ML20236C881
Text
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May 9,1975 SFPFLDENT NO. 2
_TO THE SAFFIY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY CGIMISSION IN THE MATTER OF PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR P0'wE STATION, UNITS 1 AND 2 DOCKET NOS. 50-275 Ale 50-323
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i TABLE OF CONTENTS PAGE
1.0 INTRODUCTION
............................ ............... 1-1
- 2. 0 S ITE CHARACT ERI ST IC S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2- 1 2.3 Meteorology....................................... 2-1 2.3.3 Onsite Meteorological Measurements Program................................. 2-1 2.3.6 Co nclus ions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-3 4.0 REACT 0R................................................. 4 4.2 Mechanical Des ign . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.2.1 Fue1...................................... 4-1 4.4 Thermal and Hyd raulic Des ign. . . . . . . . . . . . . . . . . . . . . . 4-4 ,
i 6.0 E!CINEEEED S AFETY FEATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 1
6.2 containment Systems............................... 6-1 6.2.1 Contan amen t Functional Design. . . . . . . . . . . . . 6-1 6.3 Baergency Core Cooling System (E0CS) . . . . . . . . . . . . . . 6-3 ;
- 6. 3.1 Design IA5eS............................... 6-3 ;
6.3.3 Performance Evaluation. . . . . . . . . . . . . . . . . . . . 6-4 >
6.3.5 C o nc lus ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-4 f
7.0 INSTRUMENTATION AND CONTR0LS . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 !
7.2 Reactor Trip System............................... 7-1 '
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7.2.2 Reactor Trip System Actuation Logic... ... 7-1 7.2.2.1 Physical Separation.............. 7-1 l 7.2.2.2 Electrical Isolation. . . . . . . . . . . . . 7-2 {
7.2.2.4 Conclusions...................... 7-2 1 1
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race 7.3 Engineered Safety Features Actuation System....... 7-3 7.3.4 Changeover from Injection to Recirculation Mod e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 - 3 9.0 AUKILIARY SY STD(S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.2 Fuel Storage and Handling......................... 9-1 9.2.3 Fuel Handling Syctes...................... 9-1 15.0 ACC ID ENT ANALYS ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-1 15.1 Genera 1........................................... 15-1 15.2 Design Basis Accident Assump tions . . . . . . . . . . . . . . . . . 15-1 =
l 15.2.2 Fuel Handling Acciden .................... 15-1
22.0 CONCLUSION
S............................................. 22-1 l
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-1 APPENDICES PAGE 3
APPENDIX A - CONTINUATION OF THE Clut0NOLOGY OF THE A-1 R/DIO!hGICAL REVIEW................................ l APPENDIX B - REPORT OF THE NATIONAL OCEANIC AND B-1 ATMOSPHERIC ADMINISTRATION, Dated April 10, 1975...
2 APPEND 1X C - ERRATA TO THE SAFLTY EVALUATION REPORT AND SU PP LEM ENT NO . l . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-1 ,
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TABLES f 9
TABLE 4.1 - IRRADIATION SCHEDULE OF PRECHARACTERIZED FUEL p,'
ASSEMBL1ES FOR THE GENIRTC 17 x 17 SU RV E ILLANCE P ,10 GRAM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-8 j
TABLE 15.1 - POTENTIAL OF'/ SITE DOSES DUE TO FUEL CASK DR0P' ACCIDENT..................................... 15-3 ()
i TABLE 15.2 - FUEL C\SE DROP ACCIDEKI CALCULATION INPUT -
PARAMETERS........................................ 15-4 4
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1-1 < 4 1.0 INTRCOUCTION h e Commission's Safety Evaluation Report (SER) in the matter i
of the Diablo Canyon Nuclear Power Station, Units 1 and 2 was issued on October 16, 1974. In the SER it was stated that supplemental reports would be issued to update the SER in those areas where the staf f's evaluations had not been completed. Supplement No. I to the SER, issued on January 31, 1975, documented resolution of several y S
outstanding SER items, and summarised the status of the remaining 2' outstanding items.
l The purpose of this Supplement No. 2 is to further update the 7 M
SER by providing the staff's evaluation of certain matters which M\
were not resolved when Supplement No. I was issued. Each of the S ff j
following sections of this supplement is aus6ered the same as the 3
sections of the SER that are being updated.
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Appendix A of this supplement is a continuation of the chronology j1 e
of the NRC staf f's principal actions with rcspect to radiological j ai fl natters related to the processing of the application. Appendix B is
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the reprt of our meteorology consultant, the National Oceanic and
' y'I Atmospheric Administration (NOAA). Appendiz C is a listing of
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errata to the SER and Supplement No. 1.
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3-1 2.0 SITE CHARACTERISTICS 2.3 Meteorology 2.3.3 Onsite Meteorological Measurements Program In the SER, we stated that the applicant would need to provide more detailed information on control room monitoring of meteorological-parameters before our evaluation could be completed.
I In Amendment 24, the applicant described an operational onsite meteorological program. It includes displays of thirty minute mean values of: wind speed er. direction at the 25-ft and 250-fc levels; vertical temperature gradients between the 25-ft and 250-ft levels and between the 25-ft and 150-ft levels; standard deviations of vertical-and azimuthal angle fluctuations at the 25-f t and 250-f t levels; and Jewpoint temperature at the 25-ft level. The thirty minute mean j
values will be updated at five minute intervals. The control roon a display also includes two sets of values for centerline relative concentration (X/Q) and horizontal standard dei..ation of the plume ,
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(sigma-y). These values are displayed for distances of 0.8,1.0, l 2.0, 4.0, 6.0, 8.0 and 10.0 kilometers. One set of X/Q values is based upon atmospheric stability defined by vertical temperature -
gradient. The other set is based on atmospheric stability defined by azimuthal and vertical angle fluctuations. For accident con-d:
j ditions the moat conservative set of relative concentration values
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gy will be assumed valid. y CN i y l 14
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The determination of horizontal and vertical standard deviations of the plume (sigma-y and sigma-z) 'sared on the relationships with azimuthal and vertical angle fluctuations (expressed in radians) as a
described on page 2.3-16c of the FSAR is seceptable. In the FSAR, the applicant did not specify the method of .ete rmining the values i of sigma-y and sigma-z based on vertical temperature gradient. We i
will require that sigma-y and af gra-z based on vertical temperature gradient be determined from the standard curves in Meteorology and Atomic Energy - 1968,* Appendix A Figures A.2 and A.3, with sigma-z limited to 1000 meters. Curves for type G stability can be derived from the following relationships:
2 (sigma-y) (C) = 3 (sigma-y) (F)
(sigma-z) (C) = f (sigma-z) (F)
We conclude that the parameters and mode of control room dis-play are acceptable. However, the equation for the calculation of X/Q presented on page 2.3-16b of the FSAR, is not acceptable. We will require that the A/Q values on display in the control room be based on the following standard ground-level release model with '
adjustments for building wake X 1 Q " E[(pi)(sigma-y)(sigma-z)+cA)
U = average wind speed (meters /second)
A = minimum cross-sectional area of the building (1600 square i
meters)
- Reference '!, Appendix C to the SER.
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e = 0.5 pi = 3.14 )
i We have informed the applicant of our requirements regarding the methods of determining sigma-y, sigma-z and X/Q as described above. We have requested that the applicant include these methods in future amendaentt, to the FSAR. We will include them in the technical i
specif1 cations. We consider this matter to be resolved. l l
2.3.6 Conclusions
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In the SER, we stated that the applicant must provide more detniled information on the program for control reos monitoring of meteorological parameters, and that at least one additional year of onsite meteorological data must be submitted. Resolution of the control room monitoring program has been discussed in l
Section 2.3.3 of this report.
I In Amendment 24, the applicant submitted additional joint frequency distributions of vfnd speed and direction at the 25-f t level by atmospheric stability (as defined by the vertical temperature gradient between the 25-f t and 250-f t levels) for the ]
period May 1973 through April 1974.
These joint frequency distributions were submitted in accordance with the guidelines of Regulatory Guide 1.23 and are acceptable. As oescribed in Sections ' .3.3, 2.3.4 and 2.3.5 of the SER, the applicant had utilized the meteorological assumptions of Regulatory Guide 1.4 and we had t.oncluded, based upon our review of the meteorological data which had been submitted, that these b
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assumptions are adequately conservative. We have also reviewed the additional data submitted in Amendment 24.
Our evaluation, and that of our NGAA consultant (presented in Appendix B), confits that the relative concentration values for short-tera dif fusion estimates based on the meteorological assumptions in Regulatory Guide 1.4 are adequately conservative.
1 We censider this matter to be resolved.
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4-1 4.0 REACTOR 4.2 'fechanical Design 4.2.1 Fuel In the SER, we stated that the Single Rod Burst Tests had 1
been completed and would be docurented. The inforn tion with !
l regard to the Single Rod Burst Tests has been documented in l
Westinghouse Topical Report WCAP-8289 (Proprietary) and WCAP-6290
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(Non-Proprietary), "17 x 17 Design Fuel Rod Behavior During Simulated Loss of Coolant Accident Conditions," November 15 74.
Our review of this report has been completed and we concur with the conclusion made by Westinghouse that the test program indicates that the scaling down of rod geometries from the 15 x 15 design has no effect of practical significance on burst l ductility and burst temperatures. We consider this matter to :
be resolved.
In the SER, we stated that the details of the fuel surveillance program would be reported in a supplement to the SER. Our evalu-ation of the 17 x 17 fuel design, which has been completed, in-cluded assess:acnt of the engineering analysis, operating experience on similar fuel, confirmatory test results, technical specification j l
requirements and a surveillance program to nonitor the performance '
of the i radiated fuel. The surveillance program is essentially the
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1ast in a series of fuel design confirms.tions. The routine Westing- ]
i house surveillance program consists of three monitcrings; the pcwer i j
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distribution will be monitored using excore flux and incore move-able detectors, the cooltc; activity will be monitored for l 5
indications of loss of cladding integrity and finally, tho assemblies will be monitored by observations during the refueling i
operation.
Because tre 17 x 17 fuel design is new and will be introduced into a number of plants in-a short time, a comprehensive and generic i
surveillance program has been developed. The surveillance prograa ,
.q is outlined in the Diablo Canyon FSAR. One feature of the program is the insertion of four lead burnup fuel assemblies into the two Surry reactors. One assembly in each of the Surry reacters will have removable fuel rods. These assemblies will have been 61 men- i siona11y characterized prior to insertion and will be examined at l subsequent refuelings. The fuel assemblies will be examined for dimensions, fretting, bowing, gamma activity, cladding integrity (
and surface deposits. The first of these assemblies should be !
l examined near the end of 1975. This is compatible with the 1 I
anticipated startup of Diablo Canyon Unit 1. d I
For four of the first reactors to employ a full core 17 x 17 ,
l fuel assembly, the surveillance program includes insertion of fuel assemblies with removable rods. This is shown in Table 4.1. These assemblies differ from the'17 x 17 assemblies in the Surry reactors only by the number of spacer grids. The Surry assemblies have seven.
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4-3 grids while the 17 x 17 standard design assemblies have eight grids.
2' lightly larger fuc1 The slightly larger axial interval will cause j
rod deflections on the Surry assemblies than would the standard 1 The fuel rod damage mechanisms of l design 17 x 17 assemblies. 1 f
interest (e.g., fretting from flov induced vibration) are service i i
life dependent. Thus, the progression of any fuel damage can be acceptsbly monitored by the Westinghouse surveillance program.
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A commitment to insert a removable rod assembly had been requested for the four reactors (Table 4.1) which were originally Although expected to be the first to operate with 17 x 17 fuel.
l four reactors are committed to the insertion of a pre-characterized >
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assembly, only two of these are reqs. ired to be examined as part of j the generic 17 x 17 fuel surveillance program (Trojan and Diablo
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J Canyon).
In Amendment 27, the applicant provided a consniement to per-f form visual inspections of the peripheral rods of those fuel assemblies in Diablo Canyon Unit I which have been permanently removed to the spent fuel storage pool for the purpose of discharge. l
., l This program would last until the entire initial load of fuel had tj e l been examined or a similar program has been completed on two other ?
reactors in the United States and the NRC staff has approved i cancellation of the program for Diablo Cacyon Unit 1. If warranted .)
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l by the results if these visual examinations.,further investigation of the pre-characterized removable rods would be conducted.
We have concluded that the fuel surveillance program for Diablo Canyon, as well as the generic fuel surveillance program for Westing-j Louse 17 x 17 fuel, is acceptable. We consider this matter to be resolved.
In the SER, we stated that Westinghouse would dxument the justification for applying the results of certain tests made on 17 x 17 fuel assemblies with seven spseer grids to 17 x 17 fuel .
assemblies with eight spacer grids. This information has not yet been submitted and this matter is not resolved. We will. review this information when it is documented and will report the results of our review ia a future supplement to the SEK.
Subject to a favorable finding on the justification for applying the seven grid test results to eight grid fuel assemblies, we have concluded that the mechanical design of the Diablo Canyon fuel is acceptable.
4.4 Thermal and Hydraulic Design in the SER, we stated that westinghouse had submitted two topical reports: WCAP-8346, "An Evaluation of Fuel Rod Bowing," f j
j May 1974; and WCAP-8176 (Proprietary) and WCAP-8323 (Non-Proprietary),
"Ef fect of Bowed Rod on DNB," May 1974, that describe the analytical techniques used to predict bowing and the method used for assessing !
the ef fect of bowing on thermal performance of the fuel. We stated ,
that we were reviewing these topical reports, l
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4-5 Our review of WCAP-8176 and WCAP-8323 on the effect of a bowed rod on DNB has been completed. We have informed Westinghouse that these reports are acceptable. They describe the correlation between experimental results and calculational results obtained using the 1 i
Westinghouse design methods. While we have concluded that the report i
provides an acceptable data base model for determining the effects l I
J of rod bowing on DKB heat flux during the first fuel cycle, additional g
information is required to demonstrate that the model adequately fuel cycles.
predicts the effect of rod bowing on DN3 over subsequent The applicant is aware of our need for additional information on this I subject and is expected to provide it for our review in the near future.
We consider the effects of rod bowing on DNB heat flux af ter the first fuel cycle to be unresolved. We will report the resolution of this r.atter in a future supplement to the SER prior to a decision concerning issuance of operating licenses for Diable Canyon Units 1 -i and 2.
1 We have also completed our review of WF AP-6346, and have concluded that the fuel rod bowing predicted by the calculation model is accept-able for the 15 x 15 seven grid (rods-en-bottom) design based on observed bowing in irradiated fuel. We have also concluded that application of the same calculational methods is acceptable for I design evaluation of fuel rod bowing in the 17 x 11 fuel assembly l, design. However, a 17 x 17 fuel assembly surveillance program is l
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4-6 needed to confir i the validity of this model. The progrca, dis- )
l cussed in Section 4.2.1 of this supplement, includes bowing measure-ments on irradiated fuel in the two Surry reactors and visual 9
observations in the Diablo Canyon and Trojan reactors. The-fuel
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rod bowing model vill be reviewed as the data become available. !
If changes to the model are needed, the fuel rod bowing-effects in Diablo Canyon will be reevaluated and we will make changes to the tec. alcal specifications if appropriate. We have concluded
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that the Diablo Canyon fuel design takes fuel rod bowing into account in an acceptable manner and consider that our concerns j regarding this aspect of the Diablo Canyon design, as described in Section 4.4 of the SER, are resolved.
In the SER, we stated that, if the results of the non-uniform departure from nucleate boiling (DNB) tests were not avail-
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i able wher. the technical specifications for Diablo Canyon were finalized, we would require that the minimum allowable departure from nucleate boiling ratio (DNBR) be increased 5 percent above e
that required to satisfy the 95/95 criterion. These results have not yet been documented. Our position on this matter remains unchanged.
In the SER, we stated that we would review the results of certain elements of the verification test program for the THINC code when they became available. In the event that sufficient I
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verification could not be obtained from the combined test and j analytical programe, we stated that restrictions would be included in the technical specifications for Diablo Canyon to maintain required margins to fuel rod damage.
These results have been documented in topical reports WCAF-8453, l
(Proprietary) and WCAP-8454 (non-Proprietary), " Analysis of Data l We l from the Zion (Unit 1) THINC Verification Test," December 1974 We consider have not yet completed our evaluation of these reports.
this matter to be caresolved. The resolution of this matter will be reported in a future supplement to the SER.
We have concluded, subject to favorable ' resolution of the out-standing items described above, that the thermal and hydraulic design of the Diablo Canyon reactora is acceptable, and that these reactora can operate at the proposed core power levels.
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4-8 TABLE 4.1 IRRADIA1 ION SCHEDULE OF PRECHARACTERIZED FUEL ASSEMBLIES FOR THE CENERIC 17 x 17 SURVEILLANCE PROGRAM Current
- Fuel Estimate of Number of Precharacterized Reactor Cycle _
_Startur_ _ 17 x 17 Assemblica Surry 1 2 Dec '74 2(b)
Surry 2 2 May '75 2 ID)
Trojan (c)(d) 1 Jun '75 1 Diablo Canyon I} 1 Jan '76 1 Farley 1 May '76 '1 Sequoyah 1 Dec '77 1 i
Beaver Valley 1 Dec '75 's Salem 1 Nov '76 0 TOTAL 8 (a) As of December 19, 1974 (b) 7-grid assemblies, 1-each removable rod assembly (c) Selected as lead plant prior to August 1974 ,
(d) Currently expected to execute surveillance program l
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6.0 ENGINEERED SAFETY FEATURES
- 1 6.2 Containment Systems 6.2.1 Containment Functional Design 3 In the SER, we r.tated that the applican; had performed the containment subcompartment analyses using the Trar.sient )
1 Mass Distribution (TMD) code with the augmented critical flow s i
correlation, and that we _had requested that the ana'yses be re-done using the more conservative IMD code without the 1
augmented critical flow correlation. The appilcant had presided us with an analysis of the pressure response within the pressurizer enclosures and the loop compartments using the non-augmented critical flow correlations and se were reviewing these analymes. We expected to receive similar analyses for the reactor coolant pipe annulus, reactor vessel annulus and lower reactor cavity.
We have comp 1 ted our review of the ar.: lyses of the f 1
l pressurizer enclosw.es and loop compartments using the 1HD '
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code without the augmented critical flow correlation. We 4 j have reviewed the noding arrangements used and the assumptions i made and have ccncluded that the calculated maximum differential-pressures for these compartments are acceptable. We consider 1
this tutter to be resolved.
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6-2 In Amendments 25 and 26 to the FSAR the applicant hau provided information showing the modeling assumptions and dimensions used for the analysis of the reactor cavity, reattor vessel ar.aulus, and the reactor shield structure. The analy,,is of the reactor cavity and vessel si tulus was performed usin.g the IND code withoutt ' he augmented critical flow correlation.
The analysis assumed a limited displacement rupture of a scoctor i
coolant system hot leg at the reactor vessel nozzle weld as l l
l the design basis accident. We have reviewed the noding arrangement
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i used and the assumptions made concerning piping and vessel i
insulation behavior and conclude that the dif ferential prentures calculated for the reactor cavity structures are acceptable i
for the assumed break. We will require, however, that the applicant provide additional information on the geometry of the system to justify the assumed limitation on the size of the opening that can result from a break at the reactor vessel !
nozzle weld.
With regard to the pressure response of the reactor coolant system pipe penetrations through the reactor shield structure, l l
l the applicant has not performed an acceptable analysis. We will require that the applicant (1) analyze the response of the piping penetration to a pipe break within the penetration using l
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6-3 either the TMD code without the auCaented critical flow correlation or another acceptable method of analysis; or (2) justify that a reactor coolant system pipe break need not be postulated in the reactor shield structure pipe penetration according to the reconsnendations of Regulatory Guide 1.46. ,
We will report the resolution of these items in a future supplement to the SER.
1 Emergency Core Cooling System (ECCS) j 6.3 !
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6.3.1 Design Basea In the SER, we stated that we had identified certain )
locations where a single incorrectly positioned motor-operated valve could result in total loss of the intended ECCS safety function. We also stated that the applicant would need to either lock out power to these motor-operated valves or modify the design to obtain an equivalent degree of protection.
In Amendment 27, the applicant provided a commitment to lock out power to those motor-operated valves identified by the NRC staff as affecting the function of the 'iCCS if spurious operation were to occur. We have reviewed this commitment and concluded that it provides an acceptable method of meeting the single failure criterion. The valves for which we will require locking out of power are those identified in item 7.12 ot a request for additional infor-mation (see items 42 and 45 in Appendix A to the SER). The l
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6-4 appropriate requirements for removal and restoration of I power will be included in the technical specifications.
We consider this matter to be resolved.
i 6.3.3 Performance Evaluation In the SER, we stated that the requirencnts of 10 CFR Part 50.46(b), " Acceptance Criteria for ECCS for light Water Cooled I;uclear Power Reactors," are applicable te Diablo Canyon j Units 1 and 2, and that the-applicant had submitted an analysis of the performance of the emergency core cooling system in accordance with this provision in Amendment 15 to the FSAR, dated August 2, 1974. i j
'Je have completed our review of the Westinghouse model, and on the basis of that review, the applicant will revise the analys's submitted in Amendment 15, and will submit a revised analysis for Diablo Canyon Units 1 and 2. We will review that analysis and will report the results in a future supplement to the SER prior to a decis in concerning the issuance of operating licenses for Diablo Canyon Units 1 and 2.
6.3.5 Conclusion In the SER, we stated that the acceptability of the ECCS was still being evaluated. Specifically, we stated that (1) i the applicant would need to either lock-out power to certain motor-operated valves or modify the design to meet the single l
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6 f ailure criterion in another manner; and (2) the applicant's-ECCS evaluation model and analysis results would need to be .
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found acceptable. We consider the first -item to be resolved - i as discussed above. We will report our conclusions regarding acceptability of the ECCS in a future supplement to the SER j
as discussed in Section 6.3.3 of this report. !
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- 7. 0 INSTRUMENTATION AND CONTROLS
- 7. 2 - Reactor Trip System '
7.2.2 .Mactor Trip Systes Actuation Logic 1
7.2.2.1- Physical' Separation' In the SER, we stated that we had.found that the physicel I 1
separation in the solid state protection system racks was in-adequate. The input and output wire bundles terminated'at a connon connector of the 1.olation board. The center pins of-the connectot are not used and this provides separation at that point; however, there were no physical barriers or protection 4
to separate the input and output wire bundles at the locations 4
where they are in close proximity as they are routed from the '
connectors. The applicant agreed to install barriers to provide 1 this separation.
The applicant has provided the physical barriers. The.
barriers separate the input and output wire bundles.at the locations where they are in close proximity. We have reviewed this design change and reviewed the. detailed drawings of the barriers during a site visit. We have concluded that the barriers separating'the input and output wiring provide an adequate means of meeting the requirements of General Design Criteria 22 and 24 'and the design is therefore acceptable. We consider this matter to be resolved.
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7-2 7.2.2.2 e p ectrical Isolation In the.SER, we stated that the photodiode isolators used to electrically isolate the safety signals f rom non-safety functions, as implemented in the solid state protection system, had not been qualified as acceptable isolation devices. We informed the applicant that tests would be required to verify that these isolators will meet the design basis requirements for system isolation. The applicant indicated that he would provide qualification test l procedures and results.
The applicant has provided test procedures and results which verify that the photodiode isolators will meet their design basis requirements for system isolation' .
We have reviewed' the information and concluded . that the qualification tests of the photodiode isolators provide an acceptable basis for meeting the requirements of. General Design Criterion 24 and the design is therefore, acceptable. We consider this matter to be resolved.
7.2.2.4 Conclusions In the SER, we stated that the solid state protection system would be acceptable, providing that: (1) the photodiode isolators were adequately qualified; (2) adequate separation or barriers were provided for the input and output wiring; and (3) the seismic gealification program conformed to our requirements. We consider the first two items to be resolved, as discussed above. We consider the third item to be unresolved.
6
, . _ . - . , _ - . . _ . . . x;A
7-3 We will report the final resolution of the seismic qualification program in a supplement to the SER.
Engineered Safety Featares_ Actuation System 7.3
_ Changeover f rom Iniection te Recirculation Mode 7.3.4 In the SER, we stated that the design would be acceptable, subject to satisf actory resolution of (1) our concerns about lockout of power to ECCS motor-operated valves, and (2) compliance 279-1971. The first of the level instrumentation tv IEEE Std item has been resolved as discussed in Section 6.3.1 of this Three water level instrument channels are provided supplement.
Two out of three logic f or each ref ueling water storage tank.
is provided for regenerative heat removal pump trip and initiation We have reviewed the implementation of of low level alarm.
279-1971, and conc u eldd the revised design in accordance with IEEE that it provides an acceptabic means of meeting the requirements of General Design Criteria 20, 21, 22 and 23 and the design is, l
therefore, acceptable. We also consider this matter to be I resolved. l Accordingly, as discussed in the SER, we have concluded l that the operator will have sufficient time to perform the actions required to change over to the recirculation mode of operation, and the design meets the Commission's requirements, and is acceptable. l t
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9-1 9.0 AUXILIARY SYSTEMS 9.2 Fuel Storage and Handling 9.2.3 Fuel Handling Systec In the SER, we stated that we had not reac.hed agreement with the applicant regarding storage of spent fuel in locations where it
- ould not be damagedo ' y a dropped fuel cask.
In Amendment 22, the applicant described a fuel storage scheme j
where only spent fuel assemblies which have decayed for 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> !
i or more would be stored where they could be damaged by a dropped fuel cask. The applicant stateci that no more than 20 of these assemblies, all of which have decayed for more than 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />, could be damaged by a dropped fuel cask. "
We have reviewed this information and concluded that the folloving provisions, in addition to those proposed by the applicant, are necessary in order *o determine that no more than i
20 elements, all of which have decayed for more than 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />, could be damaged by a dropped fuel cask:
(1)
No cask handling operation near the spent fuel pool will be performed unless spent fuel at any location in rack 5 or rack 6 has decayed for at least 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> since shutdown.
(The applicant proposed a similar restriction but it in-cluded all of rack 5 and part of rack 6 instead of all of )
I both racks).
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(2) kedundant interlocks will'be provided on the cask handling ,
1 crane.to limit the position'of the cask to the outermost corner of the cask recess area whenever the cask is being raised or !
lowered. (This'is in addition to the restrictions already {
1 described in the FSAR.) ' '
With these provisions, in addition to' tbse already described in the FSAR,' we concur that, based upon the configuration of the '
spent fuel pool, no more than 20 spent fuel assemblies, all 'of which have decayed for .at 'least~1000 hours, could bs damaged by a dropped fuel cask. The applicant has agreed to these additional provisions,
~
The radiological consequences of a dropped fuel cask accident in -
. I volving damage to 20 spent fuel assemblies which have decayed 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> are acceptable, as diwussed in Section 15.2.2 of this supplement. Accordingly, subject to the applicant's documentation'of a commitment to the provisions described above, we consider this matter to be resolved.
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15-1 l l
15.0 ACCIDENT ANALYSES 15.1 ceneral !
In the SER, ve stated that we vould require heaters for the auxiliary building charcoal filters for humidity control in the event of residual heat remo al system leakage following a loss of coolant accident. i In Amendment 22 to the FSAR, the applicant agreed to provide !
such heaters, and described an analysis of the humidity of the air a I
which enters the auxiliary building charcoal filters. In this )
l analysis the applicant assumed that air entering'the containment /
l l L 11 ding would be heated 15'T before it reaches the auxiliary.
l l building charcoal filters. We have requested the applicant to I provide justification for this assumption. We consider this matter unresolved. We will report the resolution of this matter l in a future supplement to the SER.
15.2 Desigr. Basis Accident Assumptions 15.2.2 Fuel Handling Accident j Since the SER was issued, the applicant has submitted, in Amendment 27 of the FSAR, an analysis of the consequences of a fuel cask drop accident. He have evaluated the radiological con-sequences of this accident based on damage to a maximum of 20 O
15-2 spent fuel assemblies, all of which have decayed for at least 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> (see Section 9.2.3 of this supplement). The assumptions used for this evaluation are.given in 72ble 15-2, while the dose conse-quences are shown in Table 15-1. The calculated thyroid dose at the exclusion zone boundary is 18 Res. The conclusion that no freshly irradiated fuel could be damaged is an important part of this evaluation since damaging a single assembly which has decayed for only 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> since shutdown could result in a thyroid dose of as much as 23 Ren at the exclusion zone boundary.
We have concluded that the dose consequen:es of this accident are well within the guideline values of 10 CFR Part 100 and are therefore acceptable.
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i 15-3 i TABLE 15-1 PMENTIAL OFFSITE DOSES DUE TO FUEL CASK DROP ACCIDDfT_
% Rour Course of Accident Exclusion Boundary Im Population Zone (800 Meters) (9600 Meters)
Ubole Body Thyroid Whole Body Thyroid, d _
Acciden (Rem)
<1 <1
- Fuel Cask Drop 1B <1 I
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15-4 I
TABLE 15-2 1 PUEL CASK Dit0P ACCIDENT CALCULATION INPUT PARAMETERS Shutdown Time 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> Number of Fuel Assemblies Damaged by Cask Drop 20 Power Peaking Factor 1.65 l j
Todine Tractions Released from Pool Elemental 75%
Organte 25%
Effective Vilter Efficiency Elemental 90%
Organic 70%
X/Q Values, sec/o 0-2 hours t J00 meters 5.45 x 10~' j 0-2 hours # 9600 meters 2.2 x 10-5 Watertight integrity of spent fuel pool maintained.
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22-1 22.0 _ CONCLUSIONS In Section 22 of Supplement No. I to the SER, we stated that several items were still outstanding, and that favorable resolution of these items would be required before operating licenses for Diablo Canyon Unitt 1 and 2 could be issued. A number of these have been resolved, in this supplement.
A revised status report on all of these j
item.: is given below:
j The matters regarding meteorology have been resolved (Sections l (1) l 2.3.3 and 2.3.6 of thic report).
(2) The applicant has provided additional information on the effects i of tsunami waves esased by near-shore generators; however, our evaluation of this information has not been completed (Sections 2.4.2, 2.4.3 and 2.4.5 of Supplement No.1) . t (3) Our comparative evaluation of the Hosgri and Santa Lucia Bank f aults, and our evaluation of the earthquake potential of the Hosgri Fault have mot been completed (Sections 2.5.1 and 2.5.2 of Supplement No. 1).
(4) The applicant has provided additional information on the poten-tial consequences of pipe breaks outside containment; however, our evaluation of this information has not been completed (SER Section 3.6).
(5) The applicant has not yet submitted required inf ormation confirming the seismic qualification of Category 1 instrumentation and electrical equipment (SER Sections 3.10 and 7.8).
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22-2 (6)
Documentation has not yet been provided justifying the use of' the i results of tests of 7-grid assemblies to prove the acceptability '
of the 8-grid design (Section 4.2.1 of this report). '
(7) Our evaluation of the results of the single rod burst tests has l
been completed, and this matter is resolved (Section 4.2.1 of this report).
(8) Our evaluation of the 17 x 17 fuel rod surveillance program has been completed, and this matter is resolved (Section 4.2.1 of. !
this report).
(9) The matters regarding uncertt nties in the thermal a4J hydraulic design have been partially resolved (Section 4.4 of this report).
(10) The matters regarding subcompartment pressure calculations using the Transient Mass Distribution (TMD) program have been partially resolved (Section 6.2.1 of this report).
i (11) The item regarding lockout of power to certata motor-operated 1 i
ECCS valves has been resolved (Sections 6.3.1.-6.3.5 and 7.3.4 of '
this report). I j
(12) The applicant has not submitted a revised ECCS analysis.in i' accordance with the Final Acceptance Criteria (FAC) (Sections 6.3.3 and 6.3.5 of this report) .
1 (13) The matters regarding physical and electsical separation in the $
solid state protection system have been resolved (Sections 7.2.2.1, 7.2.2.2 and 7.2.2.4 of this report).
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22-1 i
l (14) W applicant has provided additional information regarding physical separation in the process analog system; however, our evaluation of this information' has not been completed (SER Section 7.2.3).
J (15) Our evaluation of the West',nghouse generic AWS model is not d yet completed (SER Section 7.2.5).
(16) The applicant has not provided adequate information to confirm the environmental qualif1 cation of Category I instrumentation and electrical equipment (SER Section 7.8).
(17) The astters regarding the consequences of a postulated cask drop have been resolved (Sections 9.2.3 and 15.2.1 of this report).
(18) Our evaluatiori of the proposed design modifications to the aux-iliary building to bring about a reduction of the doses is the !
event of RHR leakage during the recirculation phase following a postulated LOCA has not been completed (Section 15.1 of this ;
report). i (19) The applicant has submitted information regarding the guidance in certain WASH documents which pertain to the operational quality assurance program; however, our evaluation of this matter has not been coupleted (Item 99 of Appendix A to Supplement No.
I and Items 109 'ind 116 of Appendix A to this report).
Subject to f avorable resolution of the outstanding matters j desexibed above, the conclusions as stated in Section 22 of the SER rensin anchangeJ.
____.____m.__m_ . _ _ _ . _ _ _ ____ ____________m_______.____..__.-. __ _ , _ _ . _ _ _ _ _ _ _ _ . _ _
'A-1 APFFl@II A CONTINUATION OF THE CHRONOLOGY OF THE RADIOLOGICAL REVIEW 109. January 30, 1975 Submittal of Amendment No. 25 consisting primarily of additional information on subcompartaent pressure' calculations and
.the operational quality assurance program.
110. January 31, 1975 Supplement No. I to the Safety Evaluation Report issued. -
111. February 6, 1975 Meeting with applicant.to discuss seismic I and environmental qualification.of. electrical-equipment, and physical and electrical separation in the solid state protection and process analog systems..
112. February 7, 1975 Meeting with applicant to discuss the geology and seismology of the Diablo Canyon site.
113. Fa'oruary 12, 1975 Request No. 11 to applicant for additional information on geology and seismology.
114. February 18 19, 1975 ACKS Subcommittee Meeting in San Luis Obispo, .
115. March 3, 1975 Meeting with PG&E management to discuss the geology and seismology of the_Diablo Canyon site.
116. March 26, 19:5 Submittal of Amendment No. 26 consisting primarily of additional information on subcompartment pressure calculations and on recent guidance for quality assurance programs.
117. April 3, 1975 Letter to applicant requesting additional . l information on the 5sergency Core Cooling l System.
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118. April 4, 1975 Meiting with applicent to discuss items relative to'the seismic design of Diablo Canyon.
119. April 7, 1975 Letter from appif cant providing supp1 wentary pages to report entitled " Westinghouse Protection System Noise Tests" whien is referenced in Section 7.2 of the 23AR.
120. April 7, 1975 Letter from applicant regarding participation in Westinghouse program to evaluate corrosion resistance of possible alternate steam generator tube materials in an operating plant, 121. April 10, 1975 Second OL Prehearing Conference, j 122. April 15, 1975 Letter from applicant transmitting 1974 Annual Financial Report.
123. April 23, 1975 Submittal of report concerning jet effects analysis for postulated pipe breaks outside-containment.
124. April 28-May 2, 1975 Review of the Diablo Canyon seismic design at PG&E offices in San Francisco.
125. April 30, 1975 Submittal of Amendment No. 27 consisting of additional information required for the resolution of outstanding items in the SER.
126. May 1-2, 1975 Site visit and meeting related to electrical, instrumentation and control systems.
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Ei[VIEONMEWTAL REsEARCH GC@@@S70.UGO
@ ' t Silver Spring, Maryland 20910 l
April 10,1975 !
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Mr. Frank Schroeder, Acting Director Division of Technical Review
['O*2,'Tb 323 United States Nuclear Regulatory-
)
Comission 4 Washington, D.C. 20555 i
Dear Mr. Schroeder:
This refers to the letter of January 23, 1975 from William P. Gamill, Chief, Site Analysis Branch, Division of Technical Review requesting coments on the following:
Diablo Canyon 1 and 2 Pacific Gas and Electric Company Final Safety Analysis Report Amendment ho. 24 dtd.1/15/75 These coments are attached.
Sincerely, h s-*< Ae[i 4 Isaac Van der Hoven ..
Air Resources 1. laboratories * -
. , . /. ,.
Attachment -
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cc: E.H. P.arkee, Site Analysis Br., USNRC l
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4421
m-Corments on 4
Diablo Canyon Site Units 1 and 2 Pactfic Gas and Electric Company Final Safety Analysis Report Amendment 24 dated January 15, 1975 Prepared by Air Resources Laboratories National Oceanic and Atmospheric Administration April 10.1975 Using the latest available data of wind speed and direction (May 1973 -
through April 1974) at the 25-f t. level and corresponding temperature differences between 25 and 250 f t., we have calculated that for a . ground lev 10 g1 release there is a probability that a relative concentration of 1.8 x see m 5 w!11 be exceeded 5% of the time at a distance of 800 m dowmuind.
The atmospheric stability was categorized according to the temperature gradient criteria listed in AEC Safety Guide 1.23. The release time was assumed to be from 0-2 hours.
For the long-ters annual release a ground source was assumed. This is in contrast to the '0-m release height assumed by the_ applicant (p. 2.3-24).
Figure ').2-22 sh, .s the top of the tallest vent duct to be about 32 m above grade, while the adjacent reactor containment building is about 67. m in height above grade. We would thus assume toat the vent release would be caught ir, the building wake. We calculate from the 25-ft. wind data that the highest annual average relative concentration ill be at the 800-m site boundary towards the NW at a value of 2 x 10- sec m-3 F
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._. 1 APPEICIX C ERRATA TO THE SAFETY EVALUATION REPORT _
AND SUPPLEMENT NO. I 1
1
')
Safety Evaluation Report Pm Line insert a comuna af ter " system" j 6-3 1 6-15 12 replace "same" with "same" l
1
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Supplement No. 1 l l
Fage, line_ 1 22-2 5 replace "has been provided in a f ilCAP report" with "has not been provided" 22-2 8 replace " report" with "itan" 22-3 20 replace " turbine" iith " auxiliary" i
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i DISTRIBUTION: 8 CENTRAL FIL5S NRR-RDG MAR 31 1975 SAB William P. Gammill, Chief, Site Analysis Branch, TR '
, INFORMAL MEMO RE: TELEPHONE CONVERSATION WITH DOUGLAS: HAMILTON OF EARTE SCIENCE ASSOCIATES-(CONSULTANTS TO PG&E RE: .DIABLO CANYON POWER PLANT). -
DATE: 24 March 1975 To keep you informed of progress' on the Diablo Canyon work, a summary of
! my 1:00 pm call to Douglas Hamilton follows.
After our trip to Houston of 18 and 19 of March to observe Shell Oil Co.
I and Western Geophysical Company data, D. Hamilton and C. R. Willingham I
were to integrate recently released or soon-to-be released USGS offshore I seismic date with their previous work. They were to keep in mind the i
Shell data we observed and the availability of Western Geophysical data.
Progress to date:
(1) Track charts of the Western Geophysical data and of the Shell data that was shown to un have been added to a master. track chart of the area. Thus duplication of coverage can be determined.
(2) Hamilton is recommending to PG&E that all Western data south of the latitude of Arroyo Grande be purchased and that line 12 opposite i the site be purchased with additional processing og. restacking, !
special filtering and migration. He latter will help establish the precise location of the "Hoagri" fault offshore from the plant. ,, !
i (3) The U.S.G.S.1972 Bartlett cruise data has been ordered and should '
arrive in microfilm- form tomorrow. Several days will be required to enlarge it to a working paper size. This is the data that was collected and analysed under the direction of Eli Silver and Roland Von Huene.
1 (4) The U.S.G.S.1972 or earlier Polaris Cruise Seismic data has been requested. However, they have been informed that the data will not be released for several weeks. This data was collected under the direction of Holly Wagner and Steve Wolfe.
)
l-Thue are three unknowns in the time table of required events which will -'
effect the completion of a combined analysis of,all data.
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Forma AFC 318 (Rev. 9-58) ABCM 0240 *A u. s; oovanuuswr eniuvene orness set 4.ess-see -
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William P. Gammill MAR 31 YRS
- 1. When will the Polaris data become available?
- 2. When can the Shell data be observed?
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- 3. When will they receive the Western Geophysical data.
The last item is presently of the least concern.
l l Carl Stepp has proposed a meeting at Menlo Park with Hamilton,' Yerkes j McKuen, You Huene and us to review progress to dats. namilton believes such a meeting could be held but without the Polaris information.
Renner B. Hofmann i
Seismologist Site Analysis Branch I Division of Technical Review l 1 Office of Nuclear Reactor Regulation i
I cc: H. Denton F. Schroeder l
D. Allison O. Parr
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DATEM _ _
Porns AllC-318 (Rev,9 58) ABCM 0240 A' u. s: ooVERNMEN'T PRifdTINS OFFICE 8 9974 824 864