked the applicant to develop a more complete tes ting technique. We will evaluate the applicant's efforts along these
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lines during plant construction.
15.8 Effects of Earthquakes on Structures and Components The Committee feels that experimental verification of the anticipated behavior of important structures and components during large earthquakes and appropriate instrumentation to observe behavior of these itons during smaller earthquakes should be considtr:4. This information would provide guidance for continued reactor operation in the event of an earthquake.
The applicant has indicated that he will consider these aspects further.
We will pursue this desita area during plant construction.
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The report of the ACRS concluded "...The Committe believes t;iat the items mentioned can be resolved during construction, and that, if due consideration is given to the foregoing, Nuclear Unit 2 proposed for the Diablo Canyon Site can be constructed with reasonable assurance that it can be operated without undue risk to the health and safety of the public."
16.0 COMMON DEFENSE AND SECURITY The applicant reflects that the activities to be conducted would be within the jurisdiction of f.he United States and that all of the directors and principal officers of the applicant are American citizens. We find nothing in the application or otherwise to suggest that the applicant is owned, controlled or dominated by an alien, a foreiga corporation or a foreign Government. The activities to be conducted do not involve any restricted data, but the applicant has agreed to safeguard any such data which might become involved in accordance with the regulations. The applicant will obtain fuel as it is needed from sources of supply available for civilian purposes, so that no diversion of special nuclear material from military purposed is involved. For these reasons and in the absence 1
of any information to the contrary, we have found that the activities to 1
be performed will not be inimical to the common defense and security.
17.0 CONCLUSION
S Based on the proposed design of the Pacific Gas and Electric Company's Diablo Canyon Unit 2 facility, on the criteria, principles and design arrangt-ments for systene and components thus far described that include all of the i
important safety items, on the calculsted potential consequences of routine i
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45 and accidental releases of radioactive materials to the environs, en the scope of the development program which will be conducted, and on the technical competence of the applicant and the principal contractors, we have concluded that, in accordance with the provisions of paragraph 50.35(a),
10 CFR Part 50 and paragraph 2.104(b) 10 CFR 2:
1.
The applicant has described the proposed design of the facility, including, the principal architectural and er.gineering criteria for the design, and has identified the major features or com-j ponents for the protection of the health and safety of the public; 2.
Such further technical or design information as may be required I
to complete the safety analysis and which can reasonably be left for later consideration, will be supplied in the final safety analysis reports; 3.
Safety features or components, which require research and develop-1 ment have been described by the applicant,and the applicant has identified, and there will be conducted, research and development programs reasonably designed to resolve any safety questions associated with such features or components; 4.
On the basis of the foregoing, there is reasonable assurance that (i) such safety questions will be satisfactorily resolved at or before the latest date stated in the application for completion of construction of the proposed facility and (ii) taking into consideration the site criteria co0tained in 10 CFR Fart 100, the i
proposed f acility can be cons tructed and operated at the proposed location without undue risk to the health and safety of the public; L------------ ----- -- ------
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5.
The applicant is technically qualified to design and construct j
the proposed facility; and 6.
The issuance of permits for the canstr.:ction of the facility will not be inimical to the comon defense and security or to the health and safety of the public.
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l APPENDIX A l
CHRONOLOGY OF THE PG&E APPLICATION REVIEW DIABLO CANYON UNIT 2 ITEM DATE COMMENTS Application for Con-7-1-68 PSAR Volume 1, 2, and s truction Permit filed 3 (Docuet No. 50-323) for Unit 2 Initial meeting with 8-13-68 Introductory session applicant Second meeting with 10-22-68 Technical discussion applicant Third meeting with 11-4-68 Technical discucsion applicant Staff question list #1 12-11-68 Leter, PAMorris to RHPe ters on Amendment #1 5-12-69 Partial response to queerion list #1 Amendment #2 5-28-69 Partial response to q
l ques tion list fl q
Amendment #3 6-23-69 Partial recponse to question list #1 Fourth meeting with 7-30-69 Technical Discussion applicant 7-31 -69 Amendment #5 9-08-69 Clarification of responses to question list #1 plus ad(itional information Amendment #6 9-25-69 Supplementary meteorological data i
ACRS Subcommittee Meeting 10-01-69 ACRS Full Committee Meeting 10-10-69 1
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