ML20033A934
| ML20033A934 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 10/31/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0515, NUREG-0515-S04, NUREG-515, NUREG-515-S4, NUDOCS 8111300180 | |
| Download: ML20033A934 (71) | |
Text
GF
.~.
"o NSIL fod 4M S
Noy] 9 ggg Ar Ti NUREG-0515 ma g p, Supplement No. 4
}
%g-1 -
Safety Evaluation Report related to the construction of
. Allens Creek Nuclear Generating Station, Unit No.1 Docket No. 50-466
! Houston Lighting & Power Company i
g.
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation October 1981 j a"%,,
f f = = !!(
i i
NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the foi!owing sources:
1.
The NRC Public Document Room,1717 H Street,, N.W.
Washington, DC 20555 2.
The NRC/GPO Sales Program, U.S. Nuclear Regulatory Commission, Washington, DC 20555 3.
The National TerhnicalInformation Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, it is nd intended to be exhaustive.
Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforce-ment bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence, The tollowing documents in the NUREG series are available for purchase from the NRC/GPO Sales Pro-formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and NRC gram:
booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of Federa!
Regulations, and Nuclear Regulatory Commission Issuances.
Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commis-sion, forerunner agency to the Nuclear Regulatory Commission.
Documents availabM,' rom public and special technical libraries include all open literature items, such as bocks, journal and periodical articles, transactions, and codes and standards. FederalRegister notices, federal and state legislation, and congressional reports can usually be obtained from these fibraries.
Documents such as theses, dissertations, foreign reports and translations, and non NRC conference pro-ceedings are available for purchase from the organization sponsoring the publication cited.
Single copies of NRC draf t reports are available free upon written request to the Division of TechnicalInfor-mation and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555.
GPO Printed copy price:.$4.25
NUREG-0515 Supplement No. 4 Safety Evaluation Report related to the construction of Allens Creek Nuclear Generating Station, Unit No.1 Docket No. 50-466 j
Houston Lighting & Power Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation October 1981 r ~,,
a p
Yh.
~......
u..
4 ABSTRACT The Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission his issued Supplement 4 to the Safety Evaluation Report (SER) for the application filed by Houston Lighting & Power Company for a construction permit (CP) to construct the Allens Creek Nuclear Generating Station Unit 1 (Docket No. 50-466).
Tha Allens Creek site is in southern Austin County, Texas, about 45 miles west of Houston. This Supplement presents the staff's analysis of information submitted by the applicant to show compliance with the Commission's rule on j
emergency planning that was promulgated since SER Supplement 2 was issued.
Tha report also updates the staff's reviews of financial qualification and g:neric issues and reports on the staff's review of updated design information provided by the applicant.
The staff's analysis in Supplement 4 completes the staff's safety evaluation required before a decision is made on issuing a CP for the proposed facility.
Th2 staff concludes that a permit can be issued for the construction of Allens Creek Nuclect Generating Station Unit 1.
i i
i 2
6 -
1 iii
- Aliens Creek SSER #4
?
CONTENTS P. age
. ABSTRACT............................,.................................
iii
1.0 INTRODUCTION
AND GENERAL DISCUSSION..............................
1-1 l
1.1 Introduction................................................
1-1 l-
- 1. 9 Outstanding Items...........................................
1-3 1.10 Permit Conditions...........................................
1-3 2.0 SITE CHARACTERISTICS............................................
.2-1 2.2.4 Nearby Industrial Facilities........................
2-1 l
2.5 Geology and Seismology.....................................
2-2 2.5.4 Stability of Subsurface Soils.......................
2-2 3.0 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS...................
3-1 3.8.1 Steel Containment...................................
3-1 3.8.2 Containment Interior Structures.....................
3-1 3.8.3 Design of Other Category I Structures...............
3-3 4.0 REACT 0R..........................................................
4-1 4.2 Fuel System Design..........................................
4-1 13.0 CONDUCT OF OPERATIONS...........................................
13-1 13.3 Emergency Planning.........................................
13-1 13.3.1 Item A.............................................
13-1 13.3.2 Item B............................................
13-6 13.3.3 Item C............................................
13-7 13.3.4 Items D and E......................................
13-9 13.3.5 Item F.............................................
13-10 13.3.6 Item G............................................
13-11 13.3.7 Item H.............................................
13-12 13.3.8 Summary...........................................
13-13 i!.
[
15.0 ACCIDENT ANALYSES...............................................
15-1 t
15.2 Abnormal Operational Transients............................
15-1 i
Allens Creek SSER #4 v
i e
L
CONTENTS (Continued)
Page j
20.0 FINANCIAL QUALIFICATIONS........................................
20-1 20.1 Introduction...............................................
20-1
.20.2 Construction Cost Estimates................................
20-1 20.3 Bases for Analysis........................................
20-2 20.4 Rate of Return on Common Equity............................
20-2 20.5 Internal Cash Generation...................................
20-5 20.6 Interest Coverage....................................
20-6 20.7 Capital Structure..........................................
20-7 20.8 Conclusion.................................................
20-7
21.0 CONCLUSION
S......................................................
21-1 REFERENCES............................................................
R-1 APPENDIX A - CONTINUATION OF CHRONOLOGY APPENDIX B - ERRATA TO ALLENS CREEK SAFETY EVALUATION REPORT SUPPLEMENT 2 APPENDIX C - NUCLEAR REGULATORY COMV;SSION STAFF GENERIC ISSUES APPENDIX D - LETTERS ON POSTACCIDENT INERTING SYSTEM APPENDIX E - MEMORANDA ON STATE AND LOCAL EMERGENCY PREPAREDNESS FOR ALLENS CREEK 4
4 Allens Creek SSER #4 vi f
w
+
.i
- 1. 0 INTRODUCTION AND GENERAL DISCUSSION 1.1 Introduction The Nuclear Regulatory Commission (NRC) Safety Evaluation Report (SER) in the matter of the application by the Houston Lighting & Power Company (the applicant) to construct and operate the proposed Allens Creek Nuclear Generating Station Units 1 and 2 (Allens Creek or the facility) was issued in November 1974.
SER Supplement 1, which was issued in June 1975, presented the staff evaluation
-of additional information submitted by the applicant after the issuance of the SER and identified 13 matters requiring additional information or resolution.
That supplement also included a discussion of the NRC staff's response to each of the comments made by the Advisory Committee on Reactor Safeguards (ACRS) in its letter of December 12, 1974 to the Commission.
SER Supplement 2 (NUREG-0515) which was issued in March 1979, presented the staff's evaluation of additional information submitted by the applicant after the issuance of SER Supplement 1.
That information was provided by Amendments 30 through 51 to the Preliminary Safety Analysis Report (PSAR) and, in addition to information addressing the 13 remaining issues in Supplement 1, included (1) descriptions of the changes in the facility resulting from the elimination of the second unit and from the applicant's updating of design criteria and preliminary design and (2) the applicant's responses to the staff's new and revised positions that resulted from the staff's further review against NRC updated licensing requirements and review guidance.
This information submittal followed the applicant's September 15, 1975 announcement of an indefinite deferral of the Allens Creek project and advising the NRC staff on October 7,1976 that the Allens Creek project would be reactivated with only Unit 1.
SER Supplement 3, which was issued in July 1981, presented the staff's evaluation of the applicant's compliance with requirements proposed as a result of the TMI-2 accident.
Those proposed requirements were set forth in NUREG-0718, Revision 1, " Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License," dated June 1981.
The applicant's responses to these proposed requirements were provided in Amendments 57 and 59.
This supplement presents the staff's analysis of PSAR Amendments 52 through 56, 58, and 60, and of financial information (see Sec. 20.0) provided by the applicant in a letter dated August 12, 1981.
As noted in the introduction to SER Supplement 3, the applicant and the staff had not resolved a difference in interpretations of the applicability of the proposed " Standard Code for Concrete Reactor Vescels and Containments," ACI-ASME (ACI-359).
Resolution of this matter is discuse.d in Section 3.8.2.
A continuation of the chronology of the staff's review that was included as Appendix A of SER Supplement 2 is included as Appendix A of this report.
Errata to SER Supplement 2 are included as Appendix B of this report.
Allens Creek SSER #4 1 _
A discussion ~of the staff's program for tt' resolution of generic issues and
-how the issues relate to a decision on the issuance of a construction permit for Allens Creek Nuclear Generating Station Unit 1 was presented in Appendix C to SER Supplement 2.
Subsequent changes in that program are described in Appendix C to this report.
As noted in Section 18.0 of Supplement 2, " Report of the Advisory Committee on Reactor Safeguards," the Committee saw no need to reopen its review of the Allens Creek plant after the project was reactivated as a single-unit facility.
In July 1981, the Committee advised the staff that it did not intend to review the applicant's response to the TMI action items after the staff issued SER Supplement 3.
Thus, the Committee's comments to the Commission before a decision on issuance of a construction permit for Allens Creek Unit 1 are limited to those in its letter of December 17, 1974 enclosed as Appendix I to SER Supplement 1.
However, as a part of the Committee's ongoing general review of postaccident hydrogen control systems, the Committee did request that the staff discuss with it, on September 10, 1981, the postaccident containment inerting system as proposed for Allens Creek. Although this discussion did not result in transmittal of Committee comments to the Commission, the Committee by letter of September 16, 1981 (Appendix D) to the staff did express an interest in a further review of this system in its ongoing review of hydrogen control qi f systems.
Specifically, the Committee stated that it desires to review the pro-posed design of the postaccident inerting system (PAIS) before it is approved by the NRC staff or other means of hydrogen control are precluded.
A copy of the Committee's letter was transmitted from the staff to the applicant by letter of October 8, 1981.
In Item II.B.8(3) of Supplement 3 the staff described the applicant's commit-ment to provide, within 2 years of the issuance of the construction permit, analyses and test data to verify compliance with staff positions on the hydrogen control system, and noted that because the applicant had made only a preliminary commitment to postaccident inerting, the submittal to the NRC of the final selection and design of the hydrogen control system should be based on a com-parison study of these alternative systems, including cost comparisons.
Additional staff responses to the Committee's letter of September 16, 1981 were provided by letter of October 7,1981 (Appendix D) to the Committee.
A copy of the staff's letter was transmitted to the applicant by letter of October 29, 1981. The staff stated that it will work with the applicant to ensure that such issues as hydrogen source-term design criteria, system actuation criteria (including inadvertent actuation), containment pressurization, and equipment survivability, among others, can be addressed at least in a preliminary manner about March or April 1982.
Further consideration suggests that a specified time.
increment after issuance of a construction permit, rather than a specified date, would give the applicant an opportunity to develop more significant information for consideration by the staff and the Committee.
Therefore, based on these considerations and the findings of the review reported'in SER Supplement 3, the staff recommends the issuance of a construction permit with the following conditions:
(1) The applicant shall report on the plans for its program on the hydrogen control systems to the NRC staff 1 month after the issuance of a construction permit.
Allens Creek SSER #4 1-2
I l
(2) The applicant shall report the status of its program on hydrogen control systems to the NRC staff 6 months, 12 months, and 18 months after the issuance of a construction permit.
(3) The applicant shall provide its final report on analyses and test data for the proposed hydrogen control system, including results of comparison studies of alternative systems to the NRC staff within 2 years of the issuance of a construction permit.
(4) The applicant snall maintain the capability of incorporating any of the alternative systems in the final design until the staff has reviewed amd accepted the final report of Condition (3).
)
Appendix E contains memoranda on st' ate and local emergency preparedness.
1.9 Outstanding Items In Section 1.9 of SER Supplement 2, the staff stated that it had completed its review of additional responses and commitmants and concluded that there were no outstanding issues in its review.
In bection 21 of Supplement 3, the staff stated the conclusion that information submitted by the applicant in PSAR Amendments 57 and 59 is sufficient to show compliance with the action items in NUREG-0718, Revision 1.
In this report the staff has found that, for the construction permit stage of review, the applicant is in compliance with the Commission's rule on emergency l
planning that was promulgated siqce Supplement 2 was issued.
The i, sue relating to the applicability of the proposed ACI-359 Code has been resolved as dis, cussed in Section 3.8.2.
The remainder of this report pertains to issues that the staff had previously concluded were not outstanding issues, but for which updating j
of review is appropriate because (1) the applicant's progress in design analyses resulted in significant design changes, (2) staff progress in its generic issues program resulted in significant changes in unresolved safety issues, and (3) staff policy is to update financial reviews when significant time has elapsed.
After considering the results of its updated review, the staff reaffirms its conclusion that none of these items represents outstanding issues in its construction permit review. Therefore, the staff concludes that its review is complete and current with respect to the construction permit stage of licensing requirements and there are no outstanding issues in its review. The staff's conclusions are reaffirmed in Section 21.0 of this supplement.
1.10 Permit Conditions The staff has identified one issue in its review where the construction permit will be conditioned.
This item, postaccident hydrogen control, is discussed further in Section 1.1 of this report.
i i
)
Allens Creek SSER #4 1-3
+
=
l
. 2.0 SITE CHARACTERISTICS 2.2.4 Nearby Industrial Facilities In Section 2.2.4 of' Supplement 2 to the Allens Creek SER, the staff discussed i
pstential hazards ' associated with a 6-in. -liquid petroleum gas (LPG) pipeline
'which, at its closest approach, is about 7000 ft away from the nearest. seismic
' Category I plant structure. The staff concluded-that the applicant's analyses had not demonstrated the assurance recommended by Standard Review Plan Sectior.s i.
2.2.1 and 2.2.2 that a postulated event associated with a break in the LPG-
~ ipeline need not be considered as a design-basis event.
p t
f The staff review. focused on several distinct aspects of the applicant's analyses of the postulated' pipeline rupture including:
(1) the extent of a low-lying I
propine cloud that potentially could be formed below the 140-ft isocline of Allens Creek; (2) the potential for the formation of deflagrable and detonable j
propane clouds above the 140-ft isocline; and (3) the potential thermal fluxes
[
and.overpressures stemming from propane deflagration and detonation, respectively The first concern was to determine the extent of potential propane cloud i
formation along Allens Creek, because this was perceived as a means of bringing propane'in significant-quantities and concentrations close to the plant.
At the request of the staff, the applicant provided a propane-vapor flow analysis -
which included gravity-induced propane flow considerations based on some i
recent experiments with dense (negatively buoyant) gas plumes.
The use of the j
experimental data was suggested because.it supported the view that relatively-little mixing can take place across the horizontal interface between a dense 4
plume and the ambient and more buoyant fluid. This condition would tend.to maximize the transport'of propane along a channel such as Allens Creek, because i
losses to the upper layers 'of-ambient air would be restricted.
The use of j
this correlation led to the result that gravity flow of negatively buoyant
{
propane could extend far enough along Allens Creek'so that it would pars by j
the site.
Th*s estimate is in agreement with the staff's concern noted in SER Supplement 2 with respect to the potential for forming extensive propane clouds as indicated by historical data of LPG pipeline failures.
(
)-
Given the potential for propane flow along Allens Creek and past the site, the staff reviewed the applicant's analysis of propane transport-from Allens Creek toward the plant.
The applicant assumed a 26,000-ft-long line source to
-represent the propane within'Allens Creek from the pipeline break to the 1
t cooling lake.
Assuming 5 percentile meteorology and a wind blowing toward the i
plant, it was estimated that the maximum' distances from the closest point on the line source (1800 ft between the plant and the closest point on Allens i'
Creek) to the lower deflagrable and-detonable concentrations were 220 ft'and 190 ft, respectively.
In other words, the deflagrable propane mixture would approach to within 1580 ft of the' plant and the detonable mixture would come to within 1610 ft'of the plant.
The staff has reviewed the applicant's calcu-l 1ations'and independently verified the parameters used in the calculations'.
Based on the maximum extent of detonable concentrations of propane toward the i
plant and a line source: length of 26,000 ft, the estimated propane. inventory j
J Allens Creek SSER #4 2-1.
4
..*,-~e-,-r, e w w
av
-www--
-.w+-----.-r-w-
- * - -e r
-- *, - -se wa
--m-e-ww--+-e
=ew",
w-eme r C r-- e
,-+-i.as,rr--ec--,w--+*=+---+-wwm-w++o-~=--w-e
i available for detonation is 4.3 x 108 fta.
This is equivalent to 5.4 x 104 lb of TNT.
Assuming that the entire inventory of propane is detonated at the closest point to the plant (that is at 1610 ft from the plant) the resulting peak positive incident overpressure is about 1.1 psi.
The NRC staff position, as stated in Regulatory Guide 1.91, " Evaluation of Explosio n Postulated To Occur on Transportation Routes Near Nuclear Power Plants," Revision 1, February 1978, is that conservatively for peak positive incident overpressure of 1 psi or less, no significant damage to critical plant structures would be expected.
Using this screening criterion, the staff concludes that a plant-specific analysis for Allens Creek is not necessary.
Finally, the potential fire and explosion effects on the plant in the event of cloud ignition were estimated. With respect to fire effects, the estimated 8
j propane inventory within deflagrable limits is 6.1 x 108 ft.
Even if the entire propane inventory available for deflagration were to be located near the closest point to the plant (that is about 1580 ft from the plant), the 2
j maximum thermal flux at the plant would not exceed 31 kW/m.
It would take about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of exposure at this rate before any significant effects were produced on plant structures.
After considering the LPG pipeline hazard in addition to the other potential hazards associated with nearby industrial facilities that were considered in Supplement 2, the staff concludes that the site location is acceptable and meets the relevant requirements of 10 CFR Part 100. This conclusion is based i
on the following:
The applicant has identified potential accidents related to the presence of hazardous materials or activities in the site vicinity which could affect the plant.
From these th3 applicant has selected those accidents which should be considered design-basis events and has provided analyses of their effects on the safety-related features of the plant.
From the analyses, the applicant has demonstrated that the plant is adequately protected and can be designed and operated with an acceptable degree of safety with regard to potential accidents which may occur as the result of the presence of hazardous
'i materials or activities at nearby industrial, military, and transportation facilities.
2.5 Geology and Seismology 2.5.4 Stability of Subsurface Soils (1) Foundation Conditions In Supplement 2 to the SER, the staff concluded that there was not enough assurance that postulated failures of slopes around the ultimate heat sink j
intake structure would not lead to unacceptable blockage of the submerged intake canal.
In subsequent analyses reported in PSAR Amendment 58, the applicant estimated a slope soil movement of less than 4 in. under seismic loading conditions. The staff has independently reviewed the potential for slumping of the slopes around the intake structure under seismic loading and concurs with the applicant's finding that for the safe-shutdown earthquake (SSE) loading, the expected slope deformation would be minor.
In order to positively restrict potential slope soil movement into the intake canal, the applicant has committed to the construction of a concrete retaining wing wall structure at the intake forebay to contain the soil adjacent to the Allens Creek SSER #4 2-2
intake structure.
Such a retaining structure can be readily constructed using standard engineering design and construction principles and procedures, and its presence would provide very high confidence that the flow of cooling water into the intake structure from the lake would not be adversely affected by a postulated failure of the ultimate heat sink causeway slopes.
During the operating license stage of review, the staff will study the final design of the retaining wall structure and its interface with the soil slopes at the ultimate heat sink forebay canal.
Allens Creek SSER #4 2-3
3 3.0 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS 3.8.1 Steel Containment In Supplement 2 to the SER, the staff' stated that the applicant had provided j~
additional information concerr.ing criteria used to design the containment and its interior structures to withstand the. effects of suppression pool dynamic loads resulting from a loss-of-coolant accident and steam relief valve actua-tions and that the details of the structural design criteria ~are described in j
a report prepared for the applicant by EBASCO Services, Inc. entitled "Contain-ment Structures Design Report," Revision 1, and dated July 1977.
The staff reviewed this report and found the structural design criteria to be acceptable.
i By Amendment 54 the applicant submitted Revision 2 to that report and documented a design change from a semiellipsoidal to a hemispherical containment dome.
}
Because the basic structural design criteria are applicable to either configura-t tion and continue to be acceptable, the staff reaffirms its conclusion that the structural design criteria are acceptable.
k
(
In Supplement 2 to the SER, the staff also described the status of its review of the loads and combinations thereof occurring in the suppression pool and their feedback effects in other locations resulting from loss-of-coolant '
i accident, air clearing, steam condensation, chugging, and steam relief valve actuation that will be considered in the design of all structures, systems, and components housed within the reactor building.
The staff stated its i
conclusion, based on its review at that time, that the containment will be j
adequately designed to withstand, in addition to the design loads, the effects of pool dynamic loads.
In Revision 2 of the containment design report, the applicant has updated its loads and combirations of loads in accordance with additional test results and analyses developed by the General Electric Company.
j The staff does not anticipate that the final definition of these loads will be determined until after the prospective decision date on the issuance of a con-i l
struction permit for Allens Creek Unit 1.
However, on the basis of its ongoing generic review of the tests and analyses to determine these loads and its review to date of the applicant's ongoing containment design program, the staff reaffirms its conclusion that the containment will be adequately designed to 1
withstand these loads.
i f
3.8.2 Containment Interior Structures i
In Amendment 54 to the PSAR, the applicant proposed a 33-1/3 percent increase l
in reinforcing steel allowable stress for test conditions applicaM e to concrete containment interior structures and the concrete basemat in which the steel i
containment is anchored.
This proposal is based on provisions contained in the 1975 edition of the ACI-ASME Standard Code for Concrete Reactor Vessels and i
f Containments,Section III, Division C.
Specifically the allowable stress for 1-reinforcement steel would increase from 0.5 to 0.67 times the yield stress j
using the working stress design method. -The current edition of the Code (1980) contains a similar but a more relaxed provision, which permits a 50 percent increase for reinforcing bar tension involving pressure test load conditions.
i Allens Creek SSER #4 3-1 n,
,-,,n.,,, - - - -,,, - - -. ~, -
.,---n
-n,,
- -,,...,,e--
The staff has taken exception to similar code provisions in Standard Review Plan Section 3.8.1 and Section 3.8.3 where the Code-specified 33-1/3 percent increase for wind.and OBE loads is'not accepted, requiring applicants to provide
-justification for any such proposals.
Tha applicant responded to the staff request for justification by letter of June 26, 1981, and in a technical meeting held on.0ctober 23, 1981.
During these discussions the applicant cited (1) the extensive peer review the codes in question received before adoption and (2) experience ~ indicating that the cracking of concrete that has occurred during testing.of nuclear structures designed to similar criteria did not lead to increased leak rates.
The applicant further stated that use of lower allowable stresses for the test condition would require more reinforcing bar in an already crowded configuration, thus increasing the risk of less than optimum concrete placement. The staff noted that the higher allowable reinforcing bar stresses infer somewhat larger allowed concrete cracks during the period of stress, thus increasing the possibility of corrosion effects over the lifetime of the plant.
In the contin'uing development of the professional society codes and regulatory criteria for design and contruction of concrete containments, matters with respect to (1) the level of allowable stresses that is appropriate for nuclear plant design, (2) whether excess conservatism in one area will yield a net loss in safety; i.e., loss of strength because of construction difficulties brought on by crowded reinforcing bar patterns; and (3) whether the increased allowable stress will result in possible future corrosion from concrete cracks larger than those normally experienced will be part of a continuing dialog between the staff and code committees with a goal of modifying the codes, the Standard Review Plan, or both as new inforreation and experience are available.
With regard to the specific acceptability of the applicant's proposed criteria for Allens Creek:
(1) The test condition is but one of a number of design conditions and may or may not control the design of various structural members.
The staff accepts all other design conditions proposed by the applicant.
(2) For designs accomplished in accord with the applicant's proposed criteria (0.67 times yield stress as the allowable under test conditions), it is reasonable to expect that the structural system will remain elastic, i.e.,
there will be no permanent deformation, during the test period. As a con-sequence the structures will return essentially to the pretest conditions with safety margins based on other acceptable load combinations available for response to accident pressures or for other containment functions.
(3) As noted above, the applicant reported that experience to date has not shown any likelihood of increased leakage during pressurization for structures designed to similar criteria.
(4) Prior to operation of the plant the ability of the containment and drywell to withstand the calculated design pressure will be confirmed by a pressure test at 1.15 times the design pressure, and posttest inspection will provide the bases for surveillance programs during operation, if necessary, to
. monitor corrosion effects.
Based on the above, the staff concludes that the applicant's proposal to use the Code-allowed 0.67 times yield stress as a design criterion for reinforcing bar under test conditions is acceptable.
Allens Creek SSER #4 3-2
3.8.3 Design of Other Category I Structures In response to a generic request by the staff, the applicant by letter dated October 19, 1981 addressed staff concerns pertaining to use of masonry walls in seismic Category I structures.
The applicant stated that concrete block walls that will be used in Category I structures are not required to carry loads of the main structure and, therefore, are not classified as Safety-Related, Category I.
Suitable barriers will be provided for protecting the masonry walls against missiles in accordance with Section 3.5 of the Standard Review Plan, " Internally Generated Missiles." The applicant will comply with criteria of Sections 3.7 and 3.8 of the Standard Review Plan that are appli-cable to seismic non-Category I structures.
The applicant described the method that will be used to evaluate in masonry walls dynamic forces that result from earthquakes.
Safety-related piping and equipment will not be attached to masonry walls.
All non-safety-related instrumentation lines, and electrical conduits and boxes, attached to the walls will be included as a part of the dead load.
No masonry walls are situat'1 where they could form pt.rt of a " containment boundary" and be subjected to t. pressurization event.
The applicant provided plan and elevation views of th! plant structures showing the location of all masonry walls.
l On the basis of its review of the information provided by the applicant, the
[
staff concludes that the preliminary design and design criteria are acceptable l
for the construction permit stage of review.
l l
L l
l n.
Allens Creek SSER #4 3-3
4.0 REACTOR 4.2 Fuel System Design In Section 4.2 of SER Supplement 2, the staff described changes in the fuel design that have been made since the SER was issued in November 1974.
In Amendment 56 to the PSAR dated March 9, 1981, the applicant provided a revision of Chapter 4 and other PSAR sections to reflect changes in the fuel design to include prepressurized fuel rods, and two, instead of one, water rods per assembly.
In Table 4.1, characteristics of the modified Allens Creek fuel dar,ign are compared with the characteristics of the fuel design for the LaSalle Station (one of several BWR plants in the operating license stage of review).
The SER for the LaSalle Station was issued in March 1981 (NOREG-0519).
Because the revised fuel design for Allens Creek is essentially identical to the fuel design for the LaSalle Station, the staff concludes on the basis of its review of the fuel design for operation of the LaSalle Station that the fuel design - lifica--
tions for Allens Creek described in Amendment 56 will be acceptable at the operating license stage and are, therefore, acceptable for the construction permit stage of review.
Table 4.1 Comparison of parameters for fuel assembly designs Parameter Allens Creek LaSalle Fuel rods per assembly 62 62 Channel thickness, inches 0.120 0.100 Active fuel length, inches 150*
150*
Uranium xeight per assembly, pounds 402.8 403.4 Rod-to-rod pitch, inches 0.636 0.640 l
Water / fuel ratio (cold) 2.60 2.75 Cladding outside diameter, inches 0.483 0.483 Cladding thickness, inches 0.032 0.032 Thickness / diameter ratio 0.0662 0.0662 Fuel pellet outside diameter, inches 0.410
'0.410 Pellet-clad diametral gap, mills 9
9 Maximum linear heat generation rate, 13.4 13.4 kilowatts per foot l
Maximum fuel temperature, degrees 3435 3435 i
Fahrenheit I
- Includes six inches of natural uranium dioxide at bottom and top of fuel column.
l
~
Allens Creek SSER #4 4-1
b 13.0 CONDUCT OF OPERATIONS 13.3 Emergency Planning The Allens Creek Nuclear Generating Station will be located in southern Austin County, Texas, just west of the Brazos River, about 45 mi west of Houston.
The plume exposure emergency planning zone (EPZ) established for the Allens Creek site is an area 10 mi in radius.
This 10-mi-radius plume exposure EPZ encompasses the 3.5-mi-radius low population zone (LPZ).
It includes portions of five counties: Austin, Fort Bend, Wharton, Colorado, and Waller.
The plume exposure EPZ for Allens Creek is shown in Figure 13.1.
The ingestion pathway EPZ is an area about 50 mi in radius around the site and is located entirely within the State of Texas.
10 CFR S0.43(a) requires each applicant for a construction permit to include in the Preliminary Safety Analysis Report (PSAR) a discussion of preliminary plans for coping with emergencies.
The PSAR must conta'i sufficient information to ensure that the proposed emergency plans for both tL site and the EPZs are compatible with the facility design features, site layouc, and site location with respect to such considerations as access routes, surrounding population distributions, land use, and local jurisdictional boundaries for the EPZs.
The areas which must be addressed in the PSAR to meet the requirements for emergency planning at the CP stage are set forth in Part II of Appendix E to 10 CFR Part 50.
In response to these requirements, the applicant filed PSAR Amendments 55 and 60, dated January 21 and August 21, 1981, respectively.
In the following discussions, the staff has evaluated the applicant's submittals and finds that the requirements of 10 CFR Part 50, Appendix E, Part II--including the requirements for information to indicate the feasibility of meeting the standards of 10 CFR 50.47(b)--are satisfied.
The individual requirements of Appendi" E, Part II (A through H) will be discussed and analyzed below.
13.3.1 Item A Requirement Describe the "onsite and offsite organizations for coping with emergencies and the means for notification, in the event of an emergency, of persons assigned to the emergency organizations."
Discussion The Allens Creek emergency organization will initially consist of the onduty operating staff and will be augmented by offduty plant personnel, designated Houston Lighting & Power Company corporate personnel, and personnel from Federal, state, and local response organizations.
The applicant plans to establish an onsite emergency organization to meet the minimum staffing require-ments for emergencies listed in Table B-1 of NUREG-0654.
The staffing levels will be met either by augmenting the normal operations staff within the time Allens Creek SSER #4 13-1
/
o p> v: prvs _ *
-f a :;/.MK '% 5A i
nin o c.ct r y - ~ f.._. y. q i
i M.
A-
"hg. _.
,s}5 d
i
/
uwe u-
!h
'"'V' "1 I l. *..m t le
-~ ?r4Y ~
- f, p 'Y? '. 7 k
I nw 1
si2
~ ?-
',.. a-fA- ' N C'%
t y,
i'
.; ' % ';- K ~ ~ a*-
1 4
I 3
f' h' i
. ~ '
D
~Y j
WALLER
' /,
7.-
c" a'
-c l
.} 's. A L f >>
,,,,b
~
~
i f
g
(
s(-
)
'j s
s
,k
~
'~
\\
{
f s.
'\\,
\\~-
"=
\\
[l *
?(
(,
)
' /
~
FORT BEND 0
j/
.s
'y COUNTT o
3
\\ ~r' m.
. i y
v COLC': ADO 4
's,
'.,i Q.. ~,,(f 1
I J C011(TT f
r.
~
%y-
'. ji li a%, **d-I s
.3f
. r ~'
?>
,'(-
e m
- s. ~
-g- -.
j
-r f
s
.e j
.5
~'
WHARTON
)
.N N
COUNTT
.c,' Ay.
., j.
s--
M.,g4.,
x.*' :,
'N. );
s
,u.,
ss.= s -
,s M
~.;~.,
1
~,
j j_ - ;' -
j "4
d 'W
, )='5'
.My M, l
\\ 2.,.. ~
~ r Wr+1 t i_ U.
Figure 11.1.
10-mi-radius study area i
i 1
i Allens Creek SSER #4 13-2
i periods specified 17 Table B-1 or by increasing the number of personnel on the normal operations staff.
An oncall system will be used to contact personnel who are to augment the operations staff.
The onsite emergency organization is shown in Figure 13.2.
No constraints have been identified that would prevent meeting the staffing criteria of Table B-1.
The applicant has ide"ified the primary responsibilities of the major elements of the onsite emergency organization in the PSAR.
The onduty Operating Super-visor immediately assumes the position of Emergency Director and is responsible for the initial evaluation of the situation, accident classification, and notification of offsite authorities.
The Operating Supervisor serves as Emergency Director until relieved by the individual designated to serve as Emergency Director through the remainder of accident response activities.
This individual will most likely be the Plant Superintendent.
The Federal agencies available to provide support in the event of an emergency include the NRC, the Federal Emergency Management Agency (FEMA), and the
' Department of Energy (DOE). The resources of other Federal agencies would also be available in a serious emergency situation.
A national radiological emergency response plan defining the role of Federal agencies is presently being developed by FEMA.
In the State of Texas, state response to any type of emergency is coordinated through the Emergency Management Council (EMC), which is composed of representatives from 29 state agencies and is chaired by the Director of the Texas Department of Public Safety (DPS). The Texas Department of Health is represented on the EMC and is the lead state agency for the coordination of state response to a radiological emergency.
In the event of an emergency, the applicant would contact the state agencies through the DPS which maintains dispatchers on duty 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day at its state headquarters and district offices.
The applicant will have dedicated telephone lines and backup radio communication to the DPS.
L Local support agencies include the sheriff's departments of the five counties within the plume exposure EPZ (Austin, Fort Bend, Wharton, Colorado, and Waller) and the police and fire departments of the nearest community, Wallis, l
l Texas. Other local support services include ambulance service and medical i
treatment.
The Austin County Sheriff's Department will be the principal point of contact for notifying local support organizations.
The sheriff's office dispatcher is on duty 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day to receive emergency calls.
The applicant will have a dedicated telephone line, with backup radio communication, to the Austin County Sheriff's Department.
The principal offsite support agencies are listed in Table 13.1, and the interfaces between the onsite and offsite support organizations are illustrated in Figure 13.3.
Conclusion Based on its review, the staff concludes that the information submitted by the applicant is suffi +.nt to meet the requirements of Appendix E, Part II, Item A.
l Allens Creek SSER #4 13-3
RADIOLOGICAL 5
SITE RECOVERY EMERGENCY SUPPORT MANAGER h
MANAGER n
I OFFSITE
}
j 1
h w
I I
ONSITE h
3 i
l
=
1 1
=
3 l
L___
EMERGENCY
__ J l
DIRECTOR I
i I
I l
l l
l OPERATIONS TECHNICAL RADIATION d
SUPPORT OPERATIONS SUPPORT ADMINISTRATIVE PROTECTION CENTER SUPERVISOR CENTER SUPERVISOR
~ ~ ~ ~
~
MANAGER SUPERVISOR SUPERVISOR I
ROL COMMAND ROOM
--__ COMMUNICATIONS STAFF I
I I
I A
A URlW A UR CE AFFAIRS Figure 13.2.
Onsite emergency organization--Allens Creek Unit 1
~
8 NE CORPORATE ME 1A PUBLIC 5
HEADOUARTERS CENTER g
n m
m X
m OPE R ATION AL OTHER SUPPORT
\\
OFFSITE CENTER
/
SUPPORT GROUPS l
.l OTHER TECHNICAL EMERGENCY AUSTIN i
LOCAL CONTROI SUPPORT OPERATIONS COUNTY U
h00M CENTER FACILITY SHERIFF SUPPORT a
GROUPS 1
EMERGENCY N8 OPERATIONS STATE DDC CtNTER AGENCIES HOUSTON AUSilN NRC NHC FEMA FEDERAL RE ON
]
WASHINGTON WASHINGTON AGENCIES i
l 1
Figure 13.3.
Emergency organization functional interface--Allens Creek Unit 1
V
.f
~
j
.)
i Table 13.1 10ffsite support agencies' Austin County Sheriff's Department Fort Bend County Sheriff's Department Waller County Sheriff's Department Colorado County Sheriff's Department Wharton County Sheriff's Department' City'of Wallis Police Department
' City of Wallis Fire Department Austin County Ambulance Department Polly Ryan Memorial Hornital Texas Department of Pubiic Safety Texas Parks and Wildlife Departmer.t Texas Department of Health U.S. Department of Energy, Albuquerque Operations Office U.S. Nuclear Regulatory Commission, Region IV Federal Emergency Management Agency 13.3.2. Item B Requirement Describe the " contacts and arrangements made and documented with local, state, and Federal governmental agencies with responsibility for coping with emergencies, including identification of the principal agencies."
Discussion In PSAR Section 13.3.2, the applicant has identified the principal local, state, and Federal agencies responsible for coping with emergencies at the Allens Creek plant.
In the State of Texas, the local government is responsi-ble for issuing emergency notifications and instructions to the public.
The County Judge is the local official responsible for authorizing emergency _
response and protective'.ctions, which are implemented under the direction of a
the County Sheriff.
The principal local support organization is the sheriff's department of Austin County, the County in which the-Allens Creek plant site is located.
Other local support agencies include the other four county sheriff's departments within the plume exposure EPZ, the police department and volunteer fire department of Wallis, Texas (the nearest community to the site), the Austin County Ambulance Corps,:and the Polly Ryan Mamorial Hospital in Richmond,
. Texas. 'The applicant has been in contact with officials of these organizations, and letters of agreement documenting these contacts and the arrangements made are provided in Appendix 13.3A of the PSAR.
The Texas Department of Health, Division of Occupational Health and Radiation Safety, is the lead state agency for responding to a radiological emergency.
The Department of Public Safety and the Parks and kildlife Department also have a direct' response role in the event of an emergency at the Allens Creek plant.
These three agencies are members of the Texas Emergency Management Council (EMC), an organization composed of representatives from.
-29 state agencies, which is responsible for coordinating the state response to Allens Creek'SSER #4 13-6
=
fany type of_ emergency-in the state..The applicant has been.in contact with
'the-three primary _ state response l agencies and has documented the state support in,a. letter of agreement with'the-Department'of Health, which.is the lead
- state agency.and member of-the EMC.
The specific response dutief of state and l:
local; agencies will.be presented in'the-state and-local. emergency plans before commercial operation of the.Allens Creek plant.
On the Federal. level, the-applicant has been in contact with the NRC; DOE, and FEMA.
Letters of agreement / understanding with the NRC' Region IV office and the Albuquerque Operations Office of. DOE are-provided in Appendix 13.3A of~the
.PSAR.
Conclusion' The information in PSAR Section 13.3.2, including the _ letters of agreement in P_SAR Appendix 13.3A, demonstrates that preliminary contacts and agreements have been made with the principal offsite support agencies.
The staff, therefore, concludes that the requirements of Appendix E, Part II, Item B are satisfied.
13.3.3 Item C Requirement
' Describe " protective measures to be taken within the site boundary and within-each EPZ to protect health and safety in the event of an accident;-p*ocedures by which these measures are to be carried out (e.g., in the case of an evacuation who authorizes an~ evacuation, how the public is to be notifiad and instructed, how'the evacuation is to be carried out); and the expected response of off-site agencies in the event of an emergency."
. Discussion
~ Emergency conditions will be classified by the applicant into four standard emergency classes which will cover the entire spectrum of probable and postulated accidents. The~four classes are:
Notification of Unusual Event, Alert, Site l
-Area = Emergency, and General Emergency.' State and local' emergency plans will l
utilize the same emergency classification system.
The Notification of Unusual-Event and Alert classes.are intended to provide early ~and' prompt notification -
to the onsite and offsite emergency response organizations that minor events-have occurred or are in progress which could lead to more serious consequences, if'there is a future degradation in plant status, or which might'be indicative of more serious conditions which are not yet fully ~ realized.
The' Site Area and General Emergency classes are intended for more severe situations,.where some significant releases are likely or are occurring and require.immediate action
-from both onsite 'nd offsite emergency response organizations.
The applicant a
will develop emergency action level (EAL) criteria for classifying emergencies
'in accordance with the guidw of. Appendix 1 to HUREG-0654.
EALs relate to particular inplant condit.c. instrument readings, and onsite and offsite monitoring.results which pruside the basis for categorizing the event into one of the four emergency classes.
-Allens Creek SSER #4--
13-7
=,
Onsite protective measures will include exposure control, contamination control, and area'and site evacuation.
The primary protective measure for onsite personnel not engaged in emergency functions will be prompt evacuation from areas which may be affected.
During emergency conditions, efforts will be made to keep exposures within the limits specified in 10 CFR Part 20 through the use of such measures as respiratory protective equipment, protective clothing, radioprotective drugs, or other health physics procedures.
Exposure criteria for emergency workers performing critical corrective or lifesaving actions will be developed consistent with Environmental Protection Agency-(EPA) Emergency Worker and Lifesaving Action Protective Action Guides.
Emergency planning will include provisions for the prompt notification of appropriate state, local, and Federal response organizations.
For all emergency situations, the applicant will initially notify officials of the State of Texas and the Austin County Sheriff's Department.
The state response would be coordinated by the Texas Department of Health, the lead state agency for radiological emergency situations.
Support activities of the Department of Health will include environmental monitoring, independent evaluation of radio-logical consequences, and the recommendation of protective actions. The Department of Health will also ensure the activation of appropriate member agencies of the Texas Emergency Management Council.
As noted above, the County Judge has the statutory responsibility for authorizing emergency operations within each county; these operations then are implemented under the direction of the County Sheriff.
The Austin County Sheriff's Department will initiate the implementation of protective measures within the 10-mi-radius plume exposure EPZ, based on recommendations from the ap,'icant and the Texas Department of Health.
Response duties of the Sheriff's Department will include notification of the public, traffic control, law enforcement, and communication with other local support organizations.
The principal protective measures to be considered within the 10 mi-radius plume exposure EPZ will be sheltering and evacuation.
Sheltering will be recommended for emergency situations where relatively low doses are involved or where rapid passage of a radioactive cloud is expected.
The primary means of evacuation within the plume exposure EPZ sili be by private automobile and school beses.
The applicant has provided a preliminary analysis (in Appendix 13.3 8 of the PSAR) of the time required to evacuate various sectors and distances within the plume exposure EPZ.
The evacuation study is discussed in the response to Item G below.
Within the 50-mi-radius ingestion exposure EPZ, the principal protective measures will involve the control of food and water supplies.
To demonstrate compliance with the requirements of 10 CFR Part 50, Appendix E, the applicant has committed to meet the criteria in Appendix 3 of NUREG-0654 regarding a public notification system; that is, to provide a system that will give both an alert signal and an informatienal or instructional message to the population on an areawide basis throug'iout the 10 mi-radius EPZ within 15 minutes.
Allens Creek SSER #4 13-8
An evaluation will be made to determine the specific public notification system tu be installed.
The final system is expected to consist of a combi-nation of alert devices such as sirens in areas of concentrated population and individual alert devices such as tone-alert radios or multiple telephone callup systems in areas of low population density.
The staff considers the employment of some combination of these alert duices to be a feasible approach to meeting the public notification criteria of Appendix 3 of NUREG-0654.
The applicant will establish a program for disseminating information to members of the public within the plume exposure EPZ about how they will be notified and what their actions should be in an emergency.
This information will be brought to public attention by such means as direct mailing, advertisements in local telephone directories, and posting in public places.
Visitors to the Allens Creek Lake and State Park will receive information as they enter the Park describing how they would be notified and what actions they should take in the event of an emergency.
Conclusion Based on a review of the information submitted by the al.clicant, the staff concludes that the requirements of Appendix E, Part II, Item C are satisfied.
13.3.4 Items D and E Requirement D Describe " features of the facility to be provided for onsite emergency first aid and decontamination and for emergency transportation of onsite individuals, to offsite treatment facilities."
Requirement E t
Describe " provisions to be made for emergency treatment at offsite facilities of individuals injured as a result of licensed activities."
Discussion - Items D and E A first-aid room with equipment and supplies appropriate for a major industrial facility will be provided at the plant.
Personnel decontamination facilities including showers and sinks that drain to the radiological waste processing system will also be provided.
Individuals on the plant staff trained in advanced first aid and decontamination methods.will be available onsite to respond to emergency situations.
Emergency planning will include provisions for the treatment at offsite facilities of personnel injured on site.
Offsite treatment facilities and personnel will be prepared to handle contaminated patients.
Preliminary arrangements have been made with the Austin County Ambulance Corps and the Polly Ryan Memorial Hospital in Richmond, Texas (as indicated in letters of agreement in PSAR Appendix 13.3A) for transporting and treating injured persons, including those who have been involved in radioactive contamination.
Similar arrangements will be made with a backup hospital.
The applicant will ensure that equipment and supplies for contamination control and personnel
.Allens Creek SSER #4 13-9
r-t decontamination are available and maintained at each of the offsite hospitals.
The medical treatment and ambulance services personnel will participate in emergency drills and exercises with the applicant.
Conclusion - Items D and E 2
The staff has ruiewed the information presented in the PSAR on emergency treatment facilities, both onsite and offsite, and concludes that the require-ments of Append'x E, Part II, Items D and E are satisfied.
13.3.5 Item F.
Requirement Describe the " provisions for a training program for emoloyees of the licensee,
.ncluding those who are assigned specific authority and responsibility in the event of an emergency, and for other persons who are not employees of the licensee but whose assistance may be needed in the event of a radiological emergency."
Discussion As discussed in PSAR Section 13.3.10, the applicant will establish an emergency response training program for the plant staff, headquarters support personnel, and local support services personnel.
Members of the plant staff and head-quarters support personnel will receive training in their specific response functions in the emergency organization.
Individuals who will be assigned specific positions of authority and responsibility in the emergency response organization, such as Emergency Director and Recovery Manager, will receive training in all aspects of the Emergency Plan and implementing procedures.
Persons working at the plant but not directly involved in plant operations will receive general eraployee training on such subjects as warning signals, assembly areas, and evacuation procedures.
Training for local offsite response personnel including attendants at the Allens Creek Lake and State Park will include an overview of the Emergency Plan and detailed instructions in the specific fur:tions each organization will be expected to perform.
Personnel in the state response organizations will receive training through the Texas Radiological Response Interagency Training Committee.
The membership of the committee will include the Director of Occupational Health and Radiation Control and a representative from each utility operating a nuclear facility within the state.
Periodic exercises and drills will be conducted to evaluate the capabilities of emergency response 4
organizations and to develop and maintain individual skills.
Conclusion The staff concludes that the applicant's training program for onsite and offsite personnel as described in the PSAR meets the requirements of Appendix E, Part II, Item F.
Allens Creek SSER #4 13-10
i 13.3.6 Item G Wequirement Describe "a preliminary analysis that projects the time and means to be employe.
in the notification of state and local governments and the public in the event of an emergency.
A nuclear power plant applicant shall perform a preliminary analysis of the time required to evacuate various sectors and distances within the plume exposure pathway EPZ for transient and permanent populations, noting major impediments to the evacuatien or taking of protective actions."
Diccussion The applicant will have direct communication links, suco as dedicated telephone lines and radios, to notify state and local officials.
The principal agencies to be initially notified are the Texas Department of Public Safety and the Austin County Sher?ff's Department.
Both of these organizations maintain 24-hour onduty dispatchers.
Backup communication systems will be in place to ensure notification.
The Department of Public Safety will in turn no+ify other state agencies, in particular the Department of Health, which is the lead state agency for coordinating state response to a radiological emergency situation.
The Austin County Sheriff's Department will alert the other local response organizations within the plume exposure EPZ.
The staff will require the licensee to have the capability to notify responsible state and local agencies within 15 minutcs after declaring an emergency.
As noted in the response to Item C, the applicant will ensure that a public notification system will be installed which meets the criteria of Appendix 3 of NUREG-0654 (a system that will provide both an alert signal and an informa-tional or instructional message to the population on an areawide basis through-out the 10-mi-radius EPZ within 15 minutes).
A combination of alerting devices such as sirens, tone-alert radios, or multiple telephone callup systems is being considered.
The applicant has performed a preliminary analysis of the time required to evacuate various sectors and distances within the plume exposure EPZ based on the permanent and transient populations projected for the year 1990 and the highway network as it existed in 1980.
Evacuation would be accomplished primarily by automobile, and school buses would be used to evacuate students.
Normal and adverse weather conditions were considered in the analysis; adverse weather is defined as severe thunderstorms or fog which would reduce visibility and lower driving speeds.
The 10-mi-radius study area was divided into subareas on the basis of geographical, meteorological, and jurisdictional considerations.
A computer model was used to simulate the evacuation scenarios.
The evacuation time estimates ranged from 15 minutes to evacuate the plant staff and permanent population within 2 mi in good weather to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 45 minutes to evacuate the peak population from the 10-mi-radius plume exposure EPZ unc er adverse weather conditions.
No major impediments tc the evacuation or to taking protective actions were identified in the evacuation study.
The staff has reviewed the applicant's preliminary analysis of evacuation time estimates and finds that the information presented and methodology employed were in-accordance'with the guidance provided by Appendix 4 of NUREG-0654.
A Allens Creek SSER #4 13-11
l f
preliminary analysis of the evacuation time estimates,- based on the:information presented by the applicant, has been performed by T. Urbanik II, of Texas A&M
' University (under contract to Battelle Pacific Northwest Laboratories).
Based cn this analysis the staff finds that the results obtained by the applicant are i
reasonable.
_ _ _In response to the-staff's request of May 21, 1981, the Federal Emergency
[
i Management Agency (FEMA) provided its assessment of the status of state and local emargency preparedness for the Allens Creek Nuclear Generating Station in memoranda dated August 17,-1981 and October _8, 1981 (Appendix E).
FEMA j
suggested in.its August 17, 1981 memorandum that " seasonal" flooding and heavy vacation-related auto traffic on interstate Route 10 presented no special problems.
In its memorandum of October 8, 1981, FEMA stated that there were sufficient alternate evacuation routes to circumvent potential problems of-seasonal flooding.
Also, FEMA stated that transient auto traffic.from vaca-a i
tioners will be considered when the emergency plans are prepared.
FEMA than 4
stated that it had determined that there is reasonable assurance that there are no unusual problems-that cannot be adequately handled as the site-specific emergency plans are developed.
l l
Conclusion Based on a review of the information presented in the PSAR and the independent _- ' '
analysis by the staff consultant, the staff concludes that the requirements of-j j
Appendix E, Part II, Item G are satisfied.
i F
13.3.7 Item H Requirement 1
l Describe "a preliminary analysis reflecting the need to include facilities, i-1 systems, and methods of identifying the degree of seriousness and potential i
j scope of radiological consequences of emergency situations within and outside i
the site boundary, including capabilities for dose projection using real-time meteorological information and for dispatch of radiological monitoring teams-j within the EPZs; and a preliminary analysis reflecting the role of the onsite technical support center and of the near-site emergency operations facility in 1
assessing information, recommending protective action, and disseminating j-information to the public."
3 J
Discussion l_
.The applicant has analyzed the requirements for emergency planning and will 1
establish (as described in PSAR Secton 13.3) systems, equipment, facilities,-
and procedures to identify and assess the potential radiological consequen::es of emergency situations within and outside the site' boundary.
The applicant 4
will develop a standard emergency classification and action level scheme based -
1 on particular inplant conditions, instrument readings, and onsite and offsite monitoring results (see'the response'to Item C for further discussion of.
R i
energancy classes).
The applicant will have the capability and resources to
[
pro' vide initial values, accident evaluation, and continuous assessment through-out the~ course of the accident.
In the event of an actual orfsuspected release of radioactivitv,2onsite and offsite monitoring teams will be-dispatched to-l
. perform direct' radiation measurements and obtain samples.
1 Allens Creek SSER #4 13-12
(
.. -,, ~
The applicant will maini.ain a continuous onsite meteorological measurements program. The program will include equipment and systems to obtain the real-time meteorological parameters necessary for determining atmospheric dispersion conditions.
Plant pers;.nal will utilize the meteorological data and radio-logical monitoring data to develop dose projections.
The final Emegency Plan will include specific provisions for recommending protective actions to state and local organizations based on predetermined dose guidelines.
l Emergency facilities will be established at or near the site for assessing emergency situations, directing response and recovery efforts, mitigating.
accident consequences, and inforning the public.
These facilities will include an onsite Technical Support Center (TSC), an onsite Operations Support Center (OSC), and an offsite Emergency Operations Facility (E0F).
The TSC will pro-vide a location for pbnt management and technical support personnel to function in support of reactor coerating personnel during emergency conditions.
The OSC will serve as an assenit,1y area separate from the control room and TSC for per-sonnel who will support station emergency response operations.
The EOF will serve as a center for the management of overall emergency response operations, 1
including the coordination of response activities with Federal, state, and local agencies.
The Katy Service Center, a facility owned by the applicant approximately 19 mi from the site, has been selected as the preliminary location l
of the E0F.
The staff has reviewed the proposed emergency response facilities for the allens Creek plant and found them to nieet the requirements. of NUREG-0718, Revision 1, " Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License," and to be acceptable for the construction l
permit stage of review.
[
Conclusion The staff concludes that the information submitted by the applicant is sufficient to meet the requirements of Appendix E, Part II, Item H.
13.3.8 Summary Based on its review of Items A through H described above, the staff concludes that the Allens Creek PSAR contains sufficient information to ensure that the proposed emergency plans for both onsite areas and the EPZs are compatible with facility design features, site layout, and site location with respect to such considerations as access routes, surrounding population distributions, land use, and local jurisdictional boundaries.
10 CFR 50.47(b) lists 16 planning sta1dards that must be met in the emergency response planning for a nuciear power reactor.
Specific criteria for these standards are contained in NUREG-0654.
The applicant has responded to each of these standards in PSAR Section 13.3.
The staff has reviewed the information on emergency planning in the PSAR and concludes that the information presented is~ sufficient in depth and scope for the construction permit stage to indicate the feasibility of meeting the planning standards in the final Emergency Plan.
l Further, no special or unique circum:,tances have been identified which would preclude the development of adequate emergency preparedness plans at the l
operating license stage of review.
Allens Creek SSER #4 13-13 l
Based on its assessment of the applicant's emergency plans and the FEMA findings regarding the status of offsite emergency planning, the staff concludes that the overall state of emergency preparedness for the Allens Creek site is acceptable for the construction permit stage o# review.
1 J
I a
Allens Creek'SSER #4 13-14
15.0 ACCIDENT ANALYSES 15.2 Abnormal Operational Transients In SER Supplement 2 the staff stated that the applicant had committed to incorporate any design modifications that may be required by the staff to resolve the anticipated transients without scram (ATWS) issue, and that such resolution may require implementing some or all of the plant modifications discussed in Volume 3 of NUREG-0460, dated December 1978.
Subsequently, the plant modifications that may be required were modified as discussed in Volume 4 of NUREG-0460, dated March 1980.
That report identified basic plant layout, diesel capacity, and seismically qualified structures in constructed plants as constraints to implementation.
By letter of September 15, 1981, the staff asked the applicant (1) to demonstrate that the preliminary design of the Allens Creek plant does not contain any such constraints or (2) to modify the preliminary design to eliminate such constraints that might now exist.
By letter dated October 28, 1981, the applicant stated that no such constraints exist in the preliminary design of the Allens Creek plant.
On this basis, the staff concludes that potential plant modifications identified in Volume 4 of NUREG-0460 (including Alternative 4A) can be fully incorporated into the Allens Creek plant even if construction is completed.
The staff finds this lack of constraint acceptable at the construction permit stage of review and concludes that there is reasonable assurance that future requirements of the Commission for ATWS will be fully implemented in the Allens Creek plant.
Allens Creek SSER #4 15-1
n l
20.0 FINANCIAL QUALIFICATIONS 20.1 Introduction
~An evaluation of Houston Lighting & Power Company's (HL&P) financial qualifi-cations to design and construct Allens Creek Nuclear Generating Stalion Unit 1 appeared in SER Supplement 2, dated March 1979.
Because of the time elapsed since the preparation of that report, a complete update of HL&P's financial qualifications is presented here.
By letter of August 12, 1981, itL&P updated the financial information that had been provided in Amendment 2 to the license application, da a d June 21, 1978.
The Commission regulations relating to the determination of the financial qualifications of an applicant for a facility construction permit are given in Section 50.33(f) of 10 CFR Part 50 and Appendix C to 10 CFR Part 50.
In accordance with these requirements, the staff evaluates whether there is reasonable auurance that an applicant can obtain the necessary funds to cover its portion of the estimated construction and related fuel-cycle costs for the proposed facility.
In the case of Allens Creek, the staff's evaluation of the financial qualifications of the applicant included consideration of the Commis-sion's decision Public Service Company of New Hampshire, et al., 7 NRC 1, at 18
-(1978) (Seabrook Station, Units 1 and 2) which states:
... the applicant must have a reasonable financing plan in light of relevant circumstances."
The Commission indicated that an applicant possessing such a financing plan satisfies the reasonable assurance requirement of the regulations.
20.2 Construction Cost Estimates The most recent estimate of Allens Creek construction costs provided by the applicant in its letter of August 12, 1981 are
($ Millions)
Nuclear production plant 2707 Transmission, distribution, and general plant 41 Nuclear fuel for first core 122 These cost estimates are based on an estimated start of construction date of April 1982 and an expected start of commercial operation in March 1991.
The estimated cost of the nuclear production plant nas been reviewed in compar-ison with the cost projected by DOE's CONCEPT capital cost model, as performed by the Oak Ridge National Laboratory. This model projected the cost of the nuclear production plant for Allens Creek Unit 1 to ba $2402 nillion compared with the applicant's estimate of $2707 million.
Because the CONCEPT estimate is used as caly a rough check on an applicant's estimate and is not intended as a substitute for detailed engineering cost estimates, the staff has determined that it is reasonable to use the applicant's estimate for this analysis.
Moreover, because of its higher capital cost projection, the applicant's estimate is inherently more conservative.
Allens Creek SSER #4 20-1
~
i 20.3 Bases for Analysis Consistent with the regulations discussed above, the staff requires that investor-owned utility applicants submit pro forma statements of sources and
]
use of funds with underlying' assumptions.
In general terms, these statements From the viewpoint of use of funds, a are best described as financial plans.
financial plan shows year-to year funds requirements for systemwide construction
. projected throughout the period of construction for a subject nuclear facility.
At the same time,'a financial plan also shows sources of funds or, stated l
simply, where the required capital is coming from.
Generally, sources of funds' for a public-utility consist of short-term borrowings, internal cash generation, and proceeds from additional sales of long-term-debt, preferred-j-
stock, and common-equity securities. -From-this perspective, and in considera-tion of important underlying assumptions to the financing plan, the staff determines the impact of this financing on significant financial parameters.
In this respect,'the reasonableness of an applicant's financial projections is determined.
j J
This reasonable assurance standard, however, must be viewed in light of the extended period of time from the start of construction to full commercial operation.
It is presently estimated that Allens Creek Unit 1 will commence commercial operation in approximately 9 to 10 years.
Consequently, one must necessarily make certain assumption 1 about future conditions.
Two fundamental assumptions that have been incorporated in the analysis of the applicant's l
projected financing are (1) that there will be a rational regulatory environment in the setting of rates for utility service and (2)'that viable capital markets 1
will exist.
The former assumption implies that rates will be set to at least cover the cost of service, including the cost of capital; the latter assumption implies that capital will be available at some cost.
The staff also utilized the following element of conservatism from the perspective of source of funds.
Because other construction is generally planned by an t
applicant during the period of construction of a subject nuclear facility, 1
expenditures required by the other construction increase the applicant's requirements for capital.
Moreover, redemptions required by maturity of an 4
applicant's outstanding debt over the period of construction further increase i
the applicant's requirements for capital.
Because total capital requirements for any given year are higher than the expenditures needed to construct the i
subject facility, use of total _ capital requirements as a basis of analysis is a more conservative approach.
~ Tables 20.1 and 20.2 show HL&P's financing plan and underlying assumptions, respectively.
The staff's evaluation, which follows, is an analysis of the l
central aspects of the plan and the assumptions.
l 0.4 Rate of Return on Common Equity i
Of all the factors considered during the review of an investor-owned utility applicant's financial projections in determination of financial qualifications, the assumptions of projected rates of return on common equity during the
-period ~of construction are most significant.
Rate of return on common equity is best described as earnings stated as a percentage of all the stockholders' j
20-2 Allens Creek SSER #4-w
nGi 3
Table 20.1. Pro forma sources of funds for systemwide constructiort* expenditures and capital structures during cw.struction.of Allens' Creek Unit 'l g-($ Millions) 5 Q
1981 1982 1983 1984 1985 1986 1987 1988 1989 1990.
1991 4
ta External financina:
F 7ommon stock.
214 116 217 221 242 247 249 230 128 226 128 (n
Feeferred stock Se 110 110 90 110 110 90 70 110 100 Ltng-term debt 30::
300 300 525 650 600 450 450 450 275 450
- n
- No,e5 payaule (97) 11 43 30 54 151 161 (37)
(17) 67 121 Con cibution from parent A
id al external funds 77 537 670 866 1056.
1108 950 713 671 569 799 Internally generated funds:
Net income 244 336 363 448 500 641 633 709 770 738 847 Less:
Preferred dividends 21 34 48 56 67 7F 86 94 103 109 114 Common dividends 143 181 217 166 306 h.
413 454 492 529 566 Retained earnings 76 121 98 126 127 20',
134 161 175 100 167 Deferred taxes 46 48 55 59 66 65 72 13 84 99 114 Investment tax credit Deferred net 52 51 69 73 99 107 105 83 67 70 82 Depreciation and amortization 117 125 144 158 184 199 250 198 351 448 m
Change in working capital 41 10 19 27 32 45 37 57 26 15 18 O
.Less AF00-(70) 109 (91)
(140)
(149)
(197)
(130)
(40)
(113)
(50)
(124)
O Total internal funds 262 246 294 303 359 42<
468 632 590 682 795 Total funds 729 783 964 1169 1415 1532 1418 1345 1261 1250 1594' Construction expenditures:
Nuclear power plants 151 198 260 295 327 334 255 212 148 109 39 Other 558 585 704 874 1018 1168 1123 1133 1058 1141:
1430 Total construction N.
expenditures 709 783 964 1169 1345 1502 1378 1345 1206
-1250 1469 Subject power plant -
~55 TI7 IBU W
753
~743 W
~717 T3B TUD-7 Other capital requirements:
'4 Redemption of maturing bonds 20 70 30 40 55 125 Acquisition of bonds for sinking funds Miscellaneous requirements Total capital requirements 729 783 964 1169 1415 1532 1418 1345 1261 1250 1594 Capital structure [%)
Long-term debt 1867 [48] 2167 [47] 2467 [47] 2992 [48] 3572 [49] 4142 [49] 4552 [49] 5002 [49] 5397 [49] 5672 [49] 5997 [48]
Preferred stock 294 [ 7] 404 [ 9]
514 (10] 604 [10]
714 [10] 824 [10] 914 [10] 984 [10]
1094 [10] 1094 [ 9] 1194 [10]
Common equity 1750 [45] 1987 [44] 2302 [43] 2649 [42] 3017 [41] 3469 [41] 3851 [41] 4;.. [41] 4543 [41] 4869 [42] 5164 [42]
Total 3911[100] 4558[100] 5283[100] 6245[100] 7303[100] 8435[100] 9317[100] 10227[100) 11 9 4[100] 11635[100] 12355[100]
t.
t equity accounts, such as capital stock, premiums, and retained earnings in a corporation.
This is derived by first deducting from gross operating revenues the company's operation and maintenance expenses, depreciation, interest charges, taxes, and preferred dividends.
This computation results in net incoma available to the common stockholder, the " bottom line" of a company's operations.
Dividing this by the total of investment dallars provided by the company's common stockholders and accumulated retained earnings results in j
per-unit return on common equity.
R wtated on a percentage basis, this translates into the rate of return on common equity.
Table 20.2 Assumptions upon which the source of funds statement is based 1.
Rate of return on average common equity:
Targeted return on common equity of 15.8% (return allowed by Public Utility Commission in Docket 3320, September 1980).
2.
Preferred stock dividend rate:
1981-12.5%; 1982-11.7%; 1983-10.8%;
1984-91-10%.
3.
Long-and short-term interest rates:
First mortgage bonds:
1981-13.5%; 1982-12.7%; 1983-11.8%; 1984-91-11%
Pollution control bonds:
1981-10%; 1982-9.5%; 1983-91-9%
Short-term debt:
1981-14.8%; 1982-13.2%; 1983-11.6%; 1984-91-10%.
4.
Market / book ratio of projected common stock offerings:
1981-85%;
1982-90%; 1983-95%; 1984-91-105%.
5.
Common stock dividend payout ratio-10% of prior year book value by 1984.
6.
Target and year-by year capital structure:
Debt not over 50%;
preferred not over 10%; common not over 45%.
7.
Resultant SEC coverages over the period of construction 1981 3.63 1987 3.45 1982 3.99 1988 3.59 1983 3.79 1989 3.50 1984 3.80 1990 3.29 1985 3.57 1991 3.44 1986 3.72 8.
Annual growth rate in kWh sales and price per kWh:
kWh growth rate = 3.23% co.npounded Price growth rate = 13.88% compounded Of all uvestors providing capital (proceeds of long-and short-term debt, preferred stock, and common stock) to a company, shareholders of common stock bear the highest risk.
Although capital costs attributable to a company by debt and preferred stock are fixed by contract, and must be paid at the agreed Allens Creek SSER #4 20-4
rate, t. hose dollars earned on common equity represent whatever remains after j
4 payment of all other charges and expenses.
By reason of its inherent risk, because holders of a company's common stock bear the lowest priority of payment to all other obligations of that company, rate of return on common equity represents the best indicator of a company's profitability.
Profitability is
'important in that it affects both interest coverage and the price of a company's securities, which bear on the company's ability to successfully market its
-securities and maintain the formation of a reasonable ctpital structure.
It is important to note that in the case of HL&P, its " common shareholders" are in the form of its holding company, Houston Industries, Inc. (HI), which purchases all of HL&P's common stock.
In turn HI issues common stock to the public and re:nvests primarily in its principal subsidiary, HL&P.
Thus, the actual, primary financial strength of HI derives from its electric energy operation conducted by HL&P.
HL&P accounts for the vast majority of the earnings and assets of HI. Through other subsidiaries, HI is engaged in oil and gas exploration and in the acquisition and delivery of fuels to electric J
generating plar us.
Because the applicant is a public utility afforded monopoly status in its area of service, it is subject to regulation. Accordingly, its rate of return and rates are set by the Public Utilities Comission of Texas and by the incorporated municipalities in HL&P's service area. Unlike the utility base rates, which are fixed, the rate of return on comon equity is only allowed to be earned and is not guaranteed. Although the concept of a fair rate of return on property used and useful in public utility service is deeply ingrained in public utility regulatory law rnd economics, there still exists no absolute certainty as to a utility's future earnings.
Consequently, one is required to consider its current level of profitability and other relevant circumstances in assessing the reasonableness of a projected return on common equity.
The staff has reviewed the assumed rates of return of HL&P's financing plan and has determined them to be reasonable (See Table 20.2 item 1).
In fact, they are the same as the return allowed the company in September 1980 (15.8 percent).
HL&P's actual earned rates of return on average common equity for the years 1978, 1979, and 1980 were 12.7 percent, 13.1 percent, and 13.4 percent, respectively.
It is common for the earned rates of return to be less than the allowed rates of return (15.8 percent in this case).
20.5 Internal Cash Generation In the meeting of an applicant's year-by year onstruction expenditures, the first item considered is the ?evel of internal cash generation.
This is because internal cash generacion reduces the level of external financing required.
By reason of certain noncash expenses (primarily depreciation and deferred income taxes) and the portion of retained earnings not attributable to allowance for funds used during construction, a company may generate funds internally.
To show an example in a simplified fashion, a company is allowed depreciation of its assets.
These amounts are reflected on the company's income statement as an expense.
However, because these funds are not dis-bursed, the company may use them for its own needs.
These dollars represent funds that the company.can apply to its capital requirements, thereby reducing Allens Creek SSER #4 20-5
its need for externally obtained funds.
As another example, when a company earns a profit, it shares that profit with its stockholders in two ways:
first, it takes some of its net income and distributes that portion to its shareholders in the form of dividends; second, after its dividends have been disbursed, the company keeps the balance of its net income and adds this amount to its retained earnings account.
Again, this represents additional funds available to the company for its capital needs.
As an incidental point, although the allowance for funds used during construction portion of earnings is not an immediate source of cash to a company, investors do recognize it as a future source of cash, because when the facility is ultimately placed into rate base (property used and useful in public utility service), it generates funds through both earnings and depreciation.
At the same time, retained earnings also benefit the shareholders in that these amounts increase the worth of their irvestment and further enable the company to grow.
The overall level of a company's internal cash generation is likewise significant to shareholders in that it provides cash coverages to dividends.
This is especially important to investors in public utility common stocks, who generally own such securities because of their income characteristics.
The continuing generation of a sufficient amount of cash flow by a utility instills a higher level of confidence in the payment of future dividends in its common stock shareholders.
This is beneficial to the company as, in part, it continues to maintain the attractiveness of its equity securities.
In calendar years 1978, 1979, and 1980, HL&P generated internally $180 million,
$198 million, and $233 million, respectively.
Projected internally generated funds compare reasonably with tnis historical experience.
For the purpose of this analysis, HL&P assumes internally generated funds in the years 1981, 1982, and 1983 amounting to $262 million, $246 million, and $294 million respectively (see Table 20.]).
The assumed increases are partially justified by planned rate increase requests.
HL&P currently has a $248 million request pending before the Public Utility Commission of Texas.
Increases in depreciation also serve to increase internally generated funds.
20.6 Interest Coverage To meet its capital requirements during the construction of Allens Creek Unit 1, HL&P will, from time to time, enter the market for the sale of long-term-debt securities.
These securities are mortgage bonds that are secured with a lien on the assets of the issuer.
To protect the assets mortgaged under a company's debt, a trust indenture agreement is made between the company and the bondholders Indentures of such mortgage bonds contain provisions that, in addition to protecting the assets mortgaged, the interest due to the bondholders is also cevered.
At the same time, to provide an adequate level of earnings cushion over and above the company's interest requirements, there generally exists in such mortgage and trust deed indentures an interest coverage test.
Inextricably related to earnings and interest charges, this provision precludes the company from issuing additional debt should there not be satisfactory earnings coverage over its interest obligations.
Because of its significance, the interest coverage ratio is a major criterion used by the financial community in making credit _ decisions with_ respect to a company's debt.
-Allens Creek SSER #4 20-6 i
The staff has reviewed the interest coverage assumptions underlying HL&P's financing plan and finds them to be reasoncble (1) when compared to the historical coverages of HL&P and'(2) in consideration of the projected debt issuances of the company. The company projects interest coverage above 3.0 through the period of construction.
It has achieved interest coverages above 3.0 for each of the years 1978, 1979, and 1980.
This is substantially in excess of the minimum 2.0 coverage required by the company's indentures.
Under such indenture restrictions HL&P could have issued in excess of $700 million in mortgage bonds as of May 1981.
20.7 Capital Structure l
For a company to conduct a viable financing plan and preserve the attractive m s of its securities, it must maintain a reasonably balanced capital structure.
The term " capital structure" refers to the c uposition of a company's capitaliza-tion, that is, the proportion of debt equity and preferred stock which constitute capitalization.
Capital structure is an important consideration in corporate financial analysis in that it shows how much equity capital is available to 1
protect the senior obligations, or in other words, to what extent are the owners using their own capital and tn what extent are they relying on creditors' money.
By maintaining a reasonable and well-balanced capital structure, latitude will exist in a company's options of financing.
This will help achieve borrowing reserve, allowing flexibility both in the timing and selection of securities to be issued to meet capital requirements.
Most important, under these cir-cumstances, its securities will maintain their attractiveness to investors by virtue of their lower risk, because capital structure affects interest coverage.
Generally speaking, investor-owned clectric utilities historically have had capital structures composed of 50 to 55 percent long-term debt, 10 to 15 percent preferred stock, and 35 to 40 percent common equity.
These ranges of capital structure are considered reasonable by the financial community in that they maintain a sufficient amount of equity capital protection to the senior security holders and, from this viewpoint, help protect the attractiveness of the securities.
The staff has reviewed the projected capital structure assumptions under'ying HL&P's financing plan (see Tables 20.1 and 20.2) and finds them to be reasonable when compared with historical experience of the company and industry norms.
For the years 1978 through 1980, the company's capital structure was as follows:
long-term debt, 48-53 percent; preferred stock, 8 percent; and common equity, 39 to 44 percent.
As indicated in Tables 20.1 and 20.2, HL&P projects capital structures. in these approximate ranges which are favorable for an investor-owned electric utility.
20.8 Conclusion Based on the preceding analysis, the staff concludes that HL&P has presented a reasonable financing plan in the light of relevant circumstances. Therefore, according to the provisions of 10 CFR Part 50, Section 50.33(f) and Appendix C to 10 CFR Part 50, HL&P is financially qualified to design and construct Allens Creek Nuclear Generating Station Unit 1.
This conclusion is based on the staff's determination that HL&P has demonstrated reasonable assurance of obtaining the funds to carry out this activity.
Allens. Creek SSER #4 20-7
In connection with the above, it should be noted that the staff does not consider an applicant's financing plan to be a forecast of what will necessarily occur. -The applicant need only demonstrate one possible way by which the planned capital requirements, including those resulting from construction of the subject facility, might reasonably be_ financed.
The staff realistically expects that the financing plans will change to accommodate changing financial and economic conditions.
The proposed financing is in accord with general i
industry practices, and the assumptions being used, although not susceptible to precise measurement against absolute criteria, are in line with what one might expect under the postulated conditions.
Because the financing projections can be characterized as reasonable, the staff concludes that the reasonable assurance standard has been satisfied.
J Allens Creek SSER #4 20-8
4
21.0 CONCLUSION
S Based on its analysis of the proposed design of the Allens Creek Nuclear 1enerat-ing Station Unit 1, the staff concludes that in accordance with the provisions of Sections 50.35(a) and 50.40 of 10 CFR Part 50:
(1) The applicant has described the proposed design of the facility, including, but not limited to, the principal architectural and engineering criteria for the design, and has identified the major features or components incorporated therein for the protection of the health and safety of the public.
(2) Further technical or design information, as may be required to complete the safety analysis and which can be reasonably left for the later consideration, will be supplied in the Allens Creek Nuclear Generating Station Unit 1 Final Safety Analysis Report.
(3) Safety features er components that require research and development have been described and identified by the applicant, and reasonably designed research and development will be conducted to resolve safety questions associated with these features or components.
(4) On the basis of the foregoing, there is reasonable assurance that (a) such safety questions will be satisfactorily r2 solved at or before the latest date stated in the application for compietion of construction of the pro-posed facility, and (b) taking into ccnsideration the site criteria con-tained in 10 CFR Part 100, the proposed facility can be constructed and operated at the proposed location without undue risk to the health and safety of the public.
(5) The Houston Lighting & Power Company is technically qualified to design and construct the proposed facility.
(6) The applicant has reasonably estimated the costs and is financially quali-fled te design and construct the proposed facility.
(7) The issuance of a permit for construction of the facility will not be inimical to the common defense and security or to the health and safety of the public.
i l
. Allens Creek SSER #4 21-1
l l
l REFERENCES U.S. Nuclear Regulatory Commission NUREG Reports NUREG-75/087 Standard Review Plan (now published as NUREG-0800)
NUREG-0460 Anticipated Transients Without Scram for Light Water l
Reactors, Vol. 3 (December 1978) and Vol. 4 (March 1980)
NUREG-0515 Safety Evaluation Report Related to Construction of Allens Creek Nuclear Generating Station Unit 1, (November 1974),
Supplement 1 (June 1975), Scoplement 2 (March 1979), and Supplement 3 (July 1981).
NUREG-0519 Safety Evaluation Report Related to the Operation of the LaSa1.le County Station, Units 1 and 2, March 1981.
NUREG-0654 Criteria for Preparation and Evaluation of Radiological Plans and Preparedness in Support of Nuclear Power Plants, Rev. 1, November 1980.
NUREG-0718 Licensing Requirements for Pending Applications for Con-struction Permits and Manufacturing License, Rev. 1, March 1981.
'ther Items Ammerican Concrete Institute, " Analysis and Design of Reinforced Concrete Bridge Structures," ACI-443-74.
American Concrete Institute, " Building Code Requirements for Reinforced Concrete,"
American Concrete Institute /American Society of Mechanical Engineers, " Standard Code for Concrete Reactor Vessels and Containments," ACI-359.
American Society of Mechanical Engineers, " Boiler and Pressure Vessel Code."
EBASCO Services, Inc., " Containment Structures Design Report," Rev.1 (July 1977) and Rev. 2 (December 1979).
Title 10 Code of Federal Regulations Part 20, " Standards for Protection Against Radiation," January 1981.
Title 10 Code of Federal Regulations Part 50, " Domestic Licensing of Production and Utilization Facilities," January 1981.
U.S. Atomic Energy Commission, " Safety Evaluation of the Allens Creek t'uclear Generating Station Units 1 and 2," November 1974.
Allens Creek SSER #4 R-1
F APPENDIX A CONTINUATION OF CHRON0 LOGY March 30, 1979 Letter from applicant transmitting Amendment 52 to the PSAR which incorporates commitments made to the staff l
during the preparation of SER Supplement 2.
April 9, 1979 Letter to applicant transmitting Supplement 2 to the SER.
l May 17, 1979 Letter from applicant concerning reactor pressure vessel off-loading facility.
June 13, 1979 Meeting with utilities in Bethesda, Maryland, to discuss policies regarding review of current CP and OL applications and criteria for establishing priorities for review of such applications.
(Summary issued July 2, 1979)
June 26, 1979 Board notification:
Fire retardant cable coatings.
July 20, 1979 Letter to applicant forwarding NUREG-0576.
August 6, 1979 Board notification:
First report of TMI-2 Lessons Learned Task Force.
j August 9, 1979 Letter from applicant concerning emergency planning.
September 17, 1979 Letter from applicant transmitting Amendment 53 to the PSAR.
October 10, 1979 Letter to applicant concerning NRC position regarding TMI-2 accident review.
October 17, 1979 Letter to applicant concerning NRC plans to use generic analyses to develop early verification program to resolve i
ATWS issues for PWRs and BWRs.
October 23, 1979 Letter to applicant concerning preliminary plans for coping with potentic.I consequences of emergencies beyond site boundary.
November 9, 1979 Letter to applicant transmitting " Discussion of TMI Lessons Learned Short-Term Requirements."
November 14, 1979 Letter from applicant concerning TMI Lessons Learned Task l
Force recemmendations.
December 4, 1979 Meeting with CP applicants in Bethesda, Maryland, to discuss emergency preparedness requirements.
(Summary issued February 11, 1980) l l
Allens Creek SSER #4 A-1
(
December 20, 1979 Letters from applicant transmitting " Containment Structures Design Report, Revision 2" and Amendment 54 to the PSAR which incorporate the latest containment load definition information from GE and document a design change from a semiellipsoidal to a hemispherical containment dome.
December 26, 1979 Letter to applicant requesting evacuation time estimates for areas near plant.
February 5, 1980 Letter to applicant advising of issuance of NUREG-0588.
February 28, 1980 Letter from applicant concerning priority of NRC CP activities.
March 19, 1980 Meeting with CD applicants in 'ethada, Maryland, regarding resumption of tP application review activities associated with THI-2 accident.
(Summary issued April 3, 1980)
March 28, 1980 Meeting with CP applicants in Bethesda, Maryland, regarding evaluation of progress made to date.
(Summary issued April 3, 1980)
April 21, 1980 Letter to applicant concerning Category I masonry walls.
May 20, 1980 Letter to applicant concerning NUREG-0577.
May 28, 1980 Letter to applicant transmitting an ACRS letter concerning near-term construction permit applications to applicant and parties.
May 30, 1980 Order issued by the ASLB instructing intervenors to file certain documents by June 9, 1980.
June 6, 1980 Letter to applicant concerning NUREG-0619.
June 12, 1980 Letter to applicant concerning reorganization of the Office of NRR.
June 25, 1980 Letter to applicant (generic) concerning Union of Concerned Scientists.
June 26, 1980 Letter to applicant concerning Commission guidance for power reactor operating licenses.
June 30, 1980 Letter to applicant (generic) concerning regional meetings for applicants and vendors.
July 2, 1980 Letter to applicant concerning evacuation times (generic).
July 22, 1980 Order scheduling prehearing conference issued by the ASLB.
Prehearing conference will be held August 13, 1980 in Houston, Texas.
Allens Craek SSER #4 A-2
-September 3, 1980-Representatives from GE and NRC meet in Bethesda, Maryland, to discuss the GE proposed inservice feedwater nozzle leakage detection system.
September 18, 1980 Memorandum for the Public Document Room issued by the chairman, ASLB.
October 1, 1980 Letter to applicant (generic) concerning environmental qualification of safety-related equipment.
October 3, 1980 Memorandum and Order issued by the ASLB.
A second prehearing conference will be held.
' October 6, 1980 Letter to applicant (generic) concerning NUREG-0577.
t October 27, 1980
-Order scheduling second prehearing conference issued by the ASLB.
The second prehearing conference will be held on December 2, 1980 at the Holiday Inn in Houston, Texas.
November 4, 1980 Letter to applicant (generic) concerning ODYN code.
November 13, 1980 Letter to applicant (generic) concerning emergency planning final regulations.
November 24, 1980 Reconstitution of ASLB.
The chairman is Alan S. Rosenthal, and members are Dr. John H. Buck and Christine N. Kohl.
November 25, 1980 Order scheduling hearing.
December 16, 1980 Memorandum and Order issued by the ASLB.
The prehearing conference held on December 2, 1980 shall control the subsequent course of the proceeding.
December 22, 1980 Letter to applicant concerning control of heavy loads (generic).
December 31, 1980 Letter from applicant transmitting a copy of the annual financial statement of the companies that make up Houston Lighting & Power Company (HL&P).
January 21, 1981 Letter from applicant transmitting Amendment 55 to the PSAR which contains the applicant's response to the emergency planning requirements of 10 CFR 50, Appendix E.
January 26, 1981 Letter to applicant transmitting a copy of the ACRS report on requirements for near-term cps, dated January 12, 1981.
February 3, 1981 Representatives from NRC, HL&P, and utilities having pending CP applications discuss hydrogen control and associated containment pressures for subject plants.
February 10, 1981 ACRS report on requirements for near-term construction permits and manufacturing licenses issued.
Allens Creek SSER #4 A-3
February 16,-1981 Letter from applicant requesting commencement of hearings on safety on or about May 1 and with a continuous session until all safety issues are heard.
February 19, 1981 Order issued by the ASLB that an additional formal evidentiary session on environmental matters be held in Houston, lexas, from March 16-19, 1981.
February 20, 1981 Letter to applicant concerning NUREG-0619 (Generic Letter 81-11).
March 9, 1981 Letter from applicant transmitting Amendment 56 to the PSAR, which'is a revision of Chapter 4 and Sections 3.9.4, 6.3.3, and 15.1 of the PSAR to reflect a change in fuel design to include prepressurized fuel rods and two, instead of.one, hollow, water-filled tubes.
-March 10, 1981 Representatives from HL&P and NRC meet in Bethesda, Maryland, to discuss the licensing plans for Allen Creek.
March 25, 1981 Letter from applicant advising it wants to begin construc-tion in 1982.
March 26, 1981 Order scheduling resumed hearings.
The hearings are scheduled for May 11-22 at Ramada Inn, Houston, Texas, and June 1-12 at Bates College of Law (Krost Hall), Houston.
April 8, 1981 Representativu from utilities that did not get construction permits before the accident at Three Mile Island meet with NRC staff to discuss plans for review of near-term CP plants considering the proposed rule on licensing require-ments for pending construction permit and manufacturing license applications.
May 1, 1981 Letter from applicant transmitting Amendment 57 to the PSAR.
May 5-7, 1981 Representatives from NRC and HL&P meet in Bethesda, Maryland, to discuss quality. assurance licensing requiremente associated with lessons learned from the accident at TMI.
May 15, 1981 Letter from the applicant transmitting Amendment 58 to the PSAR to provide updated informatic.i addressing contentions concerning charcoal adsorber fires, tutbine missiles, and blockage of the ultimate heat sink intake structure.
May 15, 1981 Letter from applicant informing that it will direct EBASCO Services to benchmark its pipe stiess computer program code against NUREG/CR-1677.
May 18-22, 1981 Representatives from NRC and HL&P meet in Bethesda, Maryland,.
to discuss licensing require:aents associated with lessons learned from the accident ~at TMI.
Allens Creek SSER #4 A-4
May 26, 1981 Letter from applicant confirming its commitment to install during construction of the plant a jetty system to provide stabilization for the bank of the Brazos River.
June 3, 1981 Letter to applicant concerning a meeting to be held in f
Bethesda, Maryland, from July 7-10, 1981 on environmental qualification of safety-related electrical equipment.
June 4, 1981 Letter to applicant requesting additional information on the emergency plan.
i June 17, 1981 Representives from NRC and HL&P meet in Bethesda, Maryland, to discuss increases in containment stresses for test conditions and during cooldown following inadvertent inerting.
June 18, 1981 Order scheduling resumed hearing--August 17-28 and September 14 through 24 at Ramada Inn, Houston, Texas, from 9 A.M. to 5 P.M.
June 18, 1981 Letter from applicant transmitting Amendment 59 to the PSAR which addresses TMI-2 requirements.
July 7, 1981 Letter from applicant advising that it will comply with the final resolution of the core thermocouple issue that is determined for the LaSalle plant.
4 July 14, 1981 Generic Letter 81-26--Licensing Requirements for Pending Construction Permit and Manufacturing License Applications.
July 20, 1981 Letter to applicant requesting updated financial information.
July 22, 1981 ASLB issues order denying Doherty motion for leave to file Contention 57.
August 5, 1981 Letter from applicant concerning impact of RCP seal damage following small-break LOCA with loss of offsite power.
August 6, 1981 SER Supplement 3 issued.
August 7, 1981 Letter to applicant requesting updated information on unresolved safety issues.
Aegust 12, 1981 Letter from applicant forwarding updated financial information.
August 12, 1981 ASLB issues order granting boherty renewed motion for reconsideration and admitting his amended Contention 21.
August 13, 1981 Meeting with applicant to discuss unresolved safety issues.
August 21, 1981 Letter from applicant providing responses to generic issues.
Allens Creek SSER #4 A-5
i f
August 21, 1981 Letter from applicant transmitting Amendment 60 to the PSAR, which is the applicant's response to'a staff request for additional information on emergency planning and clarification of liquid petroleum gas pipe line analysis.
September 1,~ 1981 'ASLB issues second order ruling upon votions for summary
. disposition.
September 1, 1981 ASLB issues order scheduling resumed hearir.gs.
September 3, 1981 ASLB issues amendment of order scheduling tesumed hearings.
Hearing will resume September 14 through September 18 and again October 5 and October 16.
September'15, 1981 Letter to applicant concerning design to accommodate ir.clusion of plant modifications for anticipated transients with. failure to scram.
September 16, 1981 Letter from ACRS concerning the post-accident inerting system.
September 22, 1981 ASLB issues order scheduling resumed hearings.
Hearings will resume October 5 through October 9 and thereafter will resume October 26 through October 30, November 16 through November 20, and December 7 through December 11.
October 7, 1981 Letter to ACRS responding to September 16, 1981 letter concerning the post-accident inerting system.
October 23, 1981 Representatives of NRC and HL&P met in Bethesda, Maryland, to discuss design allowable stresses in concret.e reinforcing for pressure loads.
Meeting summary issued on October 28, 1981.
October 28, 1981 Letter from applicant stating that the preliminary design-of the Allens Creek facility does not preclude incorporation of staff's proposed ATWS position in final design.
i Allens. Creek SSER #4 A-6 I
APPENDIX B ERRATA TO ALLENS CREEK SAFETY EVALUATION REPORT, SUPPLEMENT NO.2*~
' Cover-Change " Houston Power & Lighting Company" to " Houston Lighting & Power Company."
Title Page
- Change " March 1978" to " March 1979."
page 2-5 Section 2.2.2 Second paragraph, line 6, change "Wanne" to " Wayne."
Third paragraph, line 6, change "Wanne" to " Wayne."
Page 2-6 Table 2.1 First item, change "6-inch crude" to "10-inch crude."
Fifth item, change " Superior Oil Company" to " Peninsula Resource Corporation;" change "8. inch raw products line" to "8.6-inch 3
natural gas line;" and change "600 psi" to "up to 50,000 ft per day at up to 1,000 psi."
Sixth item, change "up to 400,000 fta per day at 750 to 900 psi" to "up'to 400,000,000 ft per day at up to 975 psi."
Page 2-17 Line 16 Change " Amendment 52" to " Amendment 53."
First full paragraph, lines 4 and 5 and lines 13 and-14 Change " seismic Category I" to " seismically designed."
Page 3-9 Section 3.10, line 9 Change "However, the applicant has provided a commitment that all seismic Category.I equipment used in the Allens Creek plant will meet the requirements of.IEEE 344-1975 and Regulatory Guide 1.100, Revision 1" to "However, the applicant has provided a commitment that this equipment will-meet the requirements of IEEE 344-1975 and Regulatory Guide 1.100, Revision 1, unless qualified by other means found acceptable by the staff.during the operating license stage.of review."
"These errata were issued in 1979.
-Allens Creek SSER 14
<B-1
l.
i Page 4-1 Section.4.2.1(1), line 9 Change 2.52 to 2.50.
Page 4-11 Second full paragraph, line 8 Change "... boiling water reactor..." to "... Boiling Water Reactor..."
Page 4-11 Line 15 Change " January 17, 1978" to " September 22, 1976."
Page 5-4 Section 5.2.3(1), line 9 Change " Appendix B" to " Appendix G."
Page 5-5 Section 5.2.6, heading Change " Protection" to " Detection."
Page 6-7 Line 2 Change "Since then we have advised the applicant that due to the conservative plant operating conditions, as indicated by the results of analyses of emergency core cooling system performance for Allens Creek as well as for other BWR 6/ Mark III plants, we believed that the assumption of one percent metal-water reaction provides a sufficiently conservative basis for combustible gas control system design." to "Also, standards for combustible gas control system in light water cooled power reactors was issued as a Commission regulation on November 10, 1978 as Section 50.44 to 10 CFR Part 50."
Line 12 Change "Since five times the metal-water reactor calculated to demonstrate compliance with Section 50.46 of 10 CFR Part 50 (0.17 percent as reported in Section 6.3.2 of this supplement) is 0.85 percent and since 0.00023 inches is only about 0.7 percent of the clad wall thickness of 0.034 inches, the use of one percent assures compliance with Section 50.44 of 10 CFR Part 50, " Standards for combustible gas control system in light water cooled power reactors."" to "The applicant has agreed to provide a commitment in Amendment 53 to the PSAR that the combustible gas control system will be designed for nydrogen released by 0.85 percent metal-water reaction, i.e., five times the 0.17 percent reported in Section 6.3.2 of this supplement for compliance with Section 50.46 of 10 CFR Part 50.
This commitment is in accordance with Section 50.44 of 10 CFR Part 50, and therefore is acceptable."
Allens Creek SSER #4 B-2
.-- _-__ j
Page 6-7 Section 6.2.5 Remove the last three paragraphs of Section 6.2.5.
Add one paragraph:
"The redundant recombiners will be perma-nently installed inside containment.
This is permitted by the provisions of Regulatory Guide 1.7 and is acceptable.
The appli-cant has committed to provide a testing program to demonstrate the operability of the recorrhiners.
We will review the testing program and operating procedures at the operating license stage of review."
Page 6-14 Section 6.5, second paragraph, line 5 Omit "(Revision 2, March 1978)."
Page 8-1 Section 8.1, line 12 Change "Use of IEEE Std. 308-1971, Criteria for Class IE Electric Systems for Nuclear Power Generating Stations" to " Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants."
Page 8-3 Section 8.3.1, last paragraph Change "IEEE Std. 208-1974" to "IEEE Std. 308-1974."
Line 2, change "... separate ventilating fan for each room..."
to "... ventilation system..."
l l
Line 3, change "There will be no common ventilation ducts for the battery rooms." to "The ventilation system includes a comn:on collection header, exhaust fans and a common discharge header."
Page 8-4 Section 8.3.2, last paragraph, line 2 Change "IEEE Std. 208-1971" to "IEEE Std. 308-1971."
Page 9-1 Section 9.2.1, line 3 Change "Section 8.9.2.1" to "Section 9.2.1."
Page 9-1 Section 9.2.2, first paragraph Line 1, change " Amendment 37" to " Amendments 37 and 48."
Last line, change "The new decay heat load is that produced by one quarter of a full core load after 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> plus four succes-sive annual batch discharges." to "The new decay heat load is the sum of that produced by one quarter of a full core load (irradiated for four years) after 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> plus nine successive annual batch discharges."
Allens Creek SSER #4 B-3
.Page 9-2 Line 7 Change "Section 9.22" to "Section 9.2.2."
Page 9-3 Section 9.5.1, first paragraph, lines 2 and 3 Change "...added a..."
to "...provided clarification that with Amendment 22 it had added a..."
Page 9-5 Section 9.5.4, second paragraph, line 4-Change "... Engine fuel 0.1 Storage..." to "... Engine Fuel Oil Storage..."
Page 11-4 Table 11.2, line 8 Change "Xe-133 66 250 66 10 2000 2300 2300" to "Xe-133 66 250 66 10 20,000 2300 23,000."
/
Allens Creek SSER #4 B-4
~
l-i l-APPENDIX C NUCLEAR-REGULATORY COMISSION STAFF GENERIC ISSUES g
, C.4 Unresolved Safety Issues l
In SER' Supplement 2 the staff stated that all of the task action plans for b
Category A tasks addressing unresolved safety issues that are applicable to the Allens Creek' facility, with the exception of Tasks A-43, Containment Emer-gency Sump Reliability, and A-44, Station Blackout, were included in " Task Action Plans for Generic Activities, Category A".(NUREG-0371), published in-November 1978. With the exception of Tasks A-9, A-43, and A-44, task action plans for those tasks are now included in NUREG-0649, " Task Action Plans for l
Unresolved Saftty Issues Related to Nuclear Power Plants," February 1980.
Rulemaking has been proposed for Task A-9, and there is no task action plan.'
The task action plan for Task A-43 was completed in January 1981, and the task action plan for A-44 was completed in July 1980.
Each task action plan pro-vides a description of the problem; the staff's approaches to its resolution;
. a gene:al discussion of the bases on which continued plant licensing or opera-tion can proceed pending completion of the task; the technical organizations
- involved in the task and estimates of the manpower required;. a' description of the interactions with other NRC offices, the Advisory Committee on Reactor Safe-guards and outside organizations; estimates of funding required for contractor-supplied technical assistance; prospective dates for completing the tasks; and a description of potential problems that could alter the planned approach or
(
schedule.
In Appendix C to SER Supplement 2, the staff identified 15 generic l
tasks, including Tasks A-9, A-43, and A-44, addressing unresolved safety' issues that were applicable to the Allens Creek facility.
The staff reviewed each of these tasks as it relates to the Allens. Creek facility and set.forth reasons based on its review of each of these items for its conclusion that Allens Creek Nuclear Generating Station Unit 1.may be constructed and operated before the ultimate resolution of these issues without endangering the health and safety of the public.
In addition to the task action plans, the staff issues the " Aqua Book" (NUREG-0606) on a quarterly basis. This book, entitled " Office of Nuclear Reactor Regulation Unresolved Safety Issues Summary, Aqua Book," provides current i
schedule information for each of the unresolved safety issues.
It also includes information relative to the implementation status of each unresolved safety issue for which technical resolution is complete.
- The' current issue of NUREG-0606, Volume 3, Number 3, dated August 21, 1981, includes information relative to the' implementation status of the following
' unresolved safety issues that are applicable to Allens Creek and for which i
l
. technical resolution is complete:
JA-2 TAsymmetric Blowdown Loads on Reactor Primary Coolant Systems-
--A-9 --Anticipated Transients Without Scram (ATWS)
Allens Creek SSER #4 C-1~
w m.s
-sa
.,e
-, =
,----.---,=,,,,,,w-*ye-y
_,.s-
_ew
f f
l~
r A-10 BWR Feedwater Nozzle Cracking l
A-24 Qualification of Class lE Safety Related Equipment A-31 Residual-Heat Removal Requirements A-36 Control of Heavy Loads Near Spent Fuel A-42 Pipe Cracks in Boiling Water Reactors In SER Supplement 2, the staff had discussed each of these issues and provided bases for a conclusion that there is reasonable assurance that Allens Creek Nuclear Generating Station Unit 1 may be constructed and operated before the ultimate resolution of each of these generic issues without endangering the health and safety of the public.
At the staff's request and as discussed in Section 15.2 of this supplement, the applicant, by letter dated October 28, 1981, stated that construction of the Allens Creek facility as now designed would not result in constraints to full implementation of the technical resolu-tion of the ATWS issue (Task A-9) as described in Volume 4 of NUREG-0460.
In response to the staff's request, the applicant, by letter dated August 21, 1981, provided information describing the status of plant-specific implementa-tion of the technical resolutions of Tasks A-10, A-31, A-36, and A-42.
The staff has not yet requested information from the applicant describing the status of the plant-specific implementation of the technical resolution of Task A-24.
Ultimate resolution of these seven issues for the Allens Creek facility will include implementation in the final design, which the staff will review during the operating license stage.
For this construction permit stage, the staff reaffirms its conclusion, stated in SER Supplement 2, that there is reasonable assurance that Allens Creek Nuclear Generating Station Unit 1 may be constructed and operated before the ultimate resolution of each of these seven generic issues without endangering the health and safety of the public.
An update of the discussion in SER Supplement 2 on the bases for this conclu-sion for these seven issues follows.
A-2 Asymmetric Blowdown Loads on Reactor Primary Coolant Systems Technical resolution of this issue includes criteria developed as a part of Task Action Plan A-2 for evaluating PWR plant assessments for asymmetric loads.
This technical resolution does not change the staff's previous conclusion for the Allens Creek facility as stated in SER Supplement 2, that is, "Since the applicant has committed to design the reactor primary coolant system for these loads, we conclude there is reasonable assurance that the Allens Creek Nuclear Generating Station, Unit 1, may be constructed and operated prior to the ulti-mate resolution of this generic issue without endangering the health and safety of the public."
A-9 Anticipated Transients Without Scram (ATWS)
A proposed staff resolution was published for comment as Volume 4 of NUREG-0460,
" Anticipated Transients Without Scram for Light Water Reactors," dated March 1980.
Allens Creek SSER #4 C-2
In SER Su'pplement 2, the staff reported that the applicant had provided a commitment that Allens Creek Unit 1 will be designed so that implementation of the potential requirements described in Volume 3 of NUREG-0460 would not be compromised by construction. The_ staff then stated its conclusion "A satis-factory solution to the generic task will be obtained before Allens Creek-Unit 1 is put in operation.
Therefore, since the applicant has committed not to preclude implementation of design modifications there is reasonable assur-ance that the proposed Allens Creek facility can be constructed'and operated at the proposed location without undue risk to the health and safety of the public."
The staff requested that the applicant reassess the impacts of construction on implementation of the provisions of Volume 4 of NUREG-0460, particularly with respect to constraints to implementation that had been identified for plants already constructed in the event that the. final rule may approximate alterna-tive 4a. As discussed in Section 15.2 of this supplement, the applicant has completed that assessment and has concluded that no constraints now exist in the preliminary design that preclude implementation of the provisions of Volume 4 of NUREG-0460. Therefore, the staff concludes that construction of the Allens Creek facility should not result in constraints to the implementa-tion of the applicable provisions of Volume 4 of NUREG-0460.
If a rule has not been promulgated by the operating license stage of review, the staff will have to consider whether to implement some of the provisions of its proposed position before operation.
Therefore, the staff concludes with respect to this issue that construction _ of the proposed Allens Creek facility.
can proceed without undue risk to the health and safety _of the public by subse-quent operation of the facility because (1) the facility will be designed and operated in accordance with an ATWS regulation or (2) on the basis of the applicant's assurances, provisions of the staff's proposed position could be implemented before operation.
A-10 BWR Feedwater Nozzle Cracking Technical resolution of this issue is described in "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," NUREG-0619, November 1980.
In SER Supplement 2, the staff noted that the nuclear steam supply system (NSSS) vender is developing design modifications to avoid the nozzle cracking problem.
In addition, the NSSS vendor plans to perform out-of-reactor verification tests to ensure the adequacy of the modifications and to modify the instrumentation of the first reactor vessel using these design changes.
The staff also noted:
.that the applicant has described the incorporation of these design changes in Section 5.4.2.3.7 of the PSAR.
In response to the staff's request, the appli-cant by letter of August 21, 1981 described the current status on incorporation of design modifications in the Allens Creek facilities.
The applicant also stated that staff conclusions would be used to develop inspection and testing procedures for the Allens Creek facility and that new testing techniques will be examined for applicability as they are developed.
In SER Supplement 2, the staff noted that the acceptability of design modifica-tions is expected to be establishec well before the proposed operation of the Allens Creek facility.
Because the applicant's modifications are now defined
.in more detail and are responsive to the technical resolution, the staff Allens Creek SSER #4 C-3 T-
reaffirms its conclusion with respect to this issue as stated in SER Supple-ment 2 that there is reasonable assurance that the Allens Creek Nuclear Gener-ating Station Unit 1 may be constructed and operated without endangering the health and safety of the public.
A-24 Qualification of Class 1E Safety-Related Equipment Technical resolution of this issue is described in " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," NUREG-0588, Revision 1, dated July 1981.
The positions contained in the report provide guidance on (1) how to establish environmental service conditions, (2) how to select methods that are considered appropriate for qualifying the equipment in different areas of the plant, and (3) other areas such as margin, aging, and documentation.
The positions in the report are intended to supplement, in selected areas of the qualification issue, the requirements described in the 1971 and the 1974 versions of Institate of Electrical and Electronics Engineers (IEEE) Standard 323.
On May 23, 1980, 3 Commission Memoranrfum and Order (CLI-80-21) endorsed the positions in the For Comment version of NUREG-0588 as the interim positions that shall be satisfied until the " final" positions are estab-lished in rulemaking.
The staff anticipates that the proposed role will be issued for comment in December 1981.
l Because (1) the applicant has been provided with interim guidance, (2) the l
Commission's Memorandum and Order endorsed that guidance as the staff's interim position, and (3) rulemaking on a final position will soon be initiated, the staff reaffirms its position with respect to this issue as stated in SER Supple-ment 2 that there is reasonable assurance that the Allens Creek Nuclear Gener-t l
ating Station Unit 1 may be constructed and operated without endangering the health and safety of the public.
A-31 Residual Heat Removal Requirements The technical resolution of this issue was encompassed in a revision to Stand-ard Review Plan Section 5.4.7 in 1978.
In Section 5.4.5 of SER Supplement 2, l
the staff described its utilization of Revision 2 of Standard Review Plan Sec-l tion 5.4.7 and concluded that the applicant's residual heat removal system coupled with an alternate method of achieving a cold shutdown was acceptable at the construction permit stage of review.
Subject to confirmation of the acceptability of the final design implementation including availability of prototype testing data, the technical resolution has been implemented in the Allens Creek facility.
Therefore, the staff reaffirms its conclusion with respect to this issue as stated in SER Supplement 2 that Allens Creek Unit may oe constructed and operated without endangering'the health and safety of the public.
A-36 Control of Heavy Loads Near Spent Fuel The technical resolution of this issue is described in " Control of Heavy Loads at Nuclear Power Plants," NUREG-0612, dated July 1980.
In response to a staff request, the applicant by letter of August 21, 1981 stated that it will utilize
,the recommended guidelines of Section 5.1 of NUREG-0612 and will adopt the
" defense-in-depth" approach.
Although a detailed evaluation of the design and equipment design specifications as compared with those in NUREG-0612 is not Allens Creek SSER #4 C-4
available at this time, the applicant has initiated an ongoing evaluation pro-
. gram and has committed to address the requirements of NUREG-0612 throughout the design, construction, and operation of the Allens Creek facility.
Therefore, the staff reaffirms its conclusion with respect to this issue as stated in SER Supplement 2 that there is_ reasonable assurance that the Allens Creek Nuclear Generating Station Unit 1 may be constructed and operated with-out endange"ing the health and safety of the public.
'A-42 Pipe Cracks in Boiling Water Reactors The technical resolution of this issue was reported in " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping,"'NUREG-0313, Revision 1, dated July 1980.
In SER Supplement 2, the staff had stated that the Allens Creek facility was already.in compliance with NUREG-0313 and that only nominal changes in the roquirements (e.g., expansion of augmented inspection programs to other areas) were expected to result from Task A-2.
In NUREG-0313, Revision 1, the actual principal differences were identified as:
(1) The guidelines for reducing the intergranular stress corrosion cracking (IGSCC) susceptibility have been extended to cover ASME Code Class 2 and 3 piping.
(2) Augmented inservice inspection requirements for nonconforming safe ends have been included.
(3) The inservice inspection sampling schemes have been updated to comply with the most recent Code edition that has been accepted by the NRC.
(4) The recommendations of NUREG-0531 (1978 Pipe Crack Study Group Report),
which cannot be implemented immediately without further NRC evaluation, have been identified.
Certain developmental items were identified in NUREG-0313, Revision 1, for future improvements tha', are not required for present plant safety for the resolution of Generic Task No. A-42.
In response to a staff request, the applicant by letter of August 21, 1981 verified that the guidelines of NUREG-0313, Revision 1, would be applied to ASME Code Class 2 and 3 piping in addition to Class 1 piping in the Allens Creek facility.
On the basis of these considerations, the staff reaffirms its conclusion with respect to this issue as stated in Supplement 2 that there is reasonable assur-ance tha'. Allens Creek Unit 1 may be constructed and operated without er. danger-ing the health and safety of the public.
Since the issuance of SER Supplement 2, four new unresolved safety issues were designated by the Commission in December 1980 and described in NUREG-0705,
" Identification of New Unresolved Safety Issues Relating to Nuclear Power Plants, Special Report to Congress," dated March 1981..In response to the staf f's request, the applicant by its _ letter of August 21, 1981 addressed Allens Creek SSER #4 C-5
these issues as they pertain to the Allens Creek facility.
The NRC staff's evaluation of Allens Creek for each of the new unresolved safety issues follows.
This evaluation includes the staff's bases for licensing before the ultimate resolution of the issues.
Its conclusions are based in part on the information provided by the applicant in a letter of August 21, 1981.
A-45 Shutdown Decay Heat Removal Requirements Following a reactor shutdown, the radioactive decay of fission products con-tinues to produce heat (decay heat) that must be removed from the primary system. The principal means for removing this heat in a boiling water reactor while at high pressure is through the steam lines to the turbine condenser.
The condensate is normally returned to the reactor vessel by the feedwater system; however, the steam turbine-driven reactor core isolation cooling system is provided to maintain primary system inventory if AC power is not available. When the system is at low pressure, the decay heat is removed by the residual heat removal (RHR) systems.
This unresolved safety issue will evaluate the benefit of providing altctnate means of decay heat removal that could substantially increase the plant's capability to handle a broader spec-trum of transients and accidents. The study will consist of a generic system evaluation and will result in recommendations regarding the desirability of and possible design requirements for improvements in existing systems or an alternative decay heat removal method if the improvements or alternative can significantly reduce the overall risk to the public.
The Allens Creek reactor has various methods for the removal of decay heat.
As discussed above, the decay heat is normally rejected to the turbine con-denser and returned to the vessel by either the feedwater system or the reactor core isolation cooling (RCC) system (from the condensate storage tank).
If the condenser is not available (e.g., through loss of offsite power), heat can be removed by means of the safety relief valves to the suppression ~ pool.
Also,'the high pressure core spray (HPCS) system is pro-vided if the RCIC system is not available.
Both of these systems can recir-culate fluid to the vess?1 from either the condensate storage tank or the suppression pool.
If the RCIC system and high pressure core spray are unavail-able, the reactor system pressure can be reduced by the automatic depressuri-zation system so that cooling by the RHR system can be initiated. When the condenser is not used, the heat rejected to the suppression pool is subsequently removed by the RHR system.
Alternatively, heat can be removed by using the RHR system in the steam condensing mode operating in conjunction with the RCIC system.
The RHR system is of safety grade quality. The RCIC system turbine pump is steam driven, and other system contrc! ccmponents are powered from safety-related onsite DC power to further enhance system availability.
The HPCS system is powered by a dedicated, onsite safety grade diesel generator.
The RHR system consists of three pumps, any one of which can maintain reactor water lael.
The RHR and LPCS systems can be powered by the redundant onsite safety grade diesel generators.
Accordingly, the staff has concluded that there is reasonable assurance that Allens Creek may be constructed and operated before the ultimate resolution of this generic issue without endangering the health and safety of the public.
Allens Creek SSER #4 C-6
A-46 Seismic Qualificaton of Equipment in Operating Plants The design criteria and methods for the seismic qualification of mechanical end electrical equipment in nuclear power plants have undergone significant change during the course of the commercial nuclear power program.
Conse-quently, the margins of safety provided in existing equipment to resist seismically induced loads and perform the Intended safety functions may vary considerably.
The seismic qualification of the equipment in operating plants cust, therefore, be reassessed to ensure.the' ability to bring the plant to a safe shutdown condition when subjected to a seismic event.
The objective of this unresulved safety issue is to establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and electrical equipment at all operating plants instead of attempting to back-fit current design criteria for new plants.
This guidance will concern equip-ment required to safely shut down the plant as well as equipment whose function is not required for safe shutdown, but whose failure could result in adverse conditions that might impaw shutdown functions.
f Allens Creek was reviewed against current seismic criteria and approved by the staff in accordance with current design criteria and methods for seismic quali-fication.
The staff's review is diz, cussed in Section 3.10 of SER Supplement 2.
Therefore, the statf concludes that there is reasonable assurance that Allens Creek may be constructed and operated before the resolution of this generic issue without undue risk to the health and safety of the public.
A-47 Safety Implications of Control Systems This issue concerns the potential for transients or accidents being made more severe as a result of control system failures or malfunctions.
These failures or malfunctions may occur independently or as a result of the accident or trans-ient under consideration.
One concern is the potential for a single failure such as a loss of a power supply, short circuit, open circuit, or sensor fail-ure to cause simultaneous malfunction of several control features.
Such an cccurrence would conceivably result in a transient more severe than those transients analyzed as anticipated cperational occurrences.
A second concern is for a postulated accident to cause control system failures that would make the accident more severe than analyzed.
Accidents could conceivably cause control system failures by creating a harsh environment in the area of the con-trol equipment or by physically damaging the control equipment.
Although it is generally believed that such control system failure:, would not lead to serious events or result in conditions that safety systems cannot handle, indepth studies have not been rigorously performed to verify this belief.
The potential for an accident that would affect a particular control system, and effects of the control system failures, may differ from plant to plant.
There-fore, it is not possible to develop generic answers to these concerns, but rather plar.t specific reviews ate required.
The purpose of this unresolved safety issue is to define generic criteria that will be used for plant-specific reviews.
The Allens Creek control and safety systems have been designed with the goal of ensurir.g that control system failures (either single or multiple) will not prevent automatic or manual initiation and operation of any safety system equipment required to trip the plant or to maintain the plant in a safe shut-Allens Creek SSER #4 C-7
down condition following any anticipated operational occurrence or accident.
1,
.This has-been accomplished by either providing. independence between safety and 1
-nonsafety-related systems or providing. isolating devices between safety and nonsafety-related systems.
These devices are designed to preclude.the propaga-tion of'nonsafety-related system equipment faults so that operation of the j
safety system equipment ~is not impaired.
~
p
-A systematic evaluation of the control system design, such as that contemplated for this unresolved safety issue, has not been' performed to determine whether postulated accidents could cause significant control' system failures that would j.
make the accident consequences more severe than presently analyzed.. However, a wide range of bounding transients and accidents has been analyzed to ensure that the postulated events would be adequately mitigated by the safety systems.
In addition ~, systematic reviews of safety systems have been performed with the goal of ensuring that control system failures (single or multiple) will not j.
defeat safety system action.
A specific subtask of this unresolved safety issue will be to study the reactor overfill transient in boiling water reactors to determine the need for prevent-ative and/or mitigating design measures to preclude or minimize the conse-r quences of this transient.
Several early boiling water reactors have experi-enced reactor vessel overfill transients with subsequent two phase or liquid i
flow through the safety relief valves.
Following these early events, commer-cial grade high-level (Level 8) trips have been installed at most boiling water reactors to terminate flow from the appropriate systems.
No overfilling (l
events have occurred since the Level 8 trips were installed.
1 The reactor overfill issue is currently under review by the 'BWR Owners' Group of which Allens Creek is a member.
Pending ultimate resolution of this item, the applicant has incorporated in the Allens Creek design a commercial grade i
high-level (Level 8) trip of the RCIC, HPCS, and feedwater systems to prevent j
the occurrence of the overfill transient.
1 Additionally, Allens Creek has committed to develop a reliability analysis pro-i gram to identify significant and practical improvements in the reliability of l
core and containment heat removal systems that do not impact excessively on the plant design.
This program will include failure modes and effects analyses (FMEA)'to investigate common cause failure mechanisms such as environmental l
factors, operator or maintenance errors, passive failures, and system interactions.
Changes in the design of control systems can be accommodated before the issu-ance of the operating license because instrumentation design.is normally com-pleted in the latter stages of plant construction.
Based on the above, the NRC staff has concluded that there is reasonable assur-ance that Allens Creek can be constructed and operated before the ultimate resolution of this generic issue without endangering the health and safety of i
the public.
5-l l~
t Allens Creek SSER J4 C-8
~ ~.
A-48 Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment.
Following a loss-of-coolant accident in a light water reactor plant, combuetible gases, principally hydrogen, may accumulate inside the primary reactor contain-ment as a result of (1) metal-water reaction involving the fuel element clad-ding; (2) the radiolytic decomposition of the water in the reactor core and the containment sump; (3) the corrosion of certain construction materials by the spray solution; and (4) any synergistic chemical, thermal, and radiolytic effects of postaccident environmental conditions on containment protective coating systems and electric cable insulation.
Because of the potential for significant hydrogen generation as the result of an accident, 10 CFR Section 50.44, " Standards for Combustible Gas Control System in Light Water Cooled Power Reactors," and General Design Criterion 41, i
" Containment Atmosphere Cleanup" (in Appendix A to 10 CFR Part 50), require that systems be provided to control hydrogen concentrations in the containment atmosphere following a postulated accident to ensure that containment integrity is maintained.
The regulation, 10 CFR 50.44, requires that the combustible gas control system i
provided be capable of handling the hydrogen generated as a result of degra-j dation of the emergency core cooling system so that the hydrogen release h five times the amount calculated in demonstrating compliance with 10 CFR 5j U or the amount corresponding to reaction of the cladding to a 6epth of 0.00023 in., whichever amount is greater.
The accident at TMI-2 on March 28, 1979 resulted in hydrogen generation well in excess of the amounts specified in 10 CFR 50.44.
As a result of this knowl-edge, it became apparent to NRC that. specific design measures are needed for handling larger hydrogen releases, particularly for smaller, low pressure con-3 tainments. As a result, the Commission determined that a rulemaking proceeding should be undertaken to define the manner and extent to which hydrogen evolu--
tion and other effects of a degraded core need to be taken into account in plant design.
An advance notice of this rulemaking proceeding on degraded core issues was published in the Federal Register on October 2, 1980.
Houston Light & Power Company has committed to provide a hydrogen control j
system in the Alleas Creek design.
Currently, a number of different methods are being considered throughout the industry, and it is expected that these efforts will produce valuable data which will help in selecting an optimum means of hydrogen control.
Further, it is expected that the pending rulemaking i
on degraded cores will determine the necessity for such a system.
For the pur-poses of meeting the stated requirement, a postaccident inerting system (PAIS) using C02 as an inerting agent is being, proposed.
However, for the reasons given above, the need for this system will be under continuing review.
The PAIS will be designed to prevent the combustion of hydrogen generated by oxidation of 100 percent of the active fuel cladding by maintaining an inert atmosphere following degraded core conditions. The containment stresses caused by. inadvertent inerting will not exceed American Society of Mechanical Engineers-(ASME) Service Level A Limits, nor will stresses resulting from pressures caused by the 100 percent metal-water reaction exceed ASME Service Level C Limits.
Allens Creek SSER #4 C-9
i i
In summary, the pending rulemaking may determine that a hydrogen control system with the capabilities projected for the proposed PAIS is not necessary.
If it does not, the applicant s already existing commitments described in SER Supple-ment 3 ensure that such a hydrogen control system will be provided for Allens Creek.
Finally, as discussed in Section 1.1 of this supplement, alternatives will be provided if the PAIS is determined to be unacceptable as a long-term solution to this issue.
The staff, therefore, concludes that there is reason-able assurance that Allens Creek may be constracted and operated before resolu-tion of this unresolved safety issue and the proposed rulemaking without undue risk to the health and aafety of the public.
REFERENCES l
U.S. Nuclear Regulatory Commission NUREG Reports NUREG-0313 Techni al Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping, Rev.1, July 1980.
NUREG-0371 Task Action Plans for Generic Activities, Category A, November 1978.
NUREG-0460 Anticipated Transients Without Scram for Light Water Reactors, l
Vol. 3, December 1978, and Vol. 4, March 1980.
l NUREG-0531 Investigation and Evaluation of Stress Corrosion Cracking in Piping of Light-Water Reactor Plants, February 1979.
NUREG-0588 Interim Staff Position on Environmental Qualification of Safety-t l
Related Electrical Equipment, Rev. 1, July 1981.
NUREG-0606 Office of Nuclear Reactor Regulation Unresolved Safety Issues Summary, Aqua Book, Vol. 3, No. 3, August 21, 1981.
NUREG-0612 Control of Heavy Loads at Nuclear Power Plants Kcsolution of Generic Technical Activity A-36, July 1980.
NUREG-0619 BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking, November 1980.
NUREG-0649 Task Action Plan for Unresolved Safety Issues R0 ated to Nuclear Power Plants, February 1980.
i NUREG-0705 Identification of New Unresolved Safety Issues Relating to Nuclear Power Plants, Special Report to Congress, March 1981.
Other Items American Society of Mechanical Engineers, " Boiler and Pressure Vessel Code."
Institute of Electrical and Electronics Engineers, Standard 323, 1971 and 1974.
Memorandum from R. M. Bernero, NRC, to H. Denton, " Approval of Task Action Plan A-44, ' Station Blackout,'" July 14, 1980.
Allens Creek SSER #4 C-10
Memorandum from L. Shao, NRC, to H. Denton, " Approval of Task Action Plan A-43,.
' Containment Emergency Sump Performance,'" January 16, 1981.
Title 10 Code of Federal Regulations Part 50, " Domestic Licensing of Production s,
and Utilization Facilities," January 1981.
f i
i i
i i
i 1
l Allens Creek SSER #4 C-11
APPENDIX D LETTERS ON POSTACCIDENT INERTING SYSTEM l
I l
I i
I I
l l
i
gn urug j
h UNITED STATES 8
i NUCLEAR REGULATORY COMMISSION
.,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8
WASHINGT ON, D. C. 20555
%q*****c*g o
September 16, 1981 Mr. William J. Dircks Executive Director for Operations U. S. Nuclear Regulatory Commission Washington, D.C.
20555
SUBJECT:
POST ACCIDENT INERTING SYSTEM
Dear Mr. Dircks:
During its 257th meeting September 10-12, 1981, the ACRS heard a presentation on the post accident inerting system (PAIS) proposed as a means of combustible J
gas control for the Allens Creek Plant in the event of a severe core damage accident.
As a result of the presentation, the Comittee concluded that there are a number of questions that remain to be addressed concerning the design bases of 'he ~ AIS including, but not limited to, hydrogen source term design t
P criteria, system actuation criteria (including inadvertent actuation),
containment pressurization, and equipment survivability.
The ACRS desires to review the proposed design of this system before it is approved by the NRC Staff or other means of hydrogen control are precluded.
Consideration of this matter by the Committee should be factored into proposed staff scheduling of ACRS project reviews.
Sincerel R. F. Fraley Executive Director, ACRS cc:
S. Chilk, SECY H. Denton, NRR /
C. Moon, NRR v D. Eisenhut, NRR R. Tedesco, NRR ACRS Members Allens Creek SSER #4 D-1
A OCT Y 1981 Mr. R. F. Fraley Executive Director i
Advisory _ Committee on Reactor Safeguards U. S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
POST ACCIDENT INERTING SYSTEM j
Dear Mr. Fralcy:
In your letter of September 16, 1981, same subject, you stated that the ACRS desires to review the proposed post accident inerting system (PAIS) for the Allens Creek plant before it is approved by the NRC staff or other means of hydrogen control are precluded.
L l
We plan to continue our review of this matter on the Allens Creek docket during the post-CP period. We believe the applicant will need about two years to complete the analyses and tests for a full assessment of the PAIS.
However, it may be beneficial for the staff and the ACRS to undertake an early review of the progress of work in this matter during the Spring of 1982.
We will work with the applicant to assure that such issues as hydrogen source l
term design criteria, system actuation criteria (including inadvertent actuation),
l containment pressurization, and equipment survivability, among others, can be
(
addressed at 'least in a preliminary manner by about March-April,1982.
We shall also propese conditioning the construction permit for the Allens Creek plant requiring the applicant to obtain NRC approval of the proposed PAIS before i
It proceeds with construction to such an extent as to foreclose the ability to effectively incorporate other means of hydrogen control. The staff will keep the ACRS informed as to the progress of its review of the proposed PAIS for the l
Allens Creek plant.
l Sincerely, Original Signed by Roger J. Mattson i
l Roger.J. Mattson, Director Division of Systems Integration
.0ffice of Nuclear Reactor Regulation cc: S. Chilk, SECY
- 11. Denton, NRR D. Eisenhut R. Vollmer Allens Creek SSER #4 D-2
l APPENDIX E MEMORANDA ON STATE AND LOCAL EMERGENCY PREPAREDNESS FOR ALLENS CREEK l
p:f:h j$
h' Federal Emergency Management Agency Washington, D.C. 20472 August 17, 1981 MEMORANDUM FOR:
Director Division of Emergency Preparedness Office of Inspection and Enforcement U. S, Nuclear Regul mission FROM:
Acting Director f.or Radiological Emerge Preparedness Divisiorr-
SUBJECT:
Status of State and Local-Emergency Preparedness for the Allens Creek Nuclear Generating Station (ACNGS)
This memorandum is in response to your letter dated May 21, 1981, invoking the NRC/ FEMA Memorandum of Understanding request that the Federal Emergency Management Agency (FEMA) review the status of State and local plans and preparedness for ACNGS.
i The State of Texas has not yet prepared any site specific plans for ACNGS.
They believe that since the facility is still being reviewed by NRC prior t
to the issuance of a Construction Permit (CP), and since there is no fuel s
present (or expected in the near future), there is no imminent potential threat to the public health and safety.
The State therefore, has set its
~
emergency preparedness priorities accordingly.
They have however, given consideration to some of the special problems that could arise that might have to be given extra attention during the licensing process.
Seasonal flooding is one of them which could make some of the roads in the plume j
exposure pathway impassable at times.
In order to mitigate this problem plans for alternate routes are being developed in the event of an accident ct the plant.
In addition part of Interstate 10 is located within the plume exposure pathway and carries heavy vacation related traffic.
Planning is also underway to provide reasonable assurance that traffic on Interstate 10 does not provide problems in the event that an accident at,ACNGS requires cctivation of State and local emergency procedures to protect persons in the 10-mile EPZ.
i Although no detailed plans are available, FEMA has determined that problems identified in this pre-construction permit stage of the NRC licensing process can be adequately managed under the REP process developed in i
accordance with 44 CFR 350 (proposed) and has set its emergency preparedness priorities accordingly.
Further, there is reasonable assurance that there ere no other anticipated major impediments or constraints that would inhibit State and local government preparedness in support of NGS.
Allens Creek SSER #4 E-1
y z
j Washington, D.C. 20472 OCT 8 19 81 E MORANDUM FOR:
Brian Grimes Director Division of Emergency Preparedness U. S. Nuclear Regulatory Commission FROM:
Mdrshall E. Sanders V
Acting Chief fff,[,2,/
Technological Hazards Division Office of Natural and Technological Hazards
SUBJECT:
Status of State and local Emergency Preparedness for the Allens Creek Nuclear Generating Station (ACNGS)
This memorandum provides clarification of our memorandum of August 17, 1981, subject as above.
In further discussion wit;h our FEMA Region, we perceive that the statements in this memorandum may have been construed as presenting more of a problem then we had intended.
The State of Texas has not yet prepared any site specific plans for ACNGS.
However, they anticipate no major impediments or constraints that would inhibit the normal REP process in Texas.
A staff member from the Houston Light and Power Company indicated the only potential problent might be the " seasonal" flooding.
However, recent flooding exp'eriences in the vicinity have shown that most highways and streets were past,able.
Further, if some roadway should become impassable, there are sufficient alternate ' evacuation routes to circumvent any potential problems.
The occasional heavy transient auto traffic from vacationers on Route 10 is a regular occurrence during non-emergency times that is handled effectively.
This should not be any additional problem during an emergency.
However, this aspect will be considered when the REP plans are prepared.
Therefore, on the basis of current information available, FEMA han determined that there is reasonable assurance that there are no unusual problems that can not be adequately handled as the site specific cmergency plans are developed.
ALLENS CREEK SSER E-2
' U.S. NUCLEAR REGULATORY COMMISSION NUREG-0515 BIBLIOGRAPHIC DATA SHEET.
Supplement No. 4 1 TITLE AND SUBTITLE #4dd Vo4me No.,if eprenaart
- 2. (Leave 6/m A)
S:foty Evaluation Report related to the construction of Allans Creek Nuclear Generating Station, Unit No. 1 3 RECIPIENT 3 ACCESSloN NO.
- 7. AUTHOR (S)
- 5. DATE REPORT COMPLETED MONTH l YE AR October L281
- 3. PERFORMING ORGANIZATION NAME AND MAILING ADDRESS (Include les Co*1 DATE REPORT ISSUED MONTH l YEAR U. S. Nuclear Regulatory Commission October-1981 Office of Nuclear Reactor Regulation
- s. ft,,,,,,.;
Waxhington, D. C.
20555
- 8. (Leave Nank)
- 12. SPONSORING ORGANIZATION NAME AND MAILING ADDRESS (lactude lip Codel
- 10. PROJECT / TASK / WORK UNIT NO.
Same as 9 above
- 11. CONTRACT NO.
13 TYPE OF REPORT PE RIOD Cove RE D (inclusive daars/
15 SUPPLEMENTARY NOTES 14 (Leave ole &)
Docket No. 50-466
- 16. ABSTR ACT 000 words or less)
The Office of Nuclear Reactor Regulation of the U.S. Nuclear Regulatory Commission has issued Supplement 4 to the Safety Evaluation Report (SER) for the application filed by Houston Lighting & Power Company for a construction permit (CP) to construct the Allens Creek Nuclear Generating Station Unit 1 (Docket No. 50-466). The Allens Creek site is in southern Austin County, Texas, about 45 miles west of Houston. This Supplement presents the staff's analysis of information submitted by the applicant to show compliance with the Commission's rule on emergency planning that was promulgated since SER Supplement 2 was issued. The report also updates the staff's reviews of financial qualification and generi c issues and reports on the staff's review of updated design information provided by the applicant.
The staff's analysis in Supplement 4 completes the staff's safety evaluation required before a decision is made on issuing a CP for the proposed facility. The staff concludes that a permit can be issued for the constr*. tion of Allens Creek Nuclear Generating Station Unit 1.
- 17. KE Y WORDS AND DOCUMENT ANALYSIS 17a. DESCRIPTORS 1itt IDENTIFIE RSJOPEN ENDE D TERMS.
18 AVAILABILITY ST ATEMENT
- 19. SECURITY CLASS (This report) 21 NO. OF P AGES Unclassified Unlimited 2o SECURITY CLASS (Th,s pavel 22 PRICE tw 1.saa4f4od S
NRC 5 oRV 335 47 71) j
.. -. ~..
... ~.. _
r UNITED STATE 3 l'
NUCLEAR REGULATORY CCMMIS$1CN F
]
W ASHINGTON, D. C. 20898 post A38 AND PEES PASO u.S. MUCLE AR NEGULATOmv opplCI AL suSINESS COMMstSION PEN ALTY FOR PRIVATE USE. $300 -
L J
i i
i 2
I i
f
- i 4
i i
f.-
t 2 AN 120555064215 US NRCADM DOCUMENT CONTROL DESA 3
4 POR i
i 316 DC 20555 W ASHI NGT CN o
I i
k
~
1-1 1
4 i
4 i
t 1
i 4
4 i
i i