ML20039F023
| ML20039F023 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 12/31/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0831, NUREG-0831-S01, NUREG-831, NUREG-831-S1, NUDOCS 8201110730 | |
| Download: ML20039F023 (62) | |
Text
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5 NUREG-0831 Supplement No.1 e
F Safety Evaluation Report related to the operation of Grand Gulf \\luclear Station, L
Units 1 and 2 fj Docket Nos. 50-416 and 50-417 j
Mississippi Power & Light Company Middle South Energy, Inc.
South Mississippi Electric Power Association
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e U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation December 1981 g
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Available from GPO Sales Program Division of Technical Information and Document Control U.S. Nuclear Regulatory Commission Washington, DC 20555 Printed copy pr$ce:
$4.75 and National Technical Information Service Springfield, VA 22161
NUREG-0831 Supplement No.1 Safety Evaluation Report related to the operation of Grand Gulf Nuclear Station, Units 1 and 2 Docket Nos. 50-416 and 50-417 Mississippi Power & Light Company i
Middle South Energy, Inc.
South Mississippi Electric Power Association U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation December 1981 pa, u
t TABLE OF CONTENTS P, age 1 INTRODUCTION AND GENERAL DISCUSSION...............................
1-1 1.1 Introduction.................................................
1-1 1.9 Outstanding Issues...........................................
1-1 1.10 Co n f i rma to ry I s s ue s.........................................
1-3 1.11 License Conditions...........................................
1-5 2 SITE CHARACTERISTICS..............................................
2-1 2.2.1 Transportation Routes.................................
2-1 2.5.5 Emb a n km e n t s a n d Dam s..................................
2-1 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS AND COMPONENTS............
3-1 3.7.1 Seismic Input.........................................
3-1 3.7.3 Seismic System Analysis...............................
3-1 3.8.1 Concrete Containment..................................
3-2 3.8.3 Other Category I Structures...........................
3-2 3.8.4 Foundations.........................................
3-3 3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment..........................
3-3 5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS......................
5-1 5.2.2 Overpressurization Protection........................
5-1 6 ENGINEERED SAFETY FEATURES........................................
6-1 6.2.1.8 Pool Dynamics................................
6-1 6.2.1.8.1 LOCA Pool Dynamics................
6-1 6.2.1.8.2 helief Valve Dynamics.............
6-2 6.2.2 Secondary Containment.................................
6-3 6.2.4 Containment Isolation System..........................
6-3 6.2.4.1 Containment Purge System.....................
6-4 6.2.6 Containment Leakage Testing...........................
6-5 7 INSTRUMENTATION AND CONTROL.......................................
7-1 7.8 Response to Inspection and Enforcement Bulletins and Other Safety Concerns......................................
7-1 i
TABLE OF CONTENTS (Continued)
Page 9 AUXILIARY SYSTEMS.................................................
9-1 9.5.6 Alternate Shutdown....................................
9-1 10 STEAM AND POWER CONVERSION SYSTEM................................
10-1 10.2.1 Turbine Wheel Integrity............................
10-1 13 CONDUCT OF OPERATIONS............................................
13-1 13.3 Emergency Preparedness Evaluation..........................
13-1 13.3.2 Evaluation of the Emergency Plan...................
13-1 13.3.2.1 Assignment of Responsibility (Organization Control)...................
13-2 13.3.2.2 Onsi te Emergency Organi zation............
13-2 13.3.2.4 Emergency Classification System..........
13-3 13.3.2.7 Public Information.......................
13-3 13.3.2.8 Emergency Facilities and Equipment.......
13-3 13.3.2.9 Accident Assessment......................
13-5 13.3.2.10 Protective Response......................
13-5 13.3.2.11 Radiological Exposure Control............
13-6 13.3.2.15 Radiological Emergency Response Training.
13-6 13.7 Security and Safeguards....................................
13-6 15 SAFETY ANALYSIS.................................................
15-1 15.1 Abnormal Operational Occurrences..........................
15-1 16 TECHNICAL SPECIFICATIONS.........................................
16-1 18 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS...........
18-1 22 TMI-2 REQUIREMENTS...............................................
22-1 II Siting and Design.............................................
22-1 II.B.2 Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment for Post-Accident Operation..
22-1 II.B.3 Post-Accident Sampling Capability.....................
22-1 II.E.4.2 Containment Isolation Dependability.........
22-4 II.F.1.5 Containment Water Level Monitor.............
22-5 ii
I l
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TABLE OF CONTENTS (Continued)
P. age j
II.K.1 IE Bulletins on Measures to Mitigate Small Break LOCAs and Loss-of-Feedwater Accidents.......................
22-6 II.K.1.23 Reactor Vessel Level Instrumentation.......
22-6 II.K.3 Final Recommendations of Bulletins and Orders Task Force.................................................
22-6 II.K.3.25 Effect of Loss on Alternating-Current Power on Pump Seals..............................
22-6 II.K.3.27 Common Reference Level.....................
22-6
- APPENDICES A.
CHRON0 LOGY......................................................
A-1 B.
ACRS REPORT.....................................................
B-1 C.
NRC UNRESO LVED SAFETY ISSUES....................................
C-1 D.
ERRATA..........................................................
D-1 E.
CONTROL ROOM REVIEW.............................................
E-1
-l F.
NRC REVIEW TEAM.................................................
F-1 iii
1 INTRODUCTION AND GENERAL DISCUSSSION 1.1 Introduction The Nuclear Regulatory Commission's Safety Evaluation Report (NUREG-0831) on the application by Mississippi Power and Light Company, Middle South Energy, Inc. and South Mississippi Electric Power Association (hereinafter referred to as the applicants) to operate the Grand Gulf Nuclear Power Station Units 1 and 2 was issued in September 1981.
The purpose of this supplement is to update our Safety Evaluation Report by providing the results of our review of infor-mation submitted by the applicant by letter or at meetings to address the outstanding issues identified in Sections 1.9 and 1.10 of the Safety Evaluation Report.
The information provided in these letters must be' acceptably documented in Amendments to the Final Safety Analysis Report prior to licensing.
As part of our review, we requested the applicant to verify that Grand Gulf meets the applicable regulations in 10 CFR Parts 20, 50, and 100.
The applicants responded to this request with a document, dated September 14, 1981 which summarizes compliance of design and operation of Grand Gulf to the applicable regulations set forth in 10 CFR Parts 20, 50, and 100.
I.
Since the preparation of the Safety Evaluation Report, the Advisory Committee on Reactor Safeguards considered the Grand Gulf operating license application at its 258th meeting and subsequently issued an interim report, dated October 20, 1981 to the Commission (See Appendix B of this report).
In that letter, Grand Gulf Unit 1, subject to specified recommendations and conditions, was found acceptable for operation at power levels up to 5% of full power.
A discussion of the ACRS letter can be found in Section 18 of this report.
Each section of this supplement is numbered and titled to correspond to the sectiens of the Safety Evaluation Report (SER) that-have been affected by our additional evaluation and, except where specifically noted, does not replace the corresponding section of the SER.
Appendix A is a continuation of the chronology and lists additional documents used in the supplemental review.
Copies of this report are available for public inspection at the Commission's Public Document Room at 1717 H Street, NW., Washington, D.C. and at the Hinds Jr. College, George H. McLendor Library, Raymond, Mississippi 39154.
Copies of this report are also available for purchase from the sources indicated on the inside front cover.
1.9 Outstanding Issues 4
In Section 1.9 of the Safety Evaluation Report, we identified 19 outstanding issues which were not resolved at the time of issuance of the Safety Evaluation Report.
In this report we discuss the resolution of a number of these items previously identified as open.
The items identified in Section 1.9 of the Safety Evaluation Report are listed below with the status of each item.
If the item is discussed in this supplement, then the section where the item is discussed in this report is identified.
The resolution of the remaining outstanding issues will be discussed in future supplements to the Safety Evaluation Report.
Grand Gulf SSER #1 1-1
Status Section 1
(1) Damping valve for cable tray design Resolved 3.7.3 (2) Ultimate containment capacity Resolved with 3.8.1 License Condition (3) Tangential shear - drywell Resolved 3.8.1, 3.8.4 (4) Hydrodynamic LOCA loads -
Awaiting further Mark III information (5) Load combination equations Partially Resolved 3.8.3 3.9.3 (SER)
(6) Electrical equipment Awaiting further qualification information (7) ODYN Code calculations Partially Resolved 5.2.2, 15.1 4.4.1 (SER)
(8) Containment isolation Partially Resolved 6.2.4 II.E.4.2 (SER)
(9) Containment purge Awaiting further information (10) Single failure in SRV low-low Resolved 7.8 setpoint function (11) Single sequencer reliability Under review (12) Nonsafety loads on emergency Under review sources (13) Management capability and Awaiting further organization information (14) Emergency preparedness plan Awaiting further information (15) Operating and emergency Awaiting further procedures information (16) Control room access and Under review instrumentation (17) Hydrogen ignitor system Awaiting further information (18) Reactor vessel level Resolved II.K.1.23 instrumentation Grand Gulf SSER #1 1-2
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(19) Common reference water level Resolved II.K.3.27 instrumentation 1.10 Confirmatory Issues In Section 1.10 of the Safety Evaluation Report, we identified 34 confirmatory
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issues which were not fully resolved at the time of issuance of the Safety Evaluation Report.
In this report we discuss the resolution of a number of these items previously identified.
The items identified in Section 1.10 of the
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Safety Evaluation Report are listed below with the status of each item.
If the item is discussed in this supplement, then the section where the item is discussed in this report is identified.
The resolution of the remaining con-f firmatory issues will be discussed in future supplements to the Safety Evaluation Report.
Status Section (1) Barge accident-toxic ammonia gas Resolved 2.2.1 (2) Meteorological measurements Under review I
(3) Ultimate heat sink (performance)
Under review
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(4) Seismic design (EHS vs. FES)
Awaiting further
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information (5) Seismic instrumentation Awaiting further information (6) Masonry wall design Awaiting further information (7) B0P piping - dynamic testing Awaiting further information
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(8)
Inservice testing of pumps and Resolved in SER 3.9.6 (SER) valves 4
(9) Seismic and LOCA loads Awaiting further information (10) Effects of New Madrid Fault extension Resolved 3.7.1 (11) Preservice inspection review Under review (12) Secondary containment leakage Resolved 6.2.2 limits verification
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(13) RHR and ECCS pump reliability Awaiting further information Grand Gulf SSER #1 1-3
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(14) Containment short-term and Resolved in SER 6.2.1.3 (SER) long-term pressure response (short-term only) 6.2.1.4 (SER)
(15) RHR/ containment spray cooling Awaiting further information (16) Safety-related display Awaiting further instrumentation information (17) Post-accident monitoring Under review (18) Response to IE Bulletin 79-27 Resolved in SER 7.8 (SER) with License Condition (19) Redundant safety-related electrical Under review systems (20) Control of heavy loads Resolved'in SER 9.1.4 (SER)
(21) Alternate safe shutdown panel Awaiting further information (22) Appendix R Resolved in SER 9.5.9 (SER)
(23) Noise level at working stations Under review (24) HPCS D/G reliability test report Awaiting further information (25) Operating shift work limitation Under review (26) Supplemental fire brigade training Under review (27) Physical security and safeguards Resolved with 13.7 plans License Condition (28) Low power test program Awaiting further information (29) Plant shielding for post-accident Resolved II.B.2 operation (30) IE Bulletins on measures to Awaiting further mitigate small-break LOCAs and information loss-of-feedwater accidents (31) ADS logic Awaiting further information (32) Actions for auxiliary heat removal Resolved in SER II.K.3.13 (SER) system with License Condition Grand Gulf SSER #1 1-4 l
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(33) Loss of power to pump seal coolers Resolved II.K.3.25 (34) Conformance to Commissien Resolved 1.1 regulations 1.11 License Conditions In Section 1.11 of the Safety Evaluation Report, we identified 25 license con-ditions. During our subsequent review, we have made changes to this list of license conditions.
Status Section (1) Embankment protection for culvert 1 Withdrawn 2.5.5 (17) Post-accident sampling capability Modified II.B.3 (26) Containment capacity Addition 3.8.1 (27) Turbine Disc Inspection Addition 10.2.1 (28) Security and Safeguards Addition 13.7 Grand Gulf SSER #1 1-5
2.
SITE CHARACTERISTICS 2.2.1 Transportation Routes In our September 1981 Safety Evaluation Report (SER) we expressed concern over the potential risk associated with the release of toxic gases, specifically ammonia, from a barge accident on the Mississippi River near the site. We stated that the applicants would be reo.uired either to install ammonia detectors in the control room air intakes which would automatically isolate the control room, or to provide assurance that the probability of an ammonia release which exceeded the toxicity limits in the control room is acceptably low (within the acceptance criteria of Standard Review Plan Section 2.2.3).
In a subenittal dated September 23, 1981, the applicants summarized the results of a survey made to determine the handling of and methods used in shipping ammonia, the quantities that are shipped, the frequency of shipments, and the regulatory restrictions involved.
Because of the lack of ammonia related accidents on the Mississippi River, the applicants' analysis of a serious accident involving an ammonia barge in the vicinity of the Grand Gulf Station took into account all types of accidents / spills occurring on the Mississippi between Baton Rouge, Louisiana and Cairo, Illinois, a distance of 726 miles, during a five year period between 1974 and 1979.
The topography of the area, the meteorological conditions in the vicinity of the site, and the portion of the river where an accident could pose a hazard to control room personnel were also considered in the analysis.
We have reviewed the applicants' analysis and, based on the conservative assumptions used, conclude that the risk to control room personnel from a barge accident on the Mississippi River involving ammonia is acceptably low and falls within the guidelines established in Standard Review Plan Section 2.2.3.
We, therefore, conclude that barge accidents involving the release of ammonia do not present an undue risk to the safe operation of the Grand Gulf Nuclear Station.
2.5.5 Embankment and Dams There are no earth, rock or earth and rockfill embankments used for flood protection or impounding cooling water for the operation of the plant.
In the SER, we stated that we intended to impose a license condition for embankment protection for Culvert No. 1.
Since then, the applicants performed slope stability analyses to demonstrate that during the occurrence of the Probable Maximum Flood (PMF), the slopes of the access road (at Culvert No. 1) and the drainage basin will remain stable and thus the possibility of blockage at Culvert No. 1 is not considered likely.
However, the provisions of Regulatory Guide 1.127, " Inspection of Water Control Structures Associated with Nuclear Power Plants," should be implemented by the Technical Specifications for both slopes as part of the plant start-up opera-tions and throughout the operational lifetime of this plant.
Therefore, a license condition as noted above will not be required.
Grand Gulf SSER #1 2-1
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3 DESIGN CRITERA FOR STRUCTURES, SYSTEMS, AND COMPONENTS
-1 3.7.1 Seismic Input c.
During our review of the Final Safety Analysis Report (FSAR) we requested that the applicants consider the effect on the Grand Gulf seismic design cf the pos-tulated extension of the New Madrid Fault Zone.
The applicants provided us with the design response spectrum and the New Madrid event spectrum.
Compari-
""- l son of the two spectra shows that the former envelops the latter.
Independently, 1-
the staff established that the only case when the New Madrid response spectrum
' 2 is higher than the Grand Gulf design spectrum is at 5 percent damping for the
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frequencies lower than one cycle per second, the difference being 5 to 10 percent.
Since the Grand Gulf structures all have resonant frequencies higher than two
- t cycles per second, the applicants concluded, and we concurred, that the responses derived from modes of lower frequencies are not important to the evaluation of Category I structures.
In view of the above, we consider this issue to be resolved.
3.7.3 Seismic System Analysis At an October 6, 1981, meeting with the staff, the applicants presented their reasons for using damping values for cable trays higher than those specified
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in the Regulatory Guide 1.61.
At the conclusion of the meeting the staff
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requested the following information which has since been submitted.
D ?;
-~e 1.
Design guide developed as a result of the tests.
,l.*-
2.
Typical cable trays seismic design calculations pertinent to Grand Gulf plant, illustrated by examples.
3.
Mathematical model for cable trays based on the tests.
if 4.
Presentation of tests data by grouping them according to the tray loadings.
y,. ;,
5.
Statements that damping values used for Category I cable trays seismic
'a c.
analysis were limited to 15 percent and that fireproofing spray was not used in the Grand Gulf plant.
The staff reviewed the material presented by the applicant related to the proposed damping ratios of cable trayways, including test results and design calculations, and concluded that it is acceptable.
This acceptance is based on the understanding that:
(a) Cable trays used at the Grand Gulf plant are of the same material and E
configuration as those used in the test program.
u (b) No fireproofing spray was used, which is consistent with the actual YJ test conditions.
,s Grand Gulf SSER #1 3-1
(c) Damping ratios used in seismic analysis of cable trayways were limited to 15 percent, which is conservatively justified by the tests.
Thus, the issue of cable tray design is resolved.
3.8.1 Concrete Containment We requested that an evaluation of the ultimate capacity of containment be made for the following conditions:
a.
The loads resulting from hydrogen burn; and b.
The ultimate static and dynamic pressure retaining capacity using a finite element model, taking into consideration the actual material strength vari-ations indicated by test certification and other variables.
The lowest acceptable pressure capacity of the containment is 45 psig.
In order to meet these criteria the applicants cnncluded that the following modifications of the upper personnel interlock are necessary:
1)
Modification or replacement of the inflatable seal.
2)
Strengthening of the bulkhead by means of stiffeners.
Additionally, we requested that the applicants provide us with the additional information regarding:
a)
Distribution curve related to yield strength of the reinforcement.
b)
Consideration of the effects of localized explosions on the containment.
c)
Analysis of the containment capacity at the junction of the cylinder and the base slab on the basis of the allowable liner strains.
In view of the fact that important information pertaining to the containment capacity is not available at this time and the applicants' expectation that the necessary modification of personnel interlock may not be completed in time for the plant start-up, we will require, as a license condition, that the con-tainment capacity be ascertained by analysis to be equal or greater than 45 psi, all required modifications be completed and the analytical information supporting this capacity be submitted for staff review four weeks prior to the operation of the plant above 5 percent power.
3.8.3 Other Seismic Category I Structures In order to ascertain that the Grand Gulf plant complies with the present staff design requirements, the applicants compared the load combination equations contained in the FSAR with the corresponding load combination equations in the Standard Review Plan (SRP) and identified one equation in the SRP as being more conservative than the corresponding equation in the FSAR.
Furthermore, the applicants analyzed several structural members using both equations and demon-strated that their required capacities are well below that provided in the design.
Grand Gulf SSER #1 3-2
In response to our request to justify selection of these members, the applicants
-stated that they are representative of highly stressed structural elements and-their selection was based on engineering reviews by the applicants.
We find this response of the applicants satisfactory and, based on the above described information provided by the applicants, conclude that this item is resolved.
3.8.4 Foundations Structual Audit During the structural audit we established that the containment' and drywell walls had not been designed in accordance with the current regulatory position and that the amount of the inclined reinforcing steel designed to carry
.the excess'of tangential shear stress above that resisted by concrete alone may be insufficient.
In response to our request the applicants presented a justification to account for this discrepancy.
In their justification the applicants demonstrated that in view of the recent tests on reinforced concrete samples' subjected to simul-taneous biaxial tensile loading as well as shear,-the allowable stress can be.
much higher than stipulated in the current regulatory position. Furthermore, the applicants demonstrated by calculations, using equations developed during the tests, that the design for tangential shear of drywell and containment walls is conservative.
In view of the applicants' presentation of the conservatism associated with the design, we conclude that both the containment and the drywell walls are adequately designed to resist the tangential shear and, therefore, is accep '
table. We consider this item to be resolved.
3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment In the Safety Evaluation Report, we concluded that in order to complete our review, we would require the applicants to provide additional information and to clarify the details of the qual.ification for s m pieces'of equipment.
In response to these concerns, the applicants provided a post-audit submittal on October 9, 1981. This submittal is currently under review by our Seismic Qualification Review Team (SQRT). Our concerns and the corresponding response contained in the above applicants' submittal are summarized below:
1)
Fatigue effect on equipment due to hydrodynamic loads was not adequately addressed. The applicants are required to examine some typical equipment under the hydrodynamic loads: A typical sample for NSSS equipment was provided in the applicants' submittal. The sample for a BOP equipment is currently being evaluated and the applicants are committed to submit the result and conclusion of the study for our review during November,1981.
2)
Provide assurance that retesting and redesign on Limotorque Motor Opera-tors for the hydrodynamic loading will be completed prior to fuel loading.
Provide a written confirmation when such retesting and redesign, as well Grand Gulf SSER #1 3-3
as installation have been completed:
An evaluation is currently being performed and the applicants are committed to provide response during November, 1981.
3)
Clarify the details of the qualification for some pieces of equipment as described in our October 22, 1981 trip report:
Responses to our concerns as stated in the trip report have been provided by the applicants in their October 9, 1981 submittal and are being reviewed by the SRT, with the excep-tion of the following items:
a)
Horizontal Fuel Transfer System Containment Closure (NSSS 4) b)
Control Room Panel (NSSS 7) c)
ASCO Solenoid Valve (BOP 14) d)
Reactor Core Isolation Coolant Turbine (NSSS 15)
The applicants are required to provide all the information necessary for the evaluation of the applicants' seismic and dynamic qualification program of equipment, as outlined above and in our trip report. When the applicants have provided the required information and have documented the completion of their seismic and dynamic qualification program, we will then complete our review and report on the results of our final evaluation in a future supplement to the Safety Evaluation Report.
Grand Gulf SSER #1 3-4
5 REACTOR COOLANT AND CONNECTED SYSTEMS l
5.2.2 Overpressurization Analyses Using ODYN In the course of our review of the Grand Gulf FSAR, we required that the applicants provide an overpressurization analysis of the main steam isolation valve closure event assuming failure of the first safety grade scram signal and operation of the recirculation pump trip.
By letter dated October 19, 1981, from L. F. Dale (Mississippi Power and Light) to H. R. Denton (NRC), the applicants forwarded the results of their analysis using the ODYN code. The peak vessel pressure calculated was 1260 psig which is below the limit of 1375 psig required by Article 9 of the American Society of Mechanical Engineers i
Boiler and Pressure Vessel Code - Section III.
This satisfies General Design l
Criterion 15 and is, therefore, acceptable.
I Grand Gulf SSER #1 5-1
1 e
~6 ENGINEERING SAFETY FEATURES 6.2.1.8 Pool Dynamic Loads t a The SER stated that the staff's review of the pool dynamic loads for Grand Gulf
, 'would be provided.in a supplement to the SER.
Several phenomena have been
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identified. in our' review of the Mark III containment that could result in -
dynamic loa C on trie structures located in and above the suppression pool.
These loads are related.to:
(1) pool response to the LOCA; and (2) pool responss dee td, relief' valve operation, generally associated with plant
' transient condit. ions.
These phenomena are described in more detail below.
6.2.1~.8.1
[0CAPoolDynamics Following a,.LOCA in~ the,drywell, the drywell atmosphere will be compressed due
, to blowdown cass and energy,sddition to the volume.
Following vent clearing, J
an ai?/ steam / water mixture will be forced from the drywell through the vent system and injected into thb suppression pool, approximately 7-10 feet below j
its surface.
The steam component of the flow mixture will condense in the pool,
' while the'airJcomponent will exist in the pool as high pressure bubbles.
The coatinued ' addition to and expansion of the air component causes the pool volume to swell resulting in an acceleration of the surface vertically upward.
Due to the effect of buoyaricy, the air bubbles will rise faster than the pool water mass'and will ventually break through the swollen surface and relieve the driving ~ force behind the' pool.
Due to the dynamics of vent clearing and vent flow and the vertical motion of the pool water mass, structures forming the suppression pool boundary, structures located within the pool, and structures located above the pool could be subject to hydrodynamic loads.
iheMARKIIILCCA-relatedpooldynamicloadswerereviewedattheCPstagefor Grand Gulf and at the PDA stage for GESSAR-238NI. The staff concluded at the time tnat the information available was sufficient to adequately define the pool dynsmic loads for nuclear plant applications at the construction permit stage of review..Since the issuance of the staff's SER for the GESSAR-238NI (December 1975), GE conducted further tests and analyses to confirm and refine the original load definitions.
These tests included further full scale Pres-sure Suppression Test Facility (PSTF) tests to investigate the phenomena of chugging, 1/3 area scale PSTF tests to investigate the phenomena of' steam con-densation, and 1/9 area scale PSTF tests to investigate multivent effects.
The staff has developed a generic program to review these additional tests and the refined load definitions provided by GE.
The results of this generic pro-gram, " Behavior of BWR Mark III Containment" (Task Action Plan 8-10), will be applicable to Grand Gulf.
However, the review of the generic load definition for Mark III containments will not be completed until February 1982.
T0 meet the accelerated schedule requirements for the licensing of Grand Gulf, the staff has proposed the use of bounding approaches for those areas in the generic pool dynamic loads definition where further staff review with GE needs to be accomplished.
These bounding approaches included increasing the pool-swell velocity to a conservative value of 60 ft/sec and also demonstrating that Grand Gulf SSER #1 6-1
increases in the condensation oscillation and chugging load specifications have no structural significance (i.e., other loads dominate the structures' design requirements). The applicants have provided the staff with analyses to show that the bounding approaches can be safely accommodated with minor changes in the hydraulic control unit (HCU) floor grating design.
The staff agrees with the applicants' assessment, except for the froth impact load specification on the solid concrete portion of the HCU floor. The HCU floor is approximately over the initial suppression pool surface and, follow-ing a DBA, vil.1 experience an impact load due to pool swell. The HCU floor is located above the area where the pool swell is characterized by a solid water ligament; thus, the impact.on the HCU floor is from the air-water mixture i
defined as froth. The proposed froth impact specification is 15 psid, and is based on PSTF roof f.roth impact measurements made by GE. At this time, the staff cannot conclude that 15 psid is a conservative specification for bound-ing pool swell velocities and breakthrough heights.
The staff is actively pursuing resolution of this issue and will report its conclusions in a future supplement to the SER. This matter must be resolved prior to low power-testing.
6.2.1.8.2 Relief Valve Dynamics Pressure waves are generated within the suppression pool when, on first open-ing, relief vc1ves discharge high pressure air and steam ii.to the pool water.
This phenomenon is referred to as relief valve vent clearing loads which are imparted to pools retaining structures and to structures located within the pool.
These same structures can also be subject to loads which accompany extended relief valve discharge into the pool if the pool water is at an ele-
~
vated temperature.
This effect is known as steam quenching vibrations.
The Mark III SRV-related pool dynamic loads were reviewed at the CP stage for Grand Gulf and at the PDA stage for GESSAR-238NI.
The staff concluded at the time that the information available was sufficient to adequately define the pool dynamic loads for applications at the construction permit stage of review.
This evaluation was presented in the SER for the GESSAR 238NI (December 1975) and in its Supplement 1 (September 1976).
Since then, GE has proposed three major changes related to the SRV pool dynamic lead definitions.
These include:
- 1) reduction in the load amplitude based on the Caors.o tests; 2) elimination of in phase bubble oscillation for equipment and containment internal struc-ture evaluation based on a Monte Carlo analysis; and 3) elimination of the possibility of multiple subsequent SRV actuation based on a low-low setpaint logic design.
The staff has completed its review of these proposed changes and finds them tc be acceptable. Our detailed evaluation will be provided in a NUREG report due to be issued in November 1981 (NUREG-0802, " Safety / Relief Valves - Load Evaluation Report Mark II and III containments").
In summary, the staff has completed its review of the pool dynamic loads for Grand Gulf 1/2, with the exception of the froth impact specification, and finds the load definitions to be conservative and acceptable.
The results of the staff's review of the f roth impact specification for Grand Gulf 1/2 will be reported in a future supplement to the SER.
Grand Gulf SSER #1 6-2
6.2.2 Secondary Containment Although the primary containment is enclosed by the secondary containment, there are systems which penetrate the boundaries of both the primary and secondary containments to create potential radioactivity release paths, in the event of a LOCA.
Radioactivity in the primary containment could, through these potential leakage paths, bypass the leakage collection and filtration systems associated with the secondary containment.
These leakage paths were identified as an unresolved safety concern in Section 6.2.2.4 of the SER.
The criteria by which these paths are evaluated are set forth in Branch Technical Position (BTP) CSB 6-3, " Determination of Bypass Leakage Paths in Dual Containment Plants." The applicants have used the guidelines of BTP CSB 6-3 in identifying all the potential bypass leakage paths.
The applicants have stated that these lines contain physical barriers or design provisions which can effectively eliminate leakage, e.g., water seals.
The water seal will be maintained for 30 days assuming a single active failure in any of the potential bypass leakage path.
Based on our review of the potential bypass leakage paths and our evaluation against the criteria specified in the BTP CSB 6-3, we find the design of the secondary containment system acceptable.
6.2.4 Containment Isolation System In our SER, we stated that there were several containment penetrations whose isolation provisions did not satisfy the explicit requirements of General Design Criteria 55 and 56, but that they could be found acceptable on some other defined basis.
The applicants have since justified these deviations from the explicit requirements of the General Design Criteria (GDC).
Most of those penetrations not satisfying the explicit requirements of the GDC were found acceptable on the basis of the alternative criteria specified in Section 6.2.4, item II of the SRP.
The applicants' application of these alternative acceptance criteria is summarized below:
1.
Lines that must remain in service following an accident and lines which should remain in service during normal operation for safety reasons are provided with at least one isolation valve.
A second isolation boundary is formed by a closed system outside the containment.
It should be noted that the containment isolation provisions for the RHR Heat Exchanger "B" and "A" Relief Valve Vent Header to the Suppression Pool consists of one check valve located outside containment and a closed system.
The applicants have stated that this valve arrangement provides vacuum relief protection for the relief valve discharge line penetrating containment when the RHR heat exchanger is isolated and, therefore, must be self-actuating to accomplish the design function. We agree with the applicants and have concluded that the isolation provisions for the RHR Heat Exchanger Relief Valve Header to the suppression pool is acceptable.
2.
Where a closed system outside the containment forms the second isolation boundary, each of the systems and all components which form its boundary
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Grand Gulf SSER #1 6-3
are designed to Quality Group B and seismic Category I standards. Valves which isolate the branch lines of these closed systems outside containment are normally closed and are under strict administrative control.
3.
On certain engineered safety features or related system, remote manual valves are used in lieu of automatic valves since these lines must remain in service following an accident. Where remote manual valves are used, leakage detection capabilities are provided.
4.
On some penetrations the containment isolation provisions consist of two valves in series both of which are outside the containment.
For these penetrations, locating one of the valves inside containment would subject it to more severe environmental conditions (including suppression pool dynamic loads); moreover, the inside valve would then not be easily acces-sible for inservice inspection.
Based on the above evaluation, we conclude that the applicants' proposed design for the containment isolation system satisfies the requirements of Criteria 54, 55, 56, and 57 of the GDC and is acceptable.
6.2.4.1 Containment Purge System In Section 6.2.4.1 of the SER, we stated that the applicants' plan to continu-ously purge the containment atmosphere during normal operation using both the 6-inch and 20-inch purge line during hot shutdown is not in compliance with BTP CSB 6-4, " Containment Purging During Normal Plant Operation." Since that time, the applicants have provided justification for purging but have estimated that they need to continuously operate the purge system to limit the buildup of airborne activity consistent with ALARA considerations. We firmly believe the purging must be minimized during normal reactor operation.
Therefore, we will require the applicants to provide a realistic estimate of hours per year that purging is expected to be needed through each particular purge valve.
In addition, the applicants have committed to provide administrative control over the 20-inch purge line so that both the 6-inch line and the 20-inch line cannot be operated at the same time.
As a result of numerous reports on the unsatisfactory performance of resilient seats in butterfly-type isolation valves due to seal deterioration, periodic leakage integrity tests of the 6-and 20-inch butterfly isolation valves in the purge system are necessary.
Therefore, we will require Technical Specifications for testing of the valves in accordance with the following frequency:
"The leakage integrity tests of the isolation valves in the contain-ment purge / vent lines shall be conducted at least once every three months."
The purpose of the leakage integrity tests of the isolation valves in the con-tainment purge lines is to identify excessive degradation of the resilient seats for these valves.
Therefore, they need not be conducted with the precision required for the Type C isolation valve tests in 10 CFR 50, Appendix J.
These tests would be performed in addition to the quantitative Type C tests required by Appendix J, and would not relieve the licensee of the responsibility to con-form to the requirements of Appendix J.
Grand Gulf SSER #1 6-4
Based on our review and subject to inclusion of the above Technical Specifica-tions requirements, we conclude that the applicant's purge system design is in compliance with the requirements of BTP CSB 6-4 and, therefore, is acceptable.
When the applicants have provided the required information or purging time, we will be able to complete our review.
6.2.6 Containment Leakage Testing We reviewed the applicants' containment leak testing program for compliance with the containment leakage testing requirement specified in Appendix J to 10 CFR 50.
Such compliance provides adequate assurance that the containment leak-tight integrity can be verified throughout the service life time and that the leakage rates will be periodically checked during service on a timely basis to maintain such leakage within the specified limits.
Maintaining containment within such limits provides reasonable assurance that, in the event of any radio-activity release within the containment, the loss of the containment atmosphere through potential leak paths will not be in excess of the limits specified for.
the site.
Specifically, we reviewed the containment leak testing program to assure that the containment penetrations and system isolation valve arrangements are designed to satisfy the containment integrated leak rate testing requirements and the local leak testing requirements of Appendix J.
Based on our review of the applicants' proposed leak testing program, we conclude that it meets the requirements of Appendix J to 10 CFR 50 and is acceptable.
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Grand Gulf SSER #1 6-5
3 7 INSTRUMENTATION AND CONTROL 7.8 Responses to Inspection and Enforcement Bulletins and Other Safety Concerns F.
CONSEQUENCES OF SINGLE FAILURES IN THE SAFETY RELIEF VALVE (SRV) - LOW-LOW SETPOINT FUNCTION In the SER, we indicated that this concern would be addressed later.
In our review, we determined that the instrumentation and control (I&C) portions of the design satisfy the single failure criterion as related to opening valves when required.
However, the design does not satisfy the single failure criterion with respect to preventing one additional valve from opening at the time only one valve is permitted to be open (on second and subsequent valve pops).
Inherent in the design of the Grand Gulf SRV system, which uses relay logic, are single failure points which could result in th inadvertent opening of a second valve.
A second valve opening could also be caused by a mechanical failure of any of the remaining valves.
4 As a result of this finding, we requested the Grand Gulf applicants and General Electric Company to demonstrate that the present design is adequate.
In response to our concern, the applicants and General Electric Company provided a reliability analysis which established the probability of a second valve to t
open simultaneously with the low-low or mid-low set valves during the interval of concern as a result of a single failure. This analysis has been evaluated by the Reliability and Risk Assessment Branch.
Their results were in substan-tial agreement with the General Electric results and show that the probability of a second valve to open as a result of a single failure is in the range of 10 3 -10 6/ year as compared to the General Electric value of 10 6/ year.
The difference in the results comes from the inclusion of the potential for cali-bration and drift errors considered by the staff which were not included in the General Electric analysis.
i f
In either case, the potential for a second valve to open as a result of a single failure during the short intervals of concern is very low.
In addition, the Generic Issues Branch has evaluated the consequences of simul-taneous opening of two relief valves on second and subsequent valve-pops and has found that the loading to the containment is within acceptable limits. The results of their evaluation will be documented in NUREG-0802, " Safety / Relief Valve Load Evaluation Report - Mark II and III Containment," scheduled for publication in November 1981. We, therefore, conclude based on, (1) the low probability a second valve opening simultaneously with the low-low set valve due to a single failure, and (2) the staff's evaluation that the consequences are acceptable, that the design of the SRV low-low setpoint control system is acceptable.
Grand Gulf SSER #1 7-1
9 AUXILIARY SYSTEMS 9.5.6 Alternate Shutdown By letters dated August 7, 1981 and August 27, 1981 the applicants provided 3
responses to our concerns involving fire protection for the safe shutdown capability in accordance with the requirements of Appendix R.
Further discus-sion of the safe shutdown capability including information on cable separation and safe shutdown equipment location is contained in FSAR Section 9A.
The applicants' safe shutdown analysis states that systems needed for hot shutdown and cold shutdown are redundant and that one of the redundant systems needed for safe shutdown would be free of fire damage, by providing separation, fire barriers and/or alternative shutdown capability.
For hot shutdown, at least one of the following shutdown systems would be available:
(1) the Reactor Core Isolation Cooling System (RCIC), (2) the High Pressure Core Spray System (HPCS), and (3) a combination of the pressure relief system - Automatic Depressurization System (ADS), the Low Pressure Core Spray System (LPCS), and Residual Heat Removal (RHR) System.
For cold shutdown, are appropriate portion of the RHR system wouid be available.
The safe shutdown analysis considered components, cabling and support equipment for systems identified above which are needed to achieve shutdown. The applicants have provided a cable separation analysis for all rooms of the plant housing safe shutdown equipment to ensure that at least one train of this equipment is available in the event of a fire in any of these rooms.
The analysis identifies the safety-related equipment, redundant safe shutdown system cabling and discusses the consequences of a fire in each of these rooms. We have reviewed the above method and conclude that it is an acceptable means of demonstrating that separation exists between redundant safe shutdown system trains.
The applicants' analysis indicated that alternative fire protection measures were required in certain control building corridors in order to assure the availability of the safe shutdown system because they contained more than one division of safe shutdown system cabling. The applicants have also indicated that alternative shutdown is required for the control room.
In the event that fire disables the control room, two remote shutdown panels (one for each division of redundant safe shutdown systems) located in separate fire protected rooms in the control building are provided as an alternative to providing fire protection (refer to Section 7.4.2 of the SER).
Only the Division 1 panel is electrically isolated from the control room.
Based on the above, we conclude that the shutdown capability of the Grand Gulf Nuclear Station complies with the requirements of Section III.G of Appendix R and is, therefore, acceptable.
Section 7.4.1.4 of the FSAR describes the remote shutdown panel's design and capability. The present design objective of the remote shutdown panels (one panel for each division of redundant safe shutdown systems) is to achieve and maintain cold shutdown in event of an evacuation due to a fire disabling the Grand Gulf SSER #1 9-1
control room.
The Reactor Core Isolation Cooling (RCIC) system, safety / relief valves and one division of the Residual Heat Removal (RHR) system can be controlled from one remote shutdown panel to achieve cold shutdown should a fire disable the control room.
In order to assure the availability of this remote shutdown panel in the event of a control room fire, in a letter dated August 7, 1981, the applicants committed to provide electrical isolation between the control-room and the remote shutdown panel prior to startup from the first regularly' scheduled refueling outage. We find this commitment acceptable.
However, the applicants should provide the details of this design modification to the NRC for review as soon as they become available, but no later that 4 months prior to the first refueling outage.
The design of the remote shutdown panel complies with the performance goals outlined in the requirements of Section III.L of Appendix R.
Reactivity control will be accomplished by a manual scram before the operator leaves the control room.
The RCIC system will provide reactor coolant makeup, and the RHR system and the safety / relief valves will be used for reactor decay heat removal.
Reactor vessel water level, reactor vessel pressure, suppression pool water level and temperature, RCIC pump turbine speed and RHR system flow are among instrumentation to be available at the remote shutdown panel to provide direct reading of process variables.
The remote shutdown panel will also include instrumentation and control of support functions needed for the shutdown equipment.
Based on the above, we conclude that the remote shutdown panel complies with the requirements of Section III.L of Appendix R, pending approval of the electrical isolation design by the NRC prior to the first refueling outage, and is therefore, acceptable.
Grand Gulf SSER #1 9-2
10 STEAM AND POWER CONVERSION SYSTEM 10.2.1 Turbine Disc Integrity The LP discs in the Allis-Chalmers turbine at Grand Gulf were manufactured by Siemens of Germany using a steel with above average yeild strength.
- Recently, several cases of cracking in Siemens LP discs at fossil fuel plants have been reported. We have requested information to assess the potential for turbine disc cracking and to establish an acceptable inspection schedule.
Specifically, the information requested was in regard to operating temperatures and lubricants used during assembly.
Because the applicants may require substantially more time to gather the infor-mation necessary to justify long term operations, we have taken an interim con-servative approach by assaming that the propensity for disc cracking is the same as that of Westinghouse turbines.
Therefore, we expect to impose a condi-tion to the Grand Gulf Unit 1 license which will require that the bores and keyways of the LP turbine discs be inspected for ultrasonic indications of cracking during each refueling outage.
The inspection schedule may be adjusted after licensing if the responses to our requests are favorable.
Grand Gulf SSER #1 10-1
=
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f 13 CONDUCT OF OPERATIONS 1
/
13.3 Emergency Preparedness Evaluation
.The staff's evaluation of the applicants' emergency plans is provided in Sec-tion 13.3 of the SER. The Grand Gulf Nuclear Station Radiation Emergency Plan 1
(Plan) as amended (Revision 1, April 1981), was reviewed against the. require-ments of 10 CFR Section 50.47(b),' Appendix E to 10 CFR 50, and the criteria of i
the 16 Planning Standards in Part II of the " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," NUREG-0654/ FEMA-REP-1, Rev. 1, dated November 1980.
]
In the SER, the staff specifically identified 16 items requiring resolution for which additional information and commitments were to be provided by the applicants.
~
The applicants have provided the staff with additional information and commit-ments in response to the open items.
In a submittal dated August 25, 1981, MP&L provided a response to each ites and proposed changes to the Plan.
MP&L committed to incorporating the proposed changes in' the next available revision 3
to the Plan. The applicants' responses to the 16 identified items have been 4
evaluated and are discussed in this supplement.
With regard to offsite emergency preparedness, the State of Mississippi Emer-gency Operations Plan, Volume V - Radiological Emergency Response Plan and the State of Louisiana Peacetime Radiological Response Plan have been formally submitted to the Fedaral Emergency Management Agency (FEMA). Regions IV and VI, respectively, for review by t.Ra Regional Assistance Committees (RAC IV and VI).
The plans for Claiborne County, Mississippi and Tensas Parish, Louisiana are also under review by RAC IV and VI, respectively.
FEMA will review these plans in accordance with its Memorandum of Understanding with the NRC. A joint emergency response exercise was performed November 3-4, 1981, to determine the onsite and offsite emergency response capabilities. The NRC critique was held on November 4 after the exercise, and a public critique, chaired by FEMA, was l
held on November 5.
The final NRC approval of the state of emergency preparedness for.the Grand i
4 Gulf site will be made following implementation of the site emergency plans to include development of procedures, training and qualifying of personnel, installation of equipment and facilities, and a favorable finding by FEMA on t
the adequacy of State and local plans.
13.3.2 Evaluation of Applicants' Emergency Plan The applicants' responses to the 16 items previously identified by the staff i
~ as requiring additional information and commitments have been evaluated and are discussed below.
(The order of presentation corresponds to the listing of unresolved items that appears in Section 13.3 of the SER.)
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Grand Gulf SSER #1 13-1
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13.3.2.1 Assignment of Responsibility (Organization Control)
Identify the agency or agencies having jurisdiction over the traffic on that portion of the Mississippi River that lies within the 10-mile EPZ.
Provide written agreements with the said agency or agencies.
Discussion and Conclusion Section 5.8.5 of the Plan specifies tr.at the U.S. Coast Guard will exercise its authority'to control marine traffic on the Mississippi River through the establishment of a safety zone in the immediate area.
In a Memorandum for the Record from J. E. Maher, Director of Mississippi Emergency Management Agency (MEMA), dated August 18, 1981, the safety zone is described as encompassing the predetermined 10 mile EPZ (mile 393 to 424), and based on the situation may be adjusted to any size configuration.
Letters from MEMA to the U.S.
Coast Guard, dated October 30, 1980 and August 17, 1981, describe the details of the notification arrangements between MEMA and the U.S. Coast Guard.
Based on our review of their Plan and submittal as outlined and discussed above, we find that the applicants have provided an acceptable response to this item; provided the specification of the safety zone and the details of the notification process are incorporated into the appropriate emergency plan implementing procedures.
13.3.2.2 Onsite Emergency Organization Provide shift staffing and augmentation capabilities that meet the specific staf fing guidance expressing in Table B-1 of NUREG-0654.
Discussion and Conclusion The applicants' response states that GGNS will have shift personnel who will be trained in the techniques of basic chemistry analysis, and that the station will comply with the requirements of NUREG-0737, with respect to the capability to promptly quantify (within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) certain radionuclides that are indicators of the degree of core damage.
The response also states that while the majority of the GGNS staff live at distances which precludes complete shift staff augmentation within 30 and 60 minutes, several key emergency positions could be filled within 30 minutes of emergency declaration by persons living closer to GGNS.
Additional information is required as follows:
Identify the individual (s) on shift who will perform the duties of the Chem / Rad technician during an emergency, and describe the radiochemistry training program that will be provided to these persons.
Identify the key emergency positions that can be filled within 30 minutes of emergency declaration, and describe what compensating actions will be taken for those positions that will not be filled within 30 and 60 minutes, as expressed in Table B-1, NUREG-0654.
Grand Gulf SSER #1 13-2
13.3.2.4 Emergency Classification System Make minor clarifications to the classification and EAL section of the Plan.
Discussion and conclusions The applicants have committed to include, in the next revision to the Plan, those items identified in the SER related to initiating conditions for the Unusual Event and Alert Emergen;y classifications.
Based on our review of their commitment as discussed above, we find that the applicants have provided an acceptable response to this item.
13.3.2.7 Public Information The Public Information Program currently under development by the applicants must include provisions for the annual dissemination of information to the public.
Discussion and Conclusions The applicants have committed to including, in the next revision to the Plan, provisions for annual dissemination of information to the public.
Based on our review of their commitment as discussed above, we find that the applicants have provided an acceptable response to this item.
13.3.2.8 Emergency Facilities and Equipment (1) A description of the full meteorological monitoring program that meets the criteria of Appendix 2 to NUREG-0654 must be provided.
Discussion and Conclusion The applicants have committed to include, in the next revision to the Plan, additional information on the meteorological (met) system, including a description of the back-up met system.
The new meteorological information is currently under review.
Following I
the review of the final met program, the staff will provide its conclusions as to acceptability in a future supplement to the SER.
(2) The provisions for acquiring data from offsite geophysical monitoring systems must be provided.
Discussion and Conclusion The applicants have committed to include, in the next revision to the Plan, provisions for acquiring data from offsite seismic and hydrologic monitoring systems.
Grand Gulf SSER #1 13-3
Based on our review of the commitment as outlined and discussed above, we find that the applicants have provided an acceptable response to this item.
(3) The lists of emergency equipment for the Security Guardhouse and EOF (Plan Appendix B) must include protective clothing and respirators as required by Section 7.10 of the Plan.
Discussion and Conclusions The applicants' response states that respirators / masks are not a necessary part of the Guardhouse equipment inventory.
If the Guardhouse becomes uninhabitable, the guards would be evacuated to the Site Access Point (SAP) located about 0.5 miles from the reactor.
Personnel entering the site following the declaration of an emergency would pass through the SAP and would obtain protective clothing and respirators at that point, if conditions warranted. The response also states that respirators required at the EOF would be transported from the Corporate Emergency Center upon receipt of an order for EOF activation.
The SAP is located in the same building that contains the interim EOF.
Thus, the SAP is the central point for entry, exit, and re entry into the station, and will contain a full complement of emergency supplies and equipment.
The applicants' proposed change to the Plan deletes the provision for storage of respirators / clothing at the Guardhouse and interim EOF.
Based on our review of their Plan and submittals as discussed above, we find that the applicants have provided an acceptable response to this iten.
(4) Clarification of Section 7.3.1 of the Plan is required to provide assurance that the E0F will be operational prior to fuel load, and that the EOF will be staffed and operational in accordance with the recommended guidance contained in NUREG-0654, Table B-1.
Discussion and Conclusions The applicants' submittal provided partial response to the open item.
The Plan specifies that the EOF will become operational in accordance with the implementation dates set forti. in NUREG-0737.
Reference is made, in the Plan, to an interim E0F, which will be operational prior to fuel load.
The applicants' submittal did not address the item with regard to staffing and activation of the EOF.
Based on our review of their Plan and submittal as discussed above, we find that the applicants have provided an acceptable response to the-matter of establishing an interim EOF prior to fuel load. The staffing and activation of the EOF in accordance with the recommended guidance contained in NUREG-0654, Table B-1, remains an open item.
Grand Gulf SSER #1 13-4 a
(5) The Plan must be modified to assure that the Technical Support Center.(TSC) will be functional within 30 minutes of activation consistent with the guidance specified in Table B-1 of NUREG-0654, Rev. 1.
(6) Center (OSC) in a more timely manner consistent with the emergency
_ staffing augmentation guidance contained in Table B-1 of NUREG-0654, Rev. 1.
Discussion and Conclusions The applicants' submittal did not address the items pertaining to staffing and activation of the TSC and OSC consistent with the emergency staffing augmentation guidance contained in Table B-1 of NUREG-0654, Rev. 1.
Therefore, response to the SER open items must be provided by the applicant for review by the staff.
13.3.2.9 Accident Assessment The capability for detection and measurement of radioiodine, in air, in the plume EPZ as low as 10 ' pCi/cc, under field conditions, must be described.
Discussion and Conclusions The applicants have committed to include, in the next revision to the Plan, provisions for measurement of radiciodine concentrations, in the field, down to 1 x 10 7 p Ci/cc.
Based on our review of their commitment as discussed above, we find that the applicants have provided an acceptable response to this item.
13.3.2.10 Protective Response (1) Evacuation time estimates, currently under review by the staff, must satisfy the criteria of Appendix 4 to NUREG-0654.
(2) Maps showing evacuation areas and routes in the 10-mile EPZ, that conform to the specifications in Appendix 4 to NUREG-0654, must be provided.
Discussion and Conclusions A new, computer-modeled, evacuation time estimate study performed for the 10 mile EPZ surrounding the site was submitted to the NRC on July 29, 1981, as revision 2 to the Plan.
The applicants' response states that the maps contained within the study comply with the guidelines of Appendix 4 to NUREG-0654.
f The additional information provided by the revised Appendix E to the Plan is under review by the staff.
A future supplement to the SER will provide the staff's conclusions as to the adequacy of the evacuation time estimate study.
Grand Gulf SSER #1 13-5
(3) The local protection afforded in residential units or other shelter for direct and inhalation exposure shall be included in the bases for choice of recommended protective actions in Sections 6.5.1.2 and Table 6-1 of the Plan.
Discussion and Conclusions 4
The applicants have committed to include, in the next revision to the Plan, a description of the local protection afforded in residential units in the bases for choice of recommended protective actions.
Based on our review of their commitment as discussed above, we find that the applicants have provided an acceptable response to this item.
13.3.2.11 Radiological Exposure Control Procedures must be established for contamination control measures with regard to drinking water and food supplies; action levels for decontamination of materials; and criteria for permitting return of areas to normal use.
Discussion and Conclusions The applicants have committed to include, in the next revision to the Plan, provisions for onsite contamination control measures with regard to drinking water and food supplies; action levels for decontamination of materials; and criteria for permitting return of areas to normal use.
Based on review of their commitment as discussed above, we find that the applicants have provided an acceptable response to this item.
13.3.2.15 Radiological Emergency Response Training The training and annual retraining for the emergency communicators (Auxiliary Operators) and annual retraining for those offsite support agencies identified in Section 8.2.c of the Plan must be described.
Discussion and Conclusions The applicants have committed to include, in the next revision to the Plan, a description of the training and retraining to be provided emergency communi-cators and offsite support agency personnel identified in Section 8.2.c of the Plan.
Based on our review of their commitment as discussed above, we find that the applicants have provided an acceptable response to this item.
13.7 Security and Safeguards The applicants have submitted the " Grand Gulf Nuclear Station Physical Security Plan," " Grand Gulf Nuclear Station Security Training and Qualification Plan" and the " Safeguards Contingency Plan for Grand Gulf Nuclear Power Station" for the protection of Unit 1 from acts of radiological sabotage.
Grand Gulf SSER #1 13-6
The " Grand Gulf Nuclear Station Physical Security Plan," " Grand Gulf Nuclear Station Security Training and Qualification Plan" and " Safeguards Contingency Plan for the Grand Gulf Nuclear Power Station" have been determined to meet the requirements of 10 CFR 73 and; with one exception, are acceptable.
The one exception is the basis for the following license condition:
The applicants shall develop and employ, in accordance with 10 CFR 73.55(g)(1), compensatory measures for the failure of all alarms, communication equipment, physical barriers, and other security related devices or equipment to assure that the effec-tiveness of the security system is not reduced by failure or other contingencies affecting the operation of the security related equip-ment or structures.
The identification of vital areas and measures used to control access to these areas, as described in the security plan, may be subject to amendment in the future based on a confirmatory evaluation of the plant to determine those areas where acts of sabotage might cause the release of radionuclides in sufficient quantities to result in dose rates equal to or exceeding 10 CFR 100.
The applicants' plans are being withheld from public disclosure in accordance with the provisions of Section 2.790(d) of 10 CFR 2.
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Grand Gulf SSER #1 13-7 l
15 SAFETY ANALYSIS 15.1 Abnormal Operational Occurrences In our review of transient analyses for Grand Gulf, we required that the applicants provide analyses of pressurization transients using the ODYN code.
By letter dated October 19, 1981, from L. F. Dale (Mississippi Power and Light) to H. R. Denton (NRC), the applicants forwarded the requested analyses.
The applicants' analysis inputs are consistent with the requirements of the Stan-dard Review Plan.
The results indicate that the most limiting pressurization event is the Feedwater Controller Failure to Maximum Demand Transient result-ing in a minimum critical power ratio of 1.08.
The specified fuel design limit for these events is that the minimum critical power ratio not fall below 1.06.
General Design Criterion 10 is, therefore, satisfied.
This is acceptable to us.
l Grand Gulf SSER #1 15-1
16 TECHNICAL SPECIFICATIONS As discussed in this supplement, we have identified additional issues which must be included in the Technical Specifications as a condition of our acceptance.
These issues are listed below and are discussed further.in the sections of this report as indicated:
(13) Provisions of Regulatory Guide 1.127 (2.5.5)
(14) Containment leakage test frequency (6.2.4.1)
Grand Gulf SSER #1 16-1
t 18 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS A Subcommittee of the Advisory Committee on Reactor Safeguards (Committee) considered the application for operating licenses for Grand Gulf Nuclear Station Units 1 and 2 at a meeting in Jackson, Mississippi on September 17-18, 1981.
The Subcommittee toured the facility on September 17, 1981. The full Committee reviewed tne application at its 258th meeting on October 15, 1981.
A copy of the Committee's report to the Chairman of the Nuclear Regulatory Commission dated October 20, 1981 is attached as Appendix B.
The Committee's report concluded that, subject to satisfactory completion of construction, staffing, and preoperational testing and with due consideration to recommendations contained in the report, operation of Unit 1 up to 5% of full power would be acceptable.
The Committee has not completed its review of Unit 1 for full power operation pending additional progress in resolving the outstanding issues of (a) dynamic loads on structures above the Mark III suppression pool due to froth impact and (b) hydrogen control. The report, noting that construction of Unit 2 had been temporarily suspended, did not evaluate the licensing of that unit.
The Committee has recommended that the MP&L Nuclear Safety Review Board include two or more experienced members with appropriate backgrounds from outside MP&L.
We have been orally informed by MP&L of arrangements that have been made to meet this recommendation.
A formal letter on this matter is expected in early December. We will review the contents of that letter when received and report our evaluation in a future supplement to the SER.
The Committee stated that, except for the two issues noted above, it was satisfied with the progress on the topics identified in the SER as outstanding issues, confirmatory issues, and license conditions, and indicated that they should be resolved in a manner satisfactory to the NRC staff.
Sections 1.9 and 1.10 of this report update the status of the outstanding and confirmatory issues, respectively, and Section 1.11 reports on changes to the list of matters that are expected to require license conditions.
As shown in Sections 1.9 and 1.10, six of the 19 previous outstanding items have been resolved, one requiring a license condition, and 12 of the 34 confirmatory issues have been resolved, three requiring a license condition.
Efforts are continuing to resolve the remaining topics on a schedule consistent with completion of the plant and in a manner satisfactory to the NRC staff.
Grand Gulf SSER #1 18-1
22 TMI-2 REQUIREMENTS The SER is supplemented for specific TMI-2 requirements as noted below.
II Siting and Design II.B.2 Plant Shielding to Provide Access to Vital Areas and Protect Safety Equipment for Post-Accident Operation Discussion and Conclusions In the SER, we stated that we found the review acceptable subject to receipt of the final list of vital areas, the personnel occupancy study showing pro-jected doses to be within GDC-19 criteria and description of any necessary design or administrative control changes.
The information specified in NUREG-0737 was provided.
The applicants specified vital areas, requiring post-accident access included the control room and technical support center for continuous occupancy; remote shutdown panel, motor control center (CRD RPU temperature recorder panel and B0P computer multiplexing and isolation panels area) diesel generator build-4 ings, post-accident sampling station and laboratories, for extended periods.
The applicants did not find it necessary to make any design changes during their shielding review.
The applicants have stated that access to the radwaste panel is not required during or immediately following an accident and was not included in the review.
MP&L has provided a person-rems, time and personnel occupancy study showing doses to workers for operations in vital areas to be within GDC-19 criterion.
We conclude that Grand Gulf has met the requirements of NUREG-0737 item II.B.2 for post-accident access and therefore is acceptable.
II.B.3 - Post Accident Sampling Capability Discussion and Conclusions By Amendment 49 and letters dated July 21, August 25, and October 23, 1981,
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the applicants provided a description of systems, equipment, and procedures to be used for sampling the reactor coolant, suppression pool, and drywell following an accident resulting in core degradation.
The applicants have committed to a post accident sampling system that meets the requirements of Item II.B.3 in NUREG-0737.
The post-accident sampling system is an integral part of Process Sampling.
Compliance with the license conditions proposed in our SER input of August 13, 1981 is shown below:
Condition 1:
Compliance with the requirements of NURG-0737, Item II.B.3 are presented in the following subsections:
Grand Gulf SSER #1 22-1
a.
The system is capable of obtaining and analyzing reactor coolant and containment atmosphere samples within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> from the time a deci-sion is made to take a sample.
b.
Measurement of chloride ions, conductivity, pH, dissolved oxygen 3d hydrogen and gamma activity can be performed using in-line monitorn.g.
c.
Reactor coolant and containment atmosphere sampling during post-accident conditions does not require an isolated auxiliary system to be placed in operation in order to use the sampling system.
d.
The portion of the system that measures dissolved gases in the reac-tor coolant consists of in-line measurement of hydrogen concentration in the noncondensible gases, determination of the ratio of liquid te noncondensible gases, and analysis for dissolved oxygen.
e.
Chloride analyses can be performed using in-line monitoring, therefore, the chloride analyses can be obtained within the required 24-hour period.
f.
The post-accident sampling station is designed to provide adequate radiation protection so that it is possible for an operator to obtain and analyze a sample without radiation exposures exceeding the criteria of GDC 19, assuming source terms given in Regulatory Guide 1.3, " Assumptions Used for Evaluating the Potential Radiologi-cal Consequences of a Loss of-Coolant Accident for Boiling Water Reactors."
g.
Grab samples of primary coolant can be obtained and analyzed on site for boron within three hours.
A commitment to this item is contained in a letter from MP&L dated November 24, 1981.
h.
The system is designed for in-line monitoring with grab sampling as a backup.
The equipment provided for backup sampling is capable of providing at least one sample per day for 7 days following onset of the accident and at least one sample per week until the accident condition no longer exists.
i.
The radiological and chemical analysis capability of the system includes provisions to identify and quantify the isotopes of the nuclide categories of concern to levels corresponding to the source terms given in Regulatory Guides 1.3 and 1.7 (R.G. 1.7, Control of Combustible Gases Concentrations in Containment Following a Loss-of-Coolant Accident).
The gamma detection system allows the monitoring of the reactor coolant activity over a range of 10 7 to 101 Ci/cc and the contain-ment atmosphere activity over a range of 10 8 to 10 1 Ci/cc during normal and accident conditions.
The sample station is designed to restrict background levels of radiation such that the sample analysis provides results with an acceptably small error.
Grand Gulf SSER #1 22-2
J.
The post-accident sampling system instrumentation is designed to cover adequate ranges, accuracies, and sensitivities to allow the operator to obtain pertinent data to describe the radiological and chemical status of the reactor coolant system.
k.
Reactor coolant sample lines are of a diameter such that the rupture of a sample line will limit reactor coolant loss.
The ventilation exhaust from the sampling station is filtered with a charcoal absorber and HEPA filter.
Condition 2:
Sufficient. shielding is provided to make it possible for an operator to obtain and analyze a sample with radiation exposures meeting the requirements of GDC-19, assuming Regulatory Guide 1.3 source terms.
Condition 3: The applicant will comply with the detailed requirements of Regulatory Guide 1.97 or provide adequate justification on a case-by-case basis for the use of an alternate approach.
Condition 4:
Prior to attaining 5% power operation electrically powered components associated with post-accident sampling will be capable of being supplied with power and operated within thirty minutes of an accident in which there is core degradation assuming loss of off site power.
A commitment to this item is contained in a letter from MP&L dated November 24, 1981.
Condition 5: The inboard isolation valves which are inaccessible for repairs after an accident are environmentally qualified for operation as containment isolation valve and are capable of being opened with a reliable power supply in the event of a loss of off site power.
Condition 6: An interim failed fuel estimation procedure has been submitted by the applicant.
A final procedure is being devised by General Electric for the fuel type being used at Grand Gulf and will be available by August 1982.
We are now reviewing this interim procedure and will submit our evaluation when we complete the review.
This item remains open.
Condition 7: The Grand Gulf post-accident sampling system does not use high pressure carrier gas which could be injected into the reactor coolant system.
l Condition 8:
The reactor coolant dissolved oxygen level can be verified to be
< 0.1 ppm since the instrument meets the range identified in Regulatory Guide 1.97.
Condition 9: The post-accident sampling system will be used to perform at least monthly reactor coolant sample analyses for gamma isotopic, chloride, conductivity, pH, oxygen and hydrogen.
Every six months, for training and operability testing, a diluted liquid grab sample will be drawn, transported, and analyzed in the Hot Lab for boron.
In addition, every six months, a containment air sample will be analyzed for hydrogen, oxygen, and gamma spectrum.
Classroom training will also be provided on system operation and the proper handling of highly radioactive samples.
Based on the above evaluation, we determined that the provisions in the proposed post-accident sampling system when completed, will partially meet the Grand Gulf SSER #1 22-3
l l
requirements of Item II.B.3 in NUREG-0737. However, the applicants have not provided adequate information to demonstrate that the reactor coolant system sample and suppression pool sample locations are representative of core conditions.
No information has been provided on the type of in-line instru-ments and analytical chemistry procedures and on data supporting their applicability (accuracy and sensitivity) in the post-accident environment.
To that effect, the license conditions stated below are proposed.
Implementation of all of the requirements of Item II.B.3 in NUREG-0737 is not necessary prior to low power operation because only small quantities of radionuclide inventory will exist in the reactor coolant system and therefore will not affect the health and safety of the public.
Prior to exceeding 5%
power operation the applicant must demonstrate the capability-to promptly obtain reactor coolant samples 'in the event of an accident in which there is core damage consistent with the license condition stated below:
6-Provide a procedure for relating radionuclide gaseous and ionic species to estimate core damage.
10 - Demonstrate that the reactor coolant system and suppression chamber sam-ple locations are representative of core conditions.
1 In addition to the above licensing conditions the staff is conducting a generic review of accuracy and sensitivity for analytical procedures and on-line instru-mentation to be used for post-accident analysis. We will require that the appli-cants submit data supporting the applicability of each selected analytical chemistry procedure or on-line instrument along with documentation demonstrat-ing compliance with the licensing conditions four months prior to exceeding 5%
power operation, but review and approval of these procedures will not be a con-dition for full power operation.
In the event our generic review determines a specific procedure is unacceptable, we will require the applicants to make modi-fications as determined by our generic review.
The license will be cor.ditioned for those items not yet fully resolved.
II.E.4.2 Containment Isolation Dependability We have reviewed the applicants' containment isolation system for compliance with the requirements of Item II.E.4.2 in NUREG-0737.
The following discus-i sion summarizes our evaluation.
a.
The containment isolation system design complies with the provisions of Standard Review Plan (SRP) Section 6.2.4 in that there is diversity in the parameters sensed for the initiation of containment isolation.
b.
The applicants have classified the systems penetrating containment into three groups:
- 1) essential; 2) beneficial; and 3) non essential. -Although the staff only considers classifying the systems into essential' and non-essential categories, the use of the category " beneficial" is acceptable as long as the isolation criteria are met.
All non-essential systems will be automatically isolated upon receipt of the containment signal.
The applicants have identified three systems as beneficial:
the Post Accident Sampling (PAS); the Control Rod Drive Hydraulic System (CRDHS);
and the Component Cooling Water System (CCWS).
The PAS will meet the Grand Gulf SSER #1 22-4
isolation criteria by having a locked closed isolation valve that is under administrative control.
The CRDHS is capable of supplying 200 gallons per minute of cooling water to the reactor vessel during an accident and, therefore, is an important source of high pressure vessel makeup water.
The system design is accep-table because it is not desirable to have design provisions for automatic isolation which would prevent the system from operating post LOCA.
The CCWS supplies cooling water to the recirculation pump seals. 'Without adequate cooling, the pump seals could be damaged.
Hence, the, system design is acceptable because it is desirable to have the recirculation pump available during an accident.
c.
The design of the control system for containment isolation valves is such that resetting the. isolation signal will not result in the automatic reopen-ing of the containment isolation valves, except for the isolation valves in the steam supply line to the RCIC turbine system. We will require the applicants to justify the control system design for the valves in the steam supply line to the RCIC turbine system or to modify the system to meet the requirements of this position.
d.
The containment setpoint pressure that initiates containment isolation for nonessential penetrations is approximately 2 psig which is the mini-mum value compatible with normal operating conditions.
e.
Operability bases for the containment purge valves will be addressed in Section 3.11.
f.
The containment purge and vent isolation valves are designed to close on a high radiation signal.
Based on our review, we conclude that the applicants have complied with the requirements for containment isolation dependability detailed in Section II.E.4.2 of NUREG-0737, except for the reset feature of the containment isolation signal for the valves on the steam supply line to RCIC turbine and purge valve opera-bility which will be addressed in Section 3.11.
We will report our resolution of this matter in a supplemental report.
II.F.1.5 Containment Water Level Monitor Discussion and Conclusions In our Safety Evaluation Report (SER) we stated that the lower limit of the water level monitor for the suppression pool was not based on the elevation of l
the suction line inlet for the cooling system as. required.
To satisfy the lower I
limit requirement, Mississippi Power and Light will modify the instrumentation I
to span down to the centerline of the emergency core cooling suction line inlet l
elevation. The applicants have committed to perform this modification prior to l
the fuel load date.
Hence, the applicants' proposed modification and commit-ment to perform these modifications prior to the fuel load date constitutes an acceptable fulfillment of the requirements of the NUREG-0737 II.F.1.5.
Grand Gulf SSER #1 22-5 l
II.K.1 IE Bulletins on Measures to Mitigate Small Break LOCAs and Loss-of-Feedwater Accidents II.K.l.23 Reactor Vessel Level Instrumentation Discussion and Conclusions In a letter AECM-81/311 dated August 21, 1981 from L. F. Dale (Mississippi Power & Light) to H. R. Denton (NRC), the applicants summarized the reactor vessel level instrumentation used at Grand Gulf.
The instruments that sense the water level are differential pressure devices calibrated for accuracy at a specific vessel pressure and liquid temperature condition.
This instrumenta-tion is extensively detailed in the General Electric Company Report NEDO-24708,
" Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors," and has been reviewed by us and evaluated in NUREG-0626, " Generic Evaluation of Feedwater Transients and Small-Break Loss-of-Coolant Accidents in GE-Designed Operating Plants and Near-Term Operating License Applications."
We find this acceptable for this item.
II.K.3 Final Recommendations of Bulletins and Orders Task Force II.K.3.25 Loss of Power to Pump Seal Coolers Discussion and Conclusions During our review of this item, we required that the seal cooling system remain operable if it is needed to protect against excessive leakage after a loss-of-offsite power event, or that additional data be provided showing that the pump seals will not degrade to a condition of excessive leakage following a 2-hour loss of offsite power.
The applicants have verified that no automatic isolation of the emergency power component cooling water system will result upon receipt of an ECCS initiation signal even though the system is classified as non-essential.
Pump seal cooling will, therefore, continue upon loss-of-offsite power. This is acceptable to us.
Containment isolation acceptability of this design is discussed under Item II.E.4.2.
II.K.3.27 Common Reference Level Discussion and Conclusions In a letter of September 10, 1981, from L. F. Dale to H. R. Denton, the Mis-sissippi Power and Light Company (MP&L) submitted proposed changes to the Grand Gulf Nuclear Station FSAR.
These changes are as follows:
In order to satisfy the requirements of a common reactor vessel level reference point for all reactor vessel level loops, Missis-sippi Power & Light will make the necessary modifications to reference the Fuel Zone instrument from the bottom of the reactor vessel steam dryer skirt (referenced to instrument zero, 533 vessel inches).
Grand Gulf SSER #1 22-6
New scales will be obtained for the Fuel Zone instrument indicator and recorder, and both will reflect a range from -117 to -317 inches.
. Top of Active Fuel (TOAF) will be marked on the scale at -167 inches.
These changes will be implemented prior to fuel loadino.
The proposed modifications and implementation schedule are acceptable and meet the requirement of NUREG-0737 II.K.3.27.
I Grand Gulf SSER #1 22-7
CHRONOLOGY APPENDIX A GRAND GULF NUCLEAR STATION, UNITS 1 AND 2 August 21, 1981 Letter from applicant
- providing additional information in instrumentation and control area.
August 21, 1981 Letter from applicant transmitting proposed FSAR changes -
Fire protection, materials engineering.
August 21, 1981 Letter from applicant transmitting proposed FSAR changes -
Auxiliary systems.
August 24, 1981 Letter from applicant transmitting proposed FSAR changes.
August 25, 1981 Letter to applicant concerning safeguards contingency plan.
August 25, 1981 Letter from applicant concerning emergency plan.
August 25, 1981 Letter from applicant transmitting proposed FSAR changes -
Design Basis Flood, R.G. 1.85, Roof water.
August 26, 1981 Letter from applicant transmitting proposed FSAR changes -
Section 13.1.
August 26, 1981 Letter from applicant transmitting proposed FSAR changes.
August 26, 1981 Letter from applicant concerning containment LOCA loads.
August 27, 1981 Letter from applicant transmitting proposed FSAR changes -
Section 13.5.2, Fire Protection.
August 27, 1981 Letter from applicant transmitting proposed FSAR changes -
Reactor Systems, Materials, Mechanical.
August 27, 1981 Letter from applicant concerning training for mitigation of core damage.
August 28, 1981 Letter from applicant transmitting proposed FSAR changes -
Structural Engineering.
- As used in this chronology, " applicant" refers to Mississippi Power and Light.
Grand Gulf SSER #1 A-1 wr-,
n
- ~ ~
,. -. -- ---,=,, -
.,n
1 August 28, 1981 Letter from applicant filing Amendment 50 affidavit.
August 28, 1981 Letter from applicant transmitting proposed FSAR changes.
August 28, 1981 Letter from applicant concerning emergency procedures and meteorological systems.
August 31, 1981 Letter from applicant concerning hydrogen control.
August 31, 1981 Letter from applicant concerning emergency plan.
September 1, 1981 Letter from applicant concerning Class IE electrical equipment.
September 1, 1981 Letter from applicant concerning preservice inspection.
September 1, 1981 Letter from applicant concerning plant shielding.
September 1, 1981 Letter from applicant transmitting proposed FSAR changes.
September 2, 1981 Letter from applicant transmitting proposed FSAR changes.
September 3, 1981 Letter from applicant transmitting proposed FSAR changes - UHS, soil structure interaction.
September 3, 1981 Letter from applicant concering perched aquifer level changes, PMF.
September 4, 1981 Letter from applicant concerning noble gas effluent monitor.
September 4, 1981 Letter from applicant transmitting proposed FSAR changes.
September 10, 1981 Letter from applicant transmitting proposed FSAR changes.
September 11, 1981 Letter from applicant concerning hydrogen action items.
September 14, 1981 Letter from applicant concerning compliance with 10 CFR Parts 20, 50, and 100.
September 22, 1981 Issuance of Grand Gulf Safety Evaluation Report.
September 23, 1981 Letter from applicant concerning physical security plan.
September 23, 1981 Letter from applicant providing additional information re damping value for cable tray design.
September 28, 1981 Letter from applicant concerning GGNS Class A Model.
Grand Gulf SSER #1 A-2
September 30, 1981 Letter from applicant transmitting proposed FSAR changes - containment water level.
September 30, 1981 Summary of meeting on pool dynamic loads.
October 1, 1981 Letter from applicant concerning fuel load /startup testing.
October 9, 1981 Letter from applicant concerning schedule for submittal of SER outstanding items.
October 9, 1981 Letter from applicant responding to SER items -
New Madrid Fault, noise levels.
October 9, 1981 Issuance of Grand Gulf Final Environmental Statement.
October 9, 1981 Letter from applicant transmitting responses to questions on fluid dynamics.
October 9, 1981 Letter from applicant concerning emergency preparedness evaluations report.
October 9, 1981 Letter from applicant responding to SQRT requests.
October 12, 1981 Letter from applicant responding to audit of licensee qualification branch.
October 12, 1981 Letter from applicant concerning GGNS environmental Protection Plan.
October 13, 1981 Letter from applicant responding to SER Item -
Culvert 1.
October 13, 1981 Letter from applicant transmitting corrections to SER.
October 19, 1981 Letter from applicant concerning results of ODYN transient event analyses.
October 22, 1981 Letter to applicnat transmitting ACRS letter.
October 23, 1981 Letter from applicant transmitting proposed FSAR changes - ultimate containment capacity, fire protection, RHR.
October 23, 1981 Letter from applicant concerning plant shielding.
October 28, 1981 Letter to applicant concerning hydrogen control.
October 28, 1931 Letter from applicant concerning NUREG-0654.
October 30, 1981 Letter to applicant concerning preservice inspection I
and testing of snubbers.
Grand-Gulf SSER #1 A-3
I A
0ctober 30, 1981 Letter from applicant transmitting proposed FSAR changes - UHS.
45 t
s 4
4
'N m Grand Gulf SSER #1 A-4 m
m-_ _ _ _ _ _ _ _ _. __ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _. _ _. _ _. _ _ _ _ _ _. _ _. - _ _ _ _ _ _ _ _. _ _ _ _. _ _ _ _ _ _ _. _.
APPENDIX B ACRS REPORT
[on mag'o UNITED STATES P
3,
{g NUCLEAR REGULATORY COMMISSION M.g
"/.t ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
[
WASWNGTON, D. C. 20555 October 20, 1981 Honorable Nunzio J. Palladino Chairman U.S. Nuclear Regulatory Commission Washington, DC 20555 SU3 JECT:
INTERIM REPORT ON GRAND GULF NilCLEAR STATION UNIT 1
Dear Dr. Palladino:
During its 258th meeting, October 15-17, 1981, the Advisory Committee on Reactor Safeguards reviewed the application of the Mississippi Power and Light Company (MP&L), Middle South Energy. Inc., and the South Mississippi Electric Power Association for a license to operate the Grand Gulf Nuclear Station Units 1 and 2.
The units are to be operated by the Mississippi Power and Light Company. A Subcommittee meeting was held in Jackson, Mississippi on September 17-18, 1981 to consider this project. A tour of the facility was made by members of the Subcommittee on September 17, 1981. During its review, the Committee had the benefit of discussions with representatives of the Applicant, the NRC Staff, and members of the public.
The Committee also had the benefit of the documents listed. The Committee commented on the construction pemit application for this station in its report dated May 15, 1974.
The Grand Gulf Station is located in Claiborne County, Mississippi on the east side of the Mississippi River about 25 miles south of Vicksburg, the nearest city having a population in excess of 25,000 persons.
Each Grand Gulf unit is equipped with a General Electric BWR-6 nuclear steam supply system with a rated power level of 3833 MWt and a Mark III pressure suppression containment system with a design pressure of 15 psig.
Construction of Unit 1 is over 90% complete while Unit 2 is about 20% com-plete and construction of it has been temporarily suspended.
Because of the extended schedule for Unit 2, the Committee does not believe it appropriate to report on operation of Unit 2 at this time.
The Committee review included the management organization, capability, and operator training of MP8L. This is the first nuclear power plant to be operated by this utility. While the plant staff has a reasonable amount of Grand Gulf SSER #1 B-1
Honorable Nunzio J. Palladino October 20, 1981 nuclear background, the ACRS agrees with the NRC Staff on the need for ad-ditional personnel with BWR experience, at least during the first year or two of operation. MP&L also needs to fill certain senior technical person-nel positions in its management organization. The Committee recommends that the MP&L Nuclear Safety Review Board include two or more experienced voting members from outside MP&L having appropriate backgrounds.
During this meeting, the NRC Staff identified a large number of license conditions and confirmatory matters, and several outstanding issues which remain to be resol ed.
Except for the two issues identified below, the ACRS is satisfied with progress on the other topics and believes that they should be resolved in a manner satisfactory to the NRC Staff.
We have not completed our review of the following outstanding issues identi-fied in the NRC Staff Safety Evaluation Report:
dynamic loads on structures above the Mark III suppression pool due to froth impact hydrogen control The ACRS will complete its review of the full power operating license when the Staff and the Applicant have made sufficient additional progress in resolving these items.
In the interim, the ACRS believes that if due consideration is given to the recommendations above, and subject to satis-factory completion of construction, staffing, and preoperational testing, it would be acceptable for Grand Gulf Nuclear Station Unit 1 to be operated at power levels up to 5% of full power.
Sincerely, J. Carson Mark Chairman
References:
1.
Mississippi Power and Light Company, " Final Safety Analysis Report, Grand Gulf Nuclear Station Units 1 and 2," Volumes 1-21 and Amend-ments 25-50 2.
U. S. Geological Survey Professional Paper by T. G. Hildenbrand, M. F.
Kane, and J. D. Hendricks, " Magnetic Basement in the Upper Mississippi Embayment Region - A Preliminary Report," received August,1981 3.
Report by S. W. Hatch, Sandia National Laboratories and P. Cybulskis and R. O. Wooton, Battelle Columbus Laboratories for Office of Nuclear Regulatory Research, NRC, "The Reactor Safety Study Methodology Ap-plications Program Results for the Grand Gulf fl BWR Power Plant,"
NUREG/CR-1659, Vol. 4, SAND 80-1897/4, Draft Received 2/6/81 Grand Gulf SSER #1 B-2
l l
Honorable Nunzio J. Palladino October 20, 1981 1
4.
Letter from M. D. Houston, Project Manager, Division of Licensing, NRR, to H. Alderman, ACRS,
Subject:
Staff Responses to Questions asked by ACRS at the Grand Gulf Subcommittee Meeting, September 17-18, 1981, dated October 14, 1981 5.
Letters from L. F. Dale, Mississippi Power and Light Company to USNRC, dated August 27, 1981, August 27,1981, August 26, 1981, August 24, 1981, August 21, 1981, August 21, 1981, August 21, 1981, August 19, 1981, August 18, 1981, August 18, 1981 6.
Letter from C. Stewart, Jacksonians United for Livable Energy Policies (JULEP) to R. F. Fraley, ACRS, dated October 8,1981 7.
Letter from K. Lawrence, JULEP, to ACRS Grand Gulf Subcommittee dated September 18, 1981 8.
Statement by K. Lawrence, JULEP, to ACRS Grand Gulf Subcommittee dated September 17, 1981 9.
Letter from C. Dana, et al., member of public, to ACRS Subcommittee on Reactor Safety dated September 16, 1981
- 10. Anonymous letter to H. Alderman, ACRS Staff, regarding quality assurance concern, postmarked September 18, 1981 e
Grand Gulf SSER #1 B-3
APPENDIX C NUCLEAR REGULATORY COMMISSION (NRC)
UNRESOLVED SAFETY ISSUES A-39 Safety Relief Valve Hydrodynamic Loads Safety relief valves inplant tests with cross quencher devices were performed during August 1981 at the Kuosheng Power Plant in Taiwan, the first operating BWR 6/ Mark III Plant in the world. The NRC staff participated in this tech-nical activity and has made a preliminary evaluation of the test results.
i In general, the SRV loads were within the expected values. The measured building response (acceleration, in particular) was relatively low in com-parison with the expected values. The forcing function (pressure measurements),
however, is much closer to the expected values.
Some pressure measurements even exceeded the expected values. These exceedances may only represent loca-i lized loads and will not contribute significantly in terms of global loads on the entire containment structures.
Based on these observations, we conclude that the KNPS inplant SRV tests show the design SRV loads are adequate for the Kuosheng Nuclear Power Station.
Applicability of the data to other Mark III-plants will require a more detailed evaluation of the data and the dynamic analysis of the Kuosheng building re-sponse to SRV loads, e.g., analysis of amplified response spectra.
The KNPS inplant SRV tests definitely provide a relevant data base for licens-ing Mark III plants with GE designed cross quenchers with respect to the issue of suppression pool temperature limits.
In view of geometrical similarity of the Mark III containments, the data base should be applicable for all Mark III containments.
Grand Gulf SSER #1 C-1
APPENDIX D ERRATA TO THE SAFETY EVALUATION REPORT Page 1-1 Line 29 Change with own exception to with the exception Page 1-3 Line 28 Change radial walls to radial wells Page 1-4 Line 1 Add low pressure before before core spray system Page 1-5 Line 19 Change of fuel per to of fuel rods per Page 1-6 Line 13 For core spray loops, change 1 to 2 Page 1-12 All Missing section 1.12 inserted as Appendix F in supplement Page 2-1 Line 1
-Change 1200 to 2300 Page 2-1 Line 21 Change 312,930 to 684,360 Page 2-19 Line 36 Change Lakes in to Lakes and Page 2-23 Line 18
' Change 26,4000 to 26,400 Page 2-24 Line 3 Change clay, and and to clay, sand, and Page 2-24 Line 30 Change Ksg to Ksf Page 2-25 Line 4 Change limited movements to limited horizontal movements Page 2-28 Line 8 Change method to method
- Page 2-30 Line 31 Change station to stratum Page 2-34 Line 36 Change Muttli to Nuttli Page 3-12 Line 21 change oa caissons---Catahoula to compacted backfill Page 3-26 Line 44 Change all to some Page 4-6 Line 39 change not to now Page 4-13 Line 1 Delete sentence - Design bases........(G0C 12)
Page 4-13 Line 35 Change satisfying to satisfy Grand Gulf SSER #1 0-1
~
Page 4-14 Line 2 Change in Evaluation to in Topical Report NEDO-20430, " Process Computer Performance Evaluation,"
Page 4-14 Line 44 change F to i Page 4-16 Line 11 change auxially to axially Page 4-18 Line 22 change burnup to buildup Page 4-18 Line 28 change k/k to Ak/k Page 5-10 Line 21 change temperature sensors to temperature or pressure sensors Page 5-16 Line 9 change reactor building to auxiliary building Page 5-18 Line 14 Change 150 to 200 Page 5-18 Line 15 change 300 to 500 Page 5-19 Line 27 change drywell and suppression pool to containment Page 5-19 Line 33 change reactor building to auxiliary building Page 5-20 Line 9 change two to four Page 6-15 Line 7 change drywell, to drywell on stainless steel Class 1 pipe greater than 2 inches diameter, Page 6-25 Line 28 change 5.4.7 to 5.4.2 Page 6-30 Line 4 change equipment room to air plenum Page 6-31 Line 34 change reactor to cuxiliary Page 6-31 Line 36 change 120 to 101 Page 6-34 Line 37 change at to and Page 7-12 Line 19 change and status to and inoperable status Page 7-12 Line 27 change and status to and inoperable status Page 7-13 Line 24 change reactor level to reactor water level Page 7-14 Line 6 Delete Between 20.........or patterns.
Page 7-14 Line 9 change Above 70 to Between 20 and 70 percent and above 70 Page 8-7 Line 36 change two to four Grand Gulf SSER #1 0-2
Page 8-11 Line 27 Change AC to DC Page 8-12 Line 4 Change unvoltage to undervoltage Page 8-18 Line 5 change Each bus to Each Div. I and Div. 2 bus Page 9-2 Line 3 change auxiliary to fuel handling Page 9-21 Line 35 After rooms add, fuel pool cooling and cleanup pump room Page 9-29 Line 16 Change control fire to control room fire Page 9-29 Line 22 change 1 -hour fire to bullet resistant certified Page 9-30 Line 4 Change VI to 9.5.6 Page 9-30 Line 21 Should read:
9.5.4.3 Containment and Auxiliary Building Page 9-31 Line 7 After rooms add operators manually start supply fans in two unaffected diesel generator rooms Page 9-41 Line 18 Change 9.5.7 to 9.6.6 Page 9-48 Line 24 Change 9.5.4.1 to 9.6.3.1 Page 10-1 Line 28 Change 1350 to 1525 Page 11-1 Line 5 Delete laundry Page 12-5 Line 18 Change equipment to equipment hatch Page 12-7 Line 18 change 40 to 41 Page 12-7 Line 19 change 8 to 9 Page 13-10 Line 32 change 13.1-2 to 13.1-4 Page 13-35 Line 22 change ANS 18.7 to ASNI 18.7 Page 15-17 Line 40 After full power add and to 48 inches when the power is at 70% of full power Page 15-18 Line 21 change Looped to Loaded Grand Gulf SSER #1 0-3 i
Page 15-19 Line 31 change latest to least Page 17-4 Line 34 after indicate add adverse add footnote:
Exceptions to specific Regulatory Page 17-6 Table 17.1 Guide.s are discussed in MPL-TOP-1A, Appendix A.
Page 22-9 Line 19 change 10 to 100 Page 22-14 Line 18 change gaseous, and liquid radwaste to drain Page 22-19 Line 34 change 107 to 10 7 Page 22-21 Line 35 change 4 to 5 Page 22-22 Line 30 change 1.197 to 1.97 Page 22-26 Line 40 change present to preset Page 22-27 Line 24 change 10 to 30 Page 22-31 Line 8 change Susquehanna to Grand Gulf Page A-13 Line 24 Add: July 1, 1981 - Letter from applicant submitting response to NOREG-0588 Grand Gulf SSER #1 D-4
i b
i 1 -
APPENDIX E I
CONTROL ROOM REVIEW-DISCUSSION 1
The SER~ dated August 14, 1981 identified six (6) systems and. items which were
-not available for review at the time of the staff site visit, June 8 through June 12,'1981.
In addition,'the applicant's proposals on three (3) other
~
items were not acceptable to the staff.
4 A meeting was-held in Bethesda on August 25, 1981 to discuss the review and reporting requirements for.the six items not reviewed and the staff positions on the three unacceptable items.
The systems and items not available for our review were (1) the control room environment and the environment in the area i
of the Remote Shutdown panels, (2) the NSSS monitoring equipment, (3) the j
communications equipment, (4).the storage, adequacy, and availability of i-
' emergency equipment for use by operating personnel, (5) the operational storage and availability of procedures and reference material in the control j
room, and (6) the installation of the labels and location aids on the shared-panels in the main control area, the back panels, and the Remote Shutdown.
i panels. As stated in the SER we require that the applicants perform an evaluation of these items after installation and submit their findings, proposed corrective actions, and schedule for implementing the actions. We must receive this information for our review and approval 60 days prior to_
issuance of the. operating license.
To date we_have received no report on any of the above items.
I The staff position on two of the three unresolved items remains unchanged from the SER-dated August 14,11981. These items are, (1) small diameter, flush-j mounted pushbuttons, which are difficult to depress, and (2) lack of reflash capability for annunciator windows.
l Detailed review of the operating procedures involved in the third item has led to its resolution. The item, MP&L response, and staff position, are discussed i
below.
l j
Item: While increasing or decreasing-power using the RECIR MASTER CONTROLLER on section 3D of Panel 680, the operator must simultaneously watch the IRM and
(
APRM recorders on section 58 and section 'iB.
MP&L Response:
Changes.in power'using the RECIRC MASTER CONTROLLER will normally be slow, controlled, " bump and-wait" type evolutions. An operator l
can adequately monitor the IRM and APRM recorders while performing this_
controlled evolution.
The distance from'the RECIRC MASTER CONTROLLER to the recorders on section 5B is approximately 53 inches. Additionally, video i-guides will be developed when the computer interface is complete, which will include APRM indications. These video guides, when displayed on section 4B of l
Panel 680,.will provide additional power monitors in clost proximity to the
?
recirculation flow' control station.
The computer interface is expected to be operational prior to commercial operation.
I I
l Grand Gulf SSER #1 E-1 f
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Staff Position: The IRM and APRM indications must be available in close proximity to the operator while controlling power increase or decrease using the RECIRC MASTER CONTROLLER.
The staff will accept the use of video guides on the CRT, which is in close proximity to the control station, to accomplish this function.
It is the staff position that these CRT indications must be provided within three (3) months after achieving 30 percent power operation, but no later than October 1, 1982.
CONCLUSION The applicants' response to one of the nine open items, documented in the SER dated August 14, 1981 and discussed in a meeting August 25, 1981, is acceptable to the staff.
The applicants shall evaluate and report on the six (6) items not available for our review during the staff site visit.
The report on these items identified in the discussion section of this report shall be submitted for our review and approval 60 days before issuance of the operating license.
The two (2) unresolved items will be required to be corrected prior to exceeding 5 percent power operation because the staff believes their resolu-tion will contribute to safe operation of the plant. We must receive the applicants' acceptable proposals for correcting these items 60 days prior to issuance of an operating license.
Based on our review of Mississippi Power and Light Company submittals, our control room review, and other clarifying information, and pending satisfac~
tory resolution of the eight (8) open items identified in the SER for which corrective actions shall all be implemented prior to exceeding 5 percent power operation, the potential for operator error leading to serious consequences as a result of human factors considerations in the control room will be sufficiently low to permit safe power operation of the Grand Gulf Nuclear Station Unit 1.
Our evaluation of the Grand Gulf Nuclear Station Unit 1 control room will be reported on in a later supplement to the safety evaluation report prior to issuance of an Operating License.
Grand Gulf SSER #1 E-2
APPENDIX F NRC REVIEW TEAM-Mr. M. Dean Houston is the NRC Project Manager for this project.
Mr. Houston may be contacted at the U.S. Nuclear-Regulatory Commission on 301/492-8430.
The principal NRC Staff reviewers for this project are:
D. Terao, MEB, Mechanical Engineer A. Cappuci, MEB, Mechanical Engineer E. Hemminger, MEB, Mechanical Engineer R. Lipinski, SEB Sr. Structural Engineer
- 0. Rothberg, SEB,.aructural Engineer T. Cardone, GSB, Geologist J. Kimball, GSB, Geophysicist G. Staley, HGEB, Hydraulic Engineer J. Philip, HGEB, Geotechnical Engineer M. Hum, MTEB, Sr. Materials Engineer B. Elliot, MTEB, Materials Engineer J. Halapatz, MTEB, Materials Engineer G. Harrison, CEB, Sr. Protection Engineer F. Witt, CEB, Sr. Chemical Engineer J. Kennedy, EQB, Materials Engineer T. Y. Chang, EQB, Sr. Mechanical Engineer J. Conway, QAB, Sr. Quality Assurance Engineer D. Smith, MTEB, Sr. Materials Engineer D. Huang, MTEB, Materials Engineer E. Markee, AEB, Section Leader, Meteorology W. Brooks, CPB, Sr. Reactor Physicist A. Brauner, SAB, Site Analyst F. Litton, MTEB, Sr. Materials Engineer J. Petersen, SP, Sr. Financialyst D. Kers, SGPL, Plant Protection Analyst J. Gleim, MTEB, Sr. Materials Engineer J. Wermiel, ASB, Sr. Auxiliary Systems Integration i
Grand Gulf SSER #1 F-1
T. Collins, RSB, Reactor Engineer J. Minns, RAB, Health Physicist M. Fields, CSB, Containment Systems Engineer C. Tinkler, CSB, Sr. Containment Systems Engineer T. Greene, CSB, Systems Engineer J. E. Knight, ICSB, Principal Electrical Engineer S. Rhow, PSB, Electrical Engineer R. Giardina, PSB, Reactor Systems Engineer - Mechanical J. Lee, ETSB, Sr. Nuclear Engineer M. Tokar, CPB, Reactor Engineer S. Sun, CPB, Nuclear Engineer A. Ramey-Smith, HFEB, Engineering Psychologist D. Eckenrode, HFEB, Sr. Human Factors Analyst V. Deliso, OLB, Nuclear Engineer R. Benedict, LQB, Sr. Nuclear Engineer R. Clifford, PTRB, Operational Safety Engineer D. Fischer, PTRB, Section Leader, Tests B. Long, PTRB, Sr. Operational Safety Engineer D. Perrotti, EPLB, Team Leader C. Anderson, GIB, Sr. Systems Engineer R. Bottimore, LGB, Sr. Reactor Engineer R. Kirkwood, MEB, Principal Mechanical Engineer B. Siegel, RSB, Reactor Engineer J. Read, AEB, Nuclear Engineer - Chemical J. Spraul, QAB, Sr. Quality Assurance Engineer The following consultants to the staff participated in this review:
Battelle Pacific Northwest Laboratory Bio Technology, Inc.
EG&G, Idaho Lawrence Livermore National Laboratory Sandia National Laboratory United States Geological Survey.
Grand Gulf SSER #1 F-2
NRC Po7.u 335 U.S. NUCLEA2 CEIULATORY COMMISSION
- 1. LEPORT NUMBER (Assipedby DOC /
NUREG-0831 BIBLIOGRAPHIC DATA SHEET Supplement No. 1 1 TITLE AND SUBTITLE (Add Votume Na,if appropr,a#l
- 2. (Leave D/mik).
Safety Evaluation Report related to the operation of Grand Gulf Nuclear Station, RECIPIENT'S ACCESSION NO.
Units 1 and 2
- 7. AUTHOR (S)
S. DATE REPORT COMPLETED MON TH l YEAR December 1981
- 9. PET. FORMING ORGANIZATION NAME AND MAILING ADDRESS (lactude 2,p Code /
DATE REPORT ISSUED MONTH l YEAR U. S. Nuclear Regulatory Commission December 1981 Office of Nuclear Reactor Regulation
- s. (teave nianal Washington, D. C. 20555
- 8. (Leave blank)
- 12. SPONSORING ORGANIZATION N AME AND MAILING ADDRESS //nclude 2,0 Codel p
Same as 9 above n C NTRACT NO.
Safety Evaluation Report
- 15. SUPPLEMENTARY NOTES
- 14. (Leave clanAl Pertains to Docket Nos. 50-416 and 50-417
- 16. ABSTR ACT Q00 words or less)
Supplement No. 1 to the Safety Evaluation Report on the application filed by Mississippi Power & Light Company, et al, for licenses to operate the Grand Gulf Nuclear Station, Units 1 and 2, located on the east bank of the Mississippi River near Port Gibson, in Claiborne County, Mississippi, has been prepared by the Office of Nuclear Reactor Regulation of the U.
S. Nuclear Regulatory Commission.
This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report.
- 17. KEY WORDS AND DOCUMENT ANALYSIS 17a DESCRIPTORS 1
17b IDENTIFIE RS,OPEN ENDE D TERVS 18 AV AILABILITY STATEMENT
- 19. p.E CUQlTY CLe qS.(Thy reporr) 21 NO OF PAGES UnCiaSS11100 Unlimited
- o SECURITY CLASS (Thispavel 22 PRICE Uncinngifigg s
N RC F ORV 335 (7 ? ?)
2 UNITED STATES F
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C NUCLEAR REGULATORY COMMtssiON 3
W ASHIN GTON. O. C. 20555 Post AGE AND FEES P AeD I
U.S. N UCLE A R R E GU LATO R Y OF FICI AL BussNESS couwassiON PENALTY FOR PRIVATE USE $300 L
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