ML20138A799
| ML20138A799 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/30/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0824, NUREG-0824-S01, NUREG-824, NUREG-824-S1, NUDOCS 8512120114 | |
| Download: ML20138A799 (58) | |
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NUREG-0824 Supplement No.1 Integrated Plant Safety Assessment Systematic Evaluation Program Millstone Nuclear Power Station, Unit 1
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Northeast Nuclear Energy Company i
Docket No. 50-245 1
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U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation November 1985 t
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NOTICE i'
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NUREG-0824 Supplemsnt No.1 Integrated Plant Safety Assessment Systematic Evaluation Program Millstone Nuclear Power Station, Unit 1 Northeast Nuclear Energy Company Docket No. 50-245 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation November 1985
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ABSTRACT The U.S. Nuclear Regulatory Commission (NRC) has prepared Supplement 1 to the final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0824), under 7
the scope of the Systematic Evaluation Program (SEP), for Northeast Nuclear Energy Company's Millstone Nuclear Power Station, Unit 1 located in Waterford, Connecticut.
The SEP was initiated _by the NRC to review the design of older
' operating nuclear power plants to reconfirm and document their safety. This report documents the review completed under the SEP for_those issues that required refined engineering evaluations or the continuation of ongoing evaluations subsequent to issuing the Final IPSAR for the Millstone 1 plant.
The review has provided for (1) an assessment of the significance of differences between current technical positions on selected safety issues and those that existed when Millstone 1 was licensed, (2) a basis for deciding how these differences should be resolved in an integrated plant review, and (3) a documented evaluation'of plant safety.
The final IPSAR and its supplement will form part of the bases.for considering the conversion of the existing provisional operating license to a full-term operating license.
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CONTENTS Page ABSTRACT..........................................................
iii ACRONYMS AND INITIALISMS..........................................
viii 1
INTRODUCTION.................................................
1-1
-2 TOPICS THAT REQUIRED REFINED ENGINEERING ANALYSIS OR CONTINUATION OF ONG0ING EVALUATION...........................
2-1 2.1 Topics II-3.B Flooding Potential and Protection Requirements; II-3.B.1, Capability of Operating Plants to Cape with Design-Basis Flooding Conditions; and III-3.A, Effects of High Water Level on Structures.....................................
2-1 2.1.1 Flooding Elevation............................
2-1 2.1.2 Roofs.........................................
2-2 2.2 Topic II-4.F, Settlement of Foundations and Buried Equipment........................................
2-2
- 2. 2.1 -
Turbine Building..............................
2-2 2.2.2 Gas Turbine Generator Building................
2-2 2.2.3 Buried Pipelines..............................
2-3 2.3 Topic III-1, Classification of Structures, Components, and Systems.................................
2-3 4
2.4 Topics III-2 Wind and Tornado Loadings and III-4.A, Tornado Missiles........................................
2-3 2.5 Topic III-3.A, Effects of High Water Level on Structures..........................................
2-4 2.6 Topic III-4.B, Turbine Missiles........................
2-5 2.7 Topic III-5.A, Effects of Pipe Break on Structures, Systems, and Components Inside Containment.............
2-6 2.7.1 Cascading Pipe Breaks........................
2-6 2.7.2 Jet Impingement..............................
2-7 2.7.3 Pipe Whip....................................
2-8 2.8 Topic III-5.B, Pipe Break Outside Containment..........
2-8 2.9 Topic III-6, Seismic Design Considerations.............
2-9 2.9.1 Pile Foundations.............................
2-9
~2.9.2 Motor-0perated Valves........................
2-9 2.9.3 Transformers and Control Room Panels.........
2-10 2.9.4 Quali fication of Cable Trays..................
2-10 2.9.5 Reactor Vessel Internals.....................
2-10 v~
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2.10 Topic III-7.B Design Codes, Design Criteria and Load Combinations...........................................
2-10 2.11 Topic III-10. A Thermal Overload Protection for Motors of Motor-0perated Valves........................
2-10 2.12 Topic VI-4, Containment Isolation System...............
2-11 2.12.1 Remote Manual Valves.........................
2-11 2.12.2 La c k o f I n fo rma t i o n..........................
2-11 2.13 Topic VI-7.A.3, Emergency Core Cooling System Actuation System.......................................
2-12 2.14 Topic VI-7.C.1 Appendix K-Electrical Instrumentation and Control Re-reviews.................................
2-12 2.14.1 Automatic Bus Transfers......................
2-13 2.14.2 Manual Bus Transfers.........................
2-14 2.15 Topic VII-1.A Isolation of Reactor Protection System from Non-Safety Systems, Including Qualifications of Isolation Devices...................................
2-14 2.16 Topic VIII-2 Onsite Emergency Power Systems............
2-14 2.17 Topic VIII-3.B DC Power System Bus Voltage Monitoring and Annunciation............................
2-15 2.18 Topic IX-5, Ventilation Systems........................
2-15 2.18.1 Core Spray and LPCI Systems Ventilation Systems..........................
2-16 2.18.2 Reinitiation of Ventilation Followin Loss of Offsi te Power...............g 2-16 2.18.3 FWCI and Diesel Generator Area Coolers.......
2-17 2.18.4 Intake Structure Ventilation System..........
2-17 3.
IPSAR TOPICS RESOLVED BY CHANGES TO PLANT TECHNICAL SPECIFICATIONS 3.1 Topic V-5, Reactor Coolant Pressure Boundar Leakage Detection..........................y 3-1 3.2 Topic V-12. A, Wa ter Chemis try L imi ts................... -
3-1 3.3 Topic VIII-3.A, Station Battery Test Requirements......
3-2 3.4 Topic VIII-3.B DC Power System Bus Voltage Moni toring a nd Annuncia tion............................
3-2 3.5 Topic XV-16, Radiological Consequences of Failure of Small Lines Carrying Primary Coolant Ou ts i de Co n ta i nmen t....................................3-3 3.6 Topic XV-18, Radiological Consequences of a Main Steam Line Failure Outside Containment............
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IPSAR TOPIC RESOLUTIONS CONFIRMED BY THE NRC REGION I 0FFICE.....................................................
4-1 5.
REFERENCES..................................................
5-1 APPENDIX A -- REFERENCES TO STAFF SERS FOR EACH TOPIC EVALUATED IN THE SUPPLEMENT...................
A-1 APPENDIX B -- NRC STAFF CONTRIBUTORS AND CONSULTANTS........
B-1 LIST OF TABLES 2.1 Summa ry of IPSAR and Supplement Evaluations.................
2-18 3.1 Modifications to Millstone 1 Technical Specifications a s a Re s u l t o f S E P..........................................
3-4 4.1 Items for Confi rmation by NRC Region I 0ffice...............
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Acronyms and Initialisms ABT automatic bus transfer ac alternating current APRM average power range monitor BTP branch technical position BTU British thermal unit BWR boiling water reactor CFM cubic feet per minute CFR Code of Federal Regulations CS core spray CST condensate storage tank dc direct current ECCS emergency core cooling systems EGTG emergency gas turbine generator F
Fahrenheit ft feet FTOL full-term operating license FWCI feedwater coolant injection GDC general design criterion GE General Electric ICSB Instrumentation and Control Branch IE-Inspection and Enforcement IEEE Institute of Electrical and Electronics Engineers IPSAR Integrated Plant Safety Assessment Report IREP Interim Reliability Evaluation Program IRM intermediate range monitor ISAP Integrated Safety Assessment Program LOCA loss-of-coolant accident LPCI low-pressure coolant injection MOV motor-operated valve ms1 mean sea level NNECO Northeast Nuclear Energy Company NRC Nuclear Regulatory Commission PMH Probable Maximum Hurricane PHP Probable Maximum Precipitation POL provisional operating license PRA probabilistic risk assessment psi
-pounds per square inch RCPB reactor coolant pressure boundary RPS
-reactor protection system RG Regulatory Guide RWCU reactor water cleanup SEP Systematic Evaluation Program SER safety evaluation report SRP Standard Review Plan USI Unresolved Safety Issue viii
INTEGRATED PLANT SAFETY ASSESSMENT REPORT SUPPLEMENT N0. 1 SYSTEMATIC EVALUATION PROGRAM MILLSTONE NUCLEAR POWER STATION, UNIT 1 1 INTRODUCTION The Systematic Evaluation Program (SEP) was initiated by the U.S. Nuclear Regulatory Comission to review the designs of older operating nuclear power plants to reconfirm and document their safety. The review provides (1) an assessment of the significance of differences between current technical positions on safety issues and those that existed when a particular plant was licensed, (2) a basis for deciding how these differences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety.
The results of the SEP review of Millstone I were published in NUREG-0874, the Final Integrated Plant Safety Assessment Report (IPSAR), dated February 1983.
The review compared the as-built plant design with current review criteria in 137 different areas defined as " topics." During the review, 51 topics were deleted from consideration in the SEP because a review was being conducted under other programs (unresolved safety issue or Three Mile Island Action Plan tasks), the topic was not applicable to Millstone 1, or the items to be reviewed under that topic did not exist at the site.
Of the original 137 topics, 86 were, therefore, reviewed for Millstone 1; of these, 48 met current criteria or were acceptable on another defined basis.
From the review of the 38 remaining topics, certain aspects of plant design were found to differ from current criteria. The review of these 38 topics resulted in 87 individual issues which were considered in the integrated assessment of the plant. The integrated assessment consisted of evaluating the safety significance and other factors of the identified differences from current design to arrive at decisions about whether modification was necessary from an overall plant safety viewpoint. To arrive at these decisions, engineering judgment was used as well as the results of a limited probabilistic risk assessment study.
In general, the staff's positions in the integrated assessment fell into one or more of the following categories:
(1) equipment modification or addition, (2) procedure development or Technical Specification changes, (3) refined engineeringanalysisorcontinuationofongoingevaluation,and(4)no modification necessary. Table 4.1 of the IPSAR summarizes the staff's integrated assessment positions and documents the licensee's agreement with those positions.
For those positions classified as either Category (1) or (2), the IPSAR lists the scheduled completion dates agreed upon by the staff and the licensee.
Region I has verified the implementation of these positions or identified certain corrective actions required, as described in Section 4.0 of this report.
For those positions classified as Category (3), the licensee has provided the results of their evaluation or engineering analyses. The purpose of this supplement to the IPSAR is to provide the staff's evaluation of the Category
-(3) issues and to summarize the status of all actions to be implemented as a 1-1
result of the SEP review.
In those cases where analyses are continuing, the staff's evaluation identifies the analyses to be performed and the acceptance criteria that will be used to design the optimum plant modifications, if necessary.- Any pre-implementation staff reviews required for these ongoing analyses, following this supplement, will be summarized in individual safety evaluation reports as the analyses are completed.
The Millstone 1 facility is one of the seven SEP plants that has not received a full-term operating license (FTOL).
Therefore, a Safety Evaluation Report (SER) to support the conversion of the provisional operating license (POL) to an FTOL has been prepared (NL' REG-1143). The SER consists of the IPSAR, the IPSAR supplement, a consideration of major plant modifications that have been made, substantive regulations adopted since the POL was issued, the unresolved safety issues and Three Mile Island Action Plan issues.
In a related activity, the licensee proposed in letters dated June 13, 1983, September 14, 1983, December 28, 1983 and May 17, 1985 to conduct an expanded
' integrated assessment for Millstone 1 which would address the outstanding SEP issues (Category 3), all pending licensing actions, and significant licensee-sponsored plant improvements. This effort is referred to as the Integrated SafetyAssessmentProgram(ISAP).
On November 15, 1984, the Commission published a policy statement in the
- Federal Register (49 FR 45112) which describes the elements and objectives of ISAP, as a regulatory vehicle to develop plant-specific, integrated implemen-tation schedules for plant modifications. Millstone 1 has been selected as one of two plants which will participate in an ISAP pilot program. Consequently, the results of the licensee's additional analyses for several of the issues discussed in this report refer to ISAP, where alternative corrective actions will be considered and a prioritized implementation schedule for any plant modifications will be developed. The entire scope of the ISAP review for Millstone 1 is detailed in a July 31, 1985 letter to the licensee.
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2 TOPICS THAT REQUIRED REFINED ENGINEERING ANALYSIS OR CONTINUATION OF ONGOING EVALUATION Table 2.1 of this report presents a list of all issues that were evaluated in the IPSAR.
The licensee has submitted an evaluation for each of the items identified in the final IPSAR as requiring additional analysis. A summary of the staff's findings of these items is presented in Sections 2.1 through 2.18 below.
Each section references the staff's Safety Evaluation Report if applicable which provides more detail regarding the basis for the staff's conclusions.
References for correspondence pertaining to Safety Evaluation Reports (SERs) for each section appear in Appendix A of this report.
Factors considered in reaching a staff conclusion for each item include the perceived safety significance of the difference from current licensing criteria and a qualitative assessment of the financial and radiation exposure costs to make a modification.
The evaluation of these issues also considered any applicable risk perspectives, developed for the integrated assessment and described in the IPSAR, and related corrective actions proposed by the license as part of the integrated assessment or as a result of the subsequent evaluations.
2.1 Topics II-3.B, Flooding Potential and Protection Requirements; II-3.B.1, Capability of Operating Plants to Cope With Design Basis Flooding Conditions; III-3.A, Effects of High Water Level on Structures (NUREG-0824 Sections 4.1 and 4.5)
As implemented by Standard Review Plan (SRP) Section 2.4.5 (NUREG-0800) and Regulatory Guide (RG) 1.59, General Design Criterion (GDC) 2 in 10 CFR 50, requires that structures, systems and components important to safety be designed to withstand the effects of natural phenomena such as floods.
2.1.1 Flooding Elevation In iPSAR Section 4.1.1, the staff concluded that the Probable Maximum Hurricane (PMH) flood level, including wave effects, is 22.3 ft ms1 (18.11 ft msl still-water level plus wave action).
Safety-related structures at Millstone 1 are protected by concrete floodwalls to elevation 19.0 ft msl.
As a result of this higher water level, including rave effects, it was determined that there could be intermittent inleakage ano, hat the floodwalls have not been analyzed for these additional forces.
The licensee submitted information addressing these issues by letters dated February 2, 1984 and March 16, 1984.
The licensee relies on the isolation coadenser, which is flood protected, to attain safe shutdown for this event.
To supply make-up to the isolation condenser, the licensee will use the firewater pumps located in the fire pumphouse.
The licensee has further proposed to perform modifications to the flood doors of this structure to assure that it is adequately designed for this event.
By letter dated October 7,1985, De staff issued its safety evaluation on this issue.
The staff concludes that the licensee's proposal is acceptable; however the licensee should confirm that the masonry blockwalls of the fire pumphouse will resist the forces induced by the ficod water including wave effects.
This concern will be evaluated as part of ISAP Issue 1.19, Integrated Structural Analysis.
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2.1.2 Roofs Section 4.1.7 of the IPSAR concluded that some roofs with parapets may be overstressed as a result of local Probable Maximum Precipitation (PMP). The licensee analyzed the roofs of safety-related structures and proposed to install scuppers on the roofs of the turbine building, reactor building, warehouse and heating and ventilation area roofs in order to assure that the loads on these roefs remain below the design roof live loads. This course of action is described in the licensee's submittal dated February 2,1984.
After reviewing the licensee's submittal, the staff concludes that these proposed modifications will adequately limit roof ponding and, therefore, are acceptable. The licensee plans to install the scuppers during the Fall 1985 refueling outage. Confirmation of installation will be made during routine inspections.
2.2 Topic II-4.F, Settlement of Foundations and Buried Equipment (NUREG-0824 Section 4.2) 10 CFR 50 (GDC 2 and 44) and 10 CFR 100, Appendix A as implemented by Regulatory Guide 1.132 and SRP Section 2.5.4, require that foundations and buried equip-ment important to safety be adequately designed to perform their intended functions.
The IPSAR identified three issues for further evaluation concerning pile foundations and buried pipelines, which are described below. The licensee i
addressed these three issues in a March 16, 1984 submittal. Upon reviewing the backup calculations, the staff questioned, in discussions with the licensee, the analysis methodology used by the licensee to show adequacy of the piles, i
The staff also questioned the ability of recent soil borings to detect the possibility of peat beneath the buried pipelines given the depth and location of the borings. The licensee is preparing a new analysis to address the staff concerns and a program to monitor settlement in the area of the buried pipelines.
The staff will present the results of its review of these forthcoming submittals in a separate Safety Evaluation Report. This topic is included in the scope of ISAP issues for Millstone 1.
1 2.2.1 Turbine Building In IPSAR Section 4.2.1, the staff recommended that the licensee demonstrate that the pile foundation of the turbine building is adequate to resist lateral and uplift loads developed during a safe shutdown earthquake-in light of only a 4 inch pile embedment into the pile cap. Additionally, the staff recommended
- that potential corrosion of the pile be investigated considering the proximity of the sea water to the steel piles. The status of this issue is discussed above.
2.2.2 Gas Turbine Generator Building J
IPSAR Section 4.2.2 identified a concern, similar to that described in Section 2.2.1, related tc the pile foundation of the gas turbine generator building. Because some of the piles under the gas turbine building are friction l
piles, the staff required the licensee to demonstrate that they will perform adequately during a dynamic loading considering the possible loss of strength
.of the saturated granular soils surrounding these piles during an earthquake.
The status of this review is discussed above.
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2.2.3 Buried Pipelines In IPSAR Section 4.2.3, the staff identified an area where safety-related buried pipelines may be supported by peat, a highly compressible material, and potentially subject to unacceptably large settlement. The licensee supplied a summary of-the results obtained from recent soil borings in their submittal dated March 16, 1984. The status of this issue is discussed above.
2.3 Topic III-1, Classification of Structures, Systems and Components (NUREG-0824 Section 4.3)
As implemented by Regulatory Guide 1.26,10 CFR 50 (GDC 1) requires that structures, systems and components important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.
In Section 4.3 of the IPSAR, the staff concluded that insufficient information existed to complete the topic review in certain areas and recommended that the licensee supply additional information and analyses on the following subjects:
1) radiography 2) fracture toughness i
3) valves 4) pumps 5) storage tanks l
The staff's position remains c: stated in the IPSAR, namely, the licensee should complete the evaluations and i. corporate the results in the Final Safety Analysis Report update.
If the results indicate that modifications are required, those actions should be reported to the staff. By letter dated April 11, 1985 the staff granted NNECO a six month extension on the required submittal date for the updated Final Safety Analysis Report..This exemption requires the licensee to submit a program plan for the update by October 11, 1985. This topic is included in the scope of ISAP issues for Millstone 1.
2.4 Topic III-2, Wind and Tornado Loadins s (NUREG-0824, Section 4.4) and III-4.A, Tornado Missiles (NUREG-082L, Section 4.7) 10 CFR 50 (GDC 2), as implemented by SRP Sections 3.3.1 and 3.3.2 and Regulatory Guides 1.76 and 1.117, requires that the plant be designed to withstand the effects of natural phenomena such as wind and tornados.
IPSAR Sections 4.4.1 through 4.4.6 identified some structures and components important to safety that did not meet current licensing criteria, i.e.
tornado winds of 300 mph and differential pressure of 2.25 psi. The following vulnerable areas were identified in the IPSAR:
1 reactor building steel structure above the operating floor 2
ventilation stack 3
effects of failure of non-qualified structures 4) components not enclosed in qualified structures i
5) roofs 4
6) load combinations The licensee responded to each of the above issues in submittals dated December 3, 1982, October 7,1983, February 2,1984, and March 16, 1984.
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Related to the tornado wind issue is the issue of tornado missiles.
IPSAR Section 4.7 identified structures and components considered vulnerable to tornado missiles. The staff's position was that the licensee must provide protection for sufficient systems and components to ensure the ability to safely shutdown the plant (i.e., to reach hot shutdown). The licensee addressed the issue of tornado missiles in their submittal dated December 2, 1983.
The licensee has proposed to rely on the isolation condenser as the primary means of achieving safe shutdown for tornado events. As part of this proposal, the licensee intends to provide a connection to the underground city water system and a missile protected pump in the reactor building to provide cooling water for the isolation condenser.
j In an evaluation dated November 25, 1985, the staff concluded that the licensee's proposal will provide adequate protection against tornado events.
That conclusion is based on (1) the low likelihood of windspeeds which would cause substantial plant damage, (2) the capability of the condensate and (ge tank (CST) and fire-water tank as alternative cooling water sources, stora
- 3) a protected cooling path to the city water system. However, because of the general importance of the CST and fire-water tank as cooling l
sources, the staff has recommended that the capability of the anchor bolts to provide substantial resistance against failure be confirmed.
The confirmatory analysis of tank anchor bolts and implementation schedule for the plant modifications associated with the connection to the city water system will be resolved in ISAP.
2.5 Topic III-3.A, Effects of High Water Level on Structures (NUREG-0824 Section 4.5) 10 CFR 50 (GDC 2), as implemented by SRP Section 3.4 and Regulatory Guide 1.59, 4
requires that plant structures be designed to withstand the effects of flooding.
IPSAR Section 4.5.1 identified additional forces that will be applied to the floodwalls as a result of flood levels associated with the FMd. This issue has been addressed in this Supplement in Section 2.1.
In Section 4.5.2 of the IPSAR, the staff required that the licensee determine whether plant structures are capable of withstanding hydrostatic and uplift forces in combination with other loads (e.g., earthquake) resulting from groundwater rising to the grade elevation.
By letter dated February 2, 1984, the licensee supplied the results of their analyses which concluded.that these structures would adequately withstand the hydrostatic and uplift forces.
The staff has reviewed the licensee's submittal and has identified in its October 7,1985 evaluation specific areas of the analysis where additional
- justification is required to ensure adequate margins to accomodate. uncertainties -
in the analysis techniques. This
..ue will be addressed as part of ISAP issue 1.19, Integrated Structural Analysis.
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2.6 Topic 111-4.8, Turbine Missiles (NUREG-0824 Section 4.8) 10 CFR 50 (GDC 4), as implemented by Regulatory Guide 1.115 and SRP Section 3.5.1.3 requires that structures, systems and components important to safety be appropriately protected from dynamic effects including missiles.
While a turbine inspection program was being established generically for General Electric (GE) turbines, the IPSAR concluded that the low pressure turbine discs and normally inaccessible parts which have not been inspected in the last three years should be inspected at the next refueling outage; based on the results of this inspection, the licensee was to propose a schedule for future inspections.
Additionally, the staff recommended that the main steam stop and control valves and reheat stop and intercept valves be disassembled and inspected at approximately 3 year intervals and be exercised at least weekly by full closure of the valves.
By letter dated September 29, 1982, the licensee informed the staff that inspections and tests of the main steam stop, reheat stop, and intercept valves were being performed in accordance with the staff's position; however, the control valves are not tested by fully closing the valves.
The staff requested in the IPSAR that the licensee evaluate the potential improvement in control valve availability associated with weekly full-closure testing and the feasibility of conducting such tests.
The licensee responded by letters dated June 22, 1983 and December 5, 1984.
In the June 22, 1983 letter, the licensee concluded that no changes to the current testing program are warranted for the following reasons:
At rated power and steam flow, the main steam control valve position is approximately 70% of full open.
Due to the design and flow characteristics of these valves, this valve position does not mean that 30% of flow capacity remains.
Close to rated steam flow, very small changes in steam flow result in large changes in valve position.
A major reduction in reactor power would be required to insure that control valve capacity is not reached, which could result in a reactor pressure transient and subsequent neutron hi-hi flux scram.
The above condition could be avoided by opening the turbine bypass valves during the test period.
This alternative is not desirable, however, due to the high potential for main condenser tube failure.
Past operating experience has demonstrated that bypassing steam to the main condenser at high flow rates can cause degradation and subsequent tube leaks, which may result in chloride contamination and high conductivity in the main condenser.
The response time of the pressure regulating system and the control valves has not been adequately demonstrated.
Additional testing would be required to determine whether the pressure regulating system will allow opening of the remaining control valves before a reactor pressure increase and possible reactor scram could occur.
Closure of one main steam control valve can result in an imbalance in inlet steam to the high pressure turbine.
The effect of this evolution on turbine reliablity is not known; however, NNEC0 has requeste:1 an evaluation of this transient from the turbine vendor.
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The main steam control valves serve to control reactor pressure and thus modulate constantly in response to changes in steam flow..Therefore, any malfunction in the main steam control valve would be noticed by the operator and corrective action would be taken.
The fact that these valves are continually changing position provides additional assurance that they will close properly on demand to isolate the steam supply from the turbine.
The main steam stop valves and reheat stop and intercept valves are cycled to the fully closed position once per week, thus assuring their reliability.
Closure of these valves alone would be sufficient to isolate the turbine and prevent a severe overspeed condition.
In the December 5, 1984 letter, the licensee stated that the "A" low pressure turbine had been ultrasonically inspected in 1982 and that the "B" low pressure and high pressure turbines' were similarly inspected in 1980.
During the 1984 refueling outage,.the high pressure turbine was inspected using the magnetic particle technique.
Based on the results of these inspections, GE has recommended that the "B" low pressure turbine be inspected during the 1985 refueling outage and that the "A" low pressure turbine be inspected during the 1987 refueling outage.
The high pressure turbi,ne will be inspected during the 1989 refueling outage.
This schedule provides for the inspection of all three turbines over the next three refueling outages.
The staff will complete its evaluation of this issue for Millstone 1 when the licensee submits the additional information being developed by the turbine manufacturer (GE), as described previously.
In a related effort, the staff is continuing its review of the generic methods being developed by GE to establish turbine surveillance and maintenance schedules.
When these generic efforts are complete, the results will be appropriately applied to all operating plants with GE turbines, including Millstone 1.
In the interim, the staff concludes that the proposed inspection schedules are acceptable; however,-the licensee should continue to monitor developments from the generic program and modify.the inspection schedules, as necessary.
2.7 Topic III-5.A, Effects of Pipe Break on Structures, Systems and Components Inside Containment (NUREG-0824 Section 4.9) 10 CFR 50 (GDC 4), as implemented by SRP Section 3.6.2, requires, in part, that structures, systems and components important to safety be appropriately protected against dynamic effects such as pipe whip and discharging fluids.
In IPSAR Section 4.9, the staff identified three areas requiring further evaluation. The three areas were related to:
- 1) cascading pipe breaks,
- 2) jet impingement, and 3) pipe whip. The licensee submitted the necessary information by letter dated April 15, 1983 and the staff issued its SER on June 29, 1983.
2.7.1 Cascading Pipe Breaks In IPSAR Section 4.9.1, the staff was unable to conclude whether cascading pipe breaks would produce conditions more severe than those analyzed by the limiting design basis loss-of-coolant accident (LOCA).
The staff also required that any leakage detection systems deemed necessary should be reviewed in conjunction with SEP Topic V-5, " Reactor' Coolant Pressure Boundary Leakage Detection."
2-6
The licensee's evaluation concluded that the emergency core cooling systems are physically isolated from each other in such a way that any cascading breaks that impact on one train could not affect the redundant train.
Also, leakage in one recirculation loop cannot impact the other loop and cascading breaks that impact on one recirculation loop cannot impact the other loop.
Therefore, sufficient safe shutdown capability exists to mitigate the consequences of cascading breaks.
The licensee also concluded that the ECCS core cooling analysis would not be significantly changed due to cascading breaks and that the potential for cascading breaks does not compromise the design basis accident analysis.
Based on a review of the licensee's evaluation the staff concludes that the licensee has presented adequate justification to provide reasonable assurance that cascading pipe break effects will not create conditions more severe than those analyzed.
2.7.2 Jet Impingement IPSAR Section 4.9.2 identified four areas of the licensee's jet impingement analysis which required further justification.
These areas are:
(1) The jet impingement model used by the _ licensee was based on a jet expansion caused by longitudinal breaks; current criteria require the consideration of both circumferential and longitudinal breaks.
(2)- In the case of circumferential breaks, jets in conjunction with pipe whip were not considered to sweep the arc traveled by the whip.
(3) The assumptions used by the licensee appears to refer only to steam jets rather than all high-energy lines.
(4) From the information presented, it was uncertain whether the jet impingement effects on the impinged target pipe system conformed with the pipe size criterion presented in the letter transmitted to the licensee on January 4, 1980.
The licensee's response to these issues was as follows:
(1) The licensee affirmed that its high energy pipe break analysis did, in fact, consider jet impingement due to both longitudinal and circumferential breaks.
(2) Jets resulting from circumferential breaks were assumed to sweep the arc traveled by the whipping pipe.
(3) Jet impingement effects were considered for all high energy line breaks.
(4) Although it is the licensee's position that the consideration of jet impingement independent of pipe size is not a valid concern, an evaluation of the effects of jet impingement on piping targets was performed.
Because of the physical separation of the ECCS trains, the licensee determined that jet impingement could not result in the loss of both trains of either safety system. Therefore, the licensee concluded that the plant could achieve a safe shutdown when considering the effects of jet impingement on piping without consideration of the ratio of pipe sizes.
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Based on the review of the information provided by the licensee, the staff concludes that these issues have been satisfactorily resolved.
2.7.3 Pipe Whip In IPSAR Section 4.9.3, the staff required that the licensee evaluate the potential for and the consequences of pipes whipping into the drywell liner.
The licensee's prior conclusion that high energy line breaks would not penetrate the drywell liner and thus violate containment integrity was based on a test of loads being applied to a spherical shell.
The staff recommended that further justification be developed to demonstrate that the test conditions were applicable to the_ pipe break scenarios for Millstone 1.
In the test, the load was applied over a 14" diameter area and the shell deformed until it made contact with the concrete backing (3" gap).
The staff was concerned that in the case of the application of a concentrated dynamic load over a small area, the steel plate may be perforated before the deformation could be backed up by the concrete shielding wall. Also, the te.t shell thickness was 3/4" and parts of the Millstone 1 liner are not as thick.
The licensee noted in their submittal dated April 15, 1983 that it is not
-possible for the broken end of a pipe to impinge directly on the liner since the broken end is never the leading surface of the pipe.
The thickness of the drywell liner plate varies from 11/16" to 1-7/16" and the gap between the 4
liner and the concrete is only 2".
Where the liner is less than 3/4",
interactions between large bore piping and the liner plate occur at oblique angles such.that only a fraction of the pipe whip energy is transmitted to the plate.
Based on review of this response and the analysis submitted previously, the staff concludes that the licensee has demonstrated that postulated high energy line breaks inside containment will not penetrate the drywell liner and, thus, this issue has been satisfactorily resolved.
2.8 Topic III-5.B, Pipe Break Outside Containment (NUREG-0824 Section 4.10) 10 CFR 50 (GDC 4) as implemented by SRP Sect ons 3.6.1 and 3.6.2 and Branch Technical Positions MEB 3-1 and ASB 3-1, requires in part, that structures, systems and components important to safety be designed to accommodate the dynamic effects of postulated pipe ruptures.
The issue from Section 4.10.2 of the IPSAR was that the jet impingement model used by the licensee was less conservative than that specified by current criteria.
In that section, the staff recommended that the licensee:
- 1) validate the Millstone 1 jet impingement evaluation methods, 2) demonstrate that the differences between the criteria used and those in SRP Section 3.6.2 are not significant from the standpoint of consequences'on systems, or
-3) perform augmented ISI to demonstrate that unstable pipe failure is unlikely-and implement _ local leakage detection.
The !teensee sutmitted the results of its evaluation by letter dated March 24, 1983, and the staff issued its SER on June 29, 1983.
2-8 e
i The licensee stated that the difference in cross-sectional area in its model does not significantly alter the results, because conservatisms in the analysis (jet propogation losses, consequence evaluation) more than co.npensate for any non-conservatism in the jet expansion model.
Typically, jet impingement loads were determined for interactions with structures.
In these cases, either the load resulted in failure (so remedial measures were required) or there was substantial margin. Thus, based on review of the licensee's March 24, 1983 submittal, the staff has not identified any jet impingement interactions that were judged acceptable by the licensee's analysis that would result in unacceptable consequences if the jet load were proportionately increased to the change in cross-sectional area consistent with current criteria.
The staff, therefore, considers this issue acceptably resolved.
2.9 Topi'c III-6, Seismic Design Considerations (NUREG-0824 Section 4.11)
.10 CFR 50 (GDC 2) and 10 CFR 100 Appendix A, as implemented by SRP Sections 2.5, 3.7, 3.8, 3.9, and 3.10 and SEP review criteria (NUREG/CR-0098), require that structures, systems and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes.
IPSAR Section 4.11 identified the following specific areas as requiring further analysis:
1) pile foundations 2) motor operated valves 3) transformer and control room panels 4) qualification of cable trays 5) reactor vessel internals The licensee has submitted the results of analyses, as described in the following subsections, to demonstrate that the structures and equipment of concern are capable of withstanding ? he site-specific earthquake for Millstone 1, developed as part of the Topic III-6 evaluation. The staff's review of this information is nearly complete.
In a few cases, additional information is necessary to complete the review.
The staff will present the results of this review in a separate Safety Evaluation Report.
Information developed from these analyses is also related to the foundation capacity (Section 2.2) and load combinations (Section 2.10).
The results of these related evaluations will be addressed in ISAP, as discussed in the introduction to this report.
2.9.1 Pile Foundations IPSAR Section 4.11.1 identified concerns related to the adequacy of the pile foundation of the turbine building.
The concerns are the same as those discussed in Section 4.2 of the IPSAR and are addressed in Sections 2.2.1 and 2.2.2 of this supplement.
2.9.2 Motor-0perated Valves In Section 4.11.2 of the IPSAR, the staff required that the licensee demonstrate that the structural integrity of motor operated valves that are in lines 4 inches or smaller is adequate.
By letter dated May 17, 1984, the licensee submitted tte results of their review to the staff.
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2.9.3 Transformers and Control Room Panels Section 4.11.4 of the IPSAR required demonstration of anchorage system adequacy for transformers and control room panels during a seismic event.
The licensee had previously provided information to the staff by letter dated September 29, i
1982, which was under review at the time of IPSAR publication.
Subsequently, the staff requested additional information from the licensee regarding this matter.
2.9.4 Qualification of Cable Trays IPSAR Section 4.11.6 required the licensee to submit a plant-specific implementa-tion program and schedule regarding the seismic qualification of cable trays.
The licensee submitted final reports, on behalf of the SEP Owners Group, describing cable tray testing and results of the program on qualification of cable trays.
The licensee has concluded that the testing and analyses conducted as part of the SEP Owners Group program adequately demonstrates the seismic capability of the cable trays in SEP plants.
The staff is reviewing the results of the SEP cable tray program in concert with the generic resolution of Unresolved Safety Issue (USI) A-46, " Seismic Qualification of Equipment in Operating Plants," which will apply to all operating plants. When a generic resolution for USI A-46 is approved, the staff will determine whether any additional action is required to resolve the SEP cable tray program in a consistent manner.
2.9.5 Reactor Vessel Internals In IPSAR Section 4.11.8, the staff was unable to conclude that reactor vessel internals are adequate to resist a seismic event due to a lack of information.
The licensee provided the required information by letter dated January 23, 1984.
2.10 Topic III-7.B, Design Codes, Design Criteria and Load Combinations (NUREG-0824 Section 4.12) 10 CFR 540 (GDC 1, 2 and 4), as implemented by SRP Section 3.8, requires that structures, systems and components be designed for the loadings they may experience and that they conform to applicable codes and standards.
As described in IPSAR Section 4.12, during the integrated assessment, the licensee proposed to perform, on a sampling bash, an evaluation of the code, load and load combination issues delineated by de staff in order to assess the adequacy of as-built structures at Millstone 1.
By letter dated February 2,1984, the licensee supplied the results of their review.
The staff's review of this information has identified a few areas where additional information or justification is necessary in order to complete the evaluation and demonstrate that the margin of safety is adequate for the affected codes, criteria and load combinations.
The results of this review will be presented in a separate Safety Evaluation Report.
2.11 Topic III-10. A, Thermal Overload Protection for Motors of Motor-0perated Valves (NUREG-0824 Section 4.14) 10 CFR 50.55a(h), as implemented by Institute of Electrical and Electronics l
Engineers (IEEE) Standard 279-1971 and 10 CFR 50 (GDC 13, 21, 22, 23, and 29),
i requires that protective actions be reliable and precise and that they satisfy 2 - 10 l
l
the single failure criterion using quality components.
Regulatory Guide 1.106 presents the staff position on how thermal-overload protection devices can be made to meet these requirements.
In IPSAR Section 4.14, the staff found that, of 59 safety-related motor-operated valves (MOVs), 12 are not normally in their emergency position and have thermal-overload protection devices that are not bypassed by an emergency signal nor has it been shown that their trip setpoints were conservatively set.
The licensee proposed to analyze the trip setpoints and modify, where necessary, thermal overload protection devices.
The analysis is to include a demonstration that proper thermal overload protection devices have been selected, that their trip setpoints have been conservatively set and a summary of the operating experience for each of the 12 valves.
By letter dated August 13, 1985, NNEC0 provided an assessment of this issue.
As discussed in the safety evaluation dated September 16, 1985, the staff concludes that the thermal overload setpoints have been conservatively established such that there is reasonable assurance the valves will operate under accident conditions. Therefore, this issue is resolved.
2.12 Topic VI-4, Containment Isolation System (NUREG-0824 Section 4.20) 70 C'T 50 (GDC 54, 55, 56, and 57), as implemented by SRP Section 6.2.4 and r.
lory Guides 1.11 and 1.141, requires isolation provisions for the lines penetrating primary containment to maintain an essentially leaktight barrier against the uncontrolled release of radioactivity to the environment.
2.12.1 Remote Manual Valves IPSAR Section 4.20.3 identified various lines that are equipped with remote manual valves rather than automatic isolation valves.
Since these lines serve an essential emergency core cooling function, the staff agrees that automatic isolation valves should not be used. However, the staff concluded that adequate leakage detection and procedures for operator action should be demonstrated and the operating station should be located in an accessible area, where necessary.
The licensee responded to this issue by letter dated February 23, 1983.
Based on this information, the staff's SER dated July 7, 1983, concluded that adequate detection devices exist in the form of pressure and f?ow indications, surge tank levels, and sump alarms for the lines in question.
These alarms and the operating station for the valves in question are located in the control room.
Thus, operation of these valves is possible during an accident.
Additionally, plant procedures exist which describe the circumstances under which these lines should be isolated. Therefore, the staff considers this issue adequately resolved.
2.12.2 Lack of Information In IPSAR Section 4.20.7, the staff identified branch lines associated with two penetrations where the isolation capability was unknown.
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+
y In a Ietter dated June 22, 1983, th.e licensee concluded that,. adequate isolation l
c exists for penetration X-211A because the branch lines off this line have two normally closed valves in series' outside the containment but that adequate isolation does not exist for penetration X-204.
To correct this deficiency, the ~ licensee has proposed to move the branch lines in question from their current point of connection. to'rew branching positions outside of the containment isolation boundary valves.
s, The licensee has further requested that this modification be ~ deferred to ISAP, s
l as described in the introduction.to this report, to reassess the benefits of this modification relative.to re11ted issues.
The staff agrees that this deferral is appropriate.
2.'13 ; Topic VI-7.A.3, Emergency Core Cooling System Actuation System (NUREG-0824 Section 4.21) 10 CFR 50.55a(h), as implemented by IEEE Standard 279-1971 and 10 CFR 50 Appendix A (GDC 37), as iniplemented by Regulatory Guide 3:22, require that equipment important to safety be tested periodically ~to ensure the operability of the system as a wh' ole and to verify, under conditions as close to design as pract! cal, the performance of the full operational sequence that brings the system into operatisn, including the operation of the associated cooling water e
system.
IPSARSection4.21.1concludgdt[hattheMillstone1TechnicalSpecifications did not require the testing of the core spray system pump space coolers which are part of the turbine building secondary closed cooling water system (cooled by the service water system).,
~
The licensee provided information by letter dated December 3, 1982, to demonstrate that ventilation is not required in the corner rooms of the reactor building to, ensure core spray and LPCI pump operability.
The basis for this conclusion relies on tests that were performed during initial plant start-up in late 1970.
The staff reviewed this information and transmitted the results of this review to the licensek by)etter dated July 5,1983.
Even though worse conditions 1
than those.used in the test are possible,Much conditions coincident with the need for this equipment are unlikely. Additionally, redundant equipment, serviced.by'a separate cooling system, is available to perform the safety functior.:
Onithis basis, the staff concludes that this issue has been y
adequately < resolved.
h 2.14 Topic /I-7.C.1, Appendix K-Electrical Instrumentation and Control Rr;-revi9ws (NUREG-0824 Section 4.23)
.( 10 CFR 50 fGDC 2, 4r 17, and 18) as implemented by SRP Sections 8.2 and 8.3 and Regulatory 1 hide IA, requires that redundant load grcups and the redundant standby electried " power sources be independent' at least to the following extent:
y
-(1)_Noprovisionsshouldehistforautomaticallyconnectingoneloadgroupto another load group (.
r 2 - 12
'g.
s-
(2) No provisions should exist for automatically transferring loads between redundant power sources.
(3) If means exist for manually connecting redundant load groups together, at least one interlock should be provided to prevent an operator error that would parallel their standby power sources.
The bases for these requirements include the following:
(1) There is evidence based on operating experience and analytical considera-tions that the parallel operation of standby power sources renders them vulnerable to common-mode failures.
Current designs are, therefore, based on the concept of independent, redundant load groups.
In these designs, the standby power source for one load group is never automatically interconnected under accident conditions with the standby power source of a redundant counterpart.
(2) There can be compromises of independence resulting from automatic bus ties that connect the loads of one load group to the power source of another in the event the power source of the first load group has failed.
The slightly improved defense against random failures achieved by these bus ties is more than offset by the additional vulnerability to common-mode failures that they create.
2.14.1 Automatic Bus Transfers IPSAR Section 4.23.1 identified buses that are supplied by automatic-transfer switches that can transfer loads between redundant sources.
The staff required that the licensee evaluate the design of the automatic bus transfer (ABT) devices and identify any necessary corrective action.
In a letter dated April 12, 1983, as supplemented by a letter dated December 6, 1983, the licensee identified actions such as a breaker coordination study to resolve this issue.
The licensee stated that a project is currently planned at Millstone 1 to analyze and document the circuit breaker coordination for the A8Ts.
The results of the licensee's circuit breaker coordination analysis and possible corrective actions will be evaluated with related issues in ISAP, as described in the introduction to this report.
As discussed in IPSAR Sections 4.23 and 4.26, Millstone 1 has a single instrument ac bus. 'Thus, failure of this bus may result in loss of essential instrumentation needed for plant shutdown.
This issue was addressed by taking credit for local indications of vital parameters to enable plant operators to achieve and maintain safe shutdown conditions in the event of bus failure. As part of the ISAP review of bus transfers,
'the need for a redundant instrument ac bus will be evaluated by the licensee.
i 2 - 13
t 8
2.14.2 Manual Bus Transfers In IPSAR Section 4.23.2, the staff identified three load centers that are manually transferred between redundant sources.
Although they are under administrative control, there are no interlocks to prevent operator error that would parallel the emergency sources.
By letter dated April 12, 1983, the licensee concluded that the present configuration provides an adequate level of independence between the dc
. systems because the plant operates with the emergency transfer source deenergized.
The staff will evaluate the licensee's conclusion for,this issue relative to related issues under ISAP, as described in the introduction to this report.
2.15 TopicVII-1.A, Iso 1'ationofReactorProfectionSystemFromNon-Safety Systems, Including Qualifications of Isolation Devices (NUREG-0824 Sectiorn 4.25) 10 CFR 50.55a(h),sthrough IEEE Std. 279-1971, requires that safety signals be isolated from nonsafety signals and that no credible failure at the output of an isolation device shall prevent the associated protection system channel from meeting the minimum performance requirements specified in the design bases.
IPSAR S ctim 4.25.1 concluded that at Millstone Unit 1, there were no isolation devices betw'een the nuclear flux monitoring systems and the process recorders and indicating instruments, nor were there any between the average power range monitor (APRM) system and the process computer.
s The licensee proposed to conduct tests to determine if adequate isolation exists for these circuits.
By letter dated January 31, 1984, the licensee supplied the results of their testing and their bases for concluding that modifications are not warranted.
In summary, the licensee concluded that although isolation between the components does not exist, the probability of. hot shorts 'of 125V dc or 120V ac is remote and electrical failures due to inadequate isolation do not contribute significantly to the overall Yeactor protection system (RPS) failure probability.
The staff will evaluate the licdnsee's conclusion for this issue: relative to' relc.ted issues under ISAP, as described in the introduction to this report.
2.16 Topic VIII-2, Onsite Emergency Power Systems (NUREG-0824_ Section 4:28) i 10 CFR 50 (GDC 17), as implemented by SRP Section 8.3 1 and BTP ICSB 17, requires that:
,7 (1). The deitgn of standby diesel generator systemdh$uld retain 'only the U engine overspeed and the generator differentia'l trips and bypass all
- I other trips under an accident condition.
',(2)
If other trips, in addition to the engine overspeed and generator differential trips, are retained for accident conditions, an acceptable design should provide two or more independent' measurements of each of the trip parameters.
Trip logic should be such that a diesel generator trip would require specific coincident logic.
(
)
2 - 14 N
('
A.
In addition, GDC 17, as implemented by IEEE Std. 279-1971, requires that all the conditions that night render the emergency power generato" incapable of automatic starting shall be unambiguously annunciated in the control room.
All current licensing criteria for emergency onsite power have been developed for a diesel generator. At Millstone Unit 1, one of the two emergency onsite generators is powered by a gas turbine. There are no specific staff criteria for a gas turbine generator.
In IPSAR Section 4.28.3, the staff noted that many of the types of gas turbine failures that have occurred may be eliminated with an effective preventive maintenance program.
Thus, the staff concluded that such a program should be implemented or, if one already exists, the licensee should review the program for possible improvement.
The licensee responded by letters dated May 10, 1983 and June 20, 1983.
In these responses, the licensee proposed changes in their preventive maintenance program regarding the engine mounted fuel pump, fuel shutoff valve, air start regulating valve, air receiver and control circuit timers.
By letter dated August 16, 1984, the staff concluded that the revised maintenance program acceptably resolves this issue.
Bypass under accident conditions of the light-off speed and generator excitation trips, the high lube oil temperature trip and of five gas turbine generator output breaker protective trips is issue 1.01 in ISAP.
2.17 Topic VIII-3.B. DC Power System Bus Voltage Monitoring and Annunciation (NUREG-0824 Section 4.30) 10 CFR 50.55a(h), through IEEE Std. 279-1971, and 10 CFR 50 (GDC 2, 4, 5,17, 18, and 19), as implemented by SRP Section 8.3.2, Regulatory Guides 1.6, 1.32, 1.47,1.75,1.118, and 1.129, and BTP ICSB 21, require that the control room operator be given timely indication of the status of the station batteries.
In IPSAR Section 4.30 the staff concluded that the Millstone 1 control room did not have all of the alarms and indications of dc power system status recomended by current criteria. As discussed in IPSAR Section 4.30.1, the licensee evaluated the need to provide battery current alarms / indications in their December 27, 1982 submittal, and concluded that present indications in the control rcom are sufficient.
The staff will consider the licensee's conclusion for this issue relative to related issues under ISAP, as described in the introduction to this report.
2.18 Topic IX-S, Ventilation Systems (NUREG-0824 Section 4.32) 10 CFR 50 (GDC 4, 60 and 61), as implemented by SRP Sections 9.4.1, 9.4.2, 9.4.3, 9.4.4, and 9.4.5, requires that the ventilation systems shall have the capability to provide a safe environment for plant personnel and for engineered safety features. The staff's review of the ventilation systems for Millstone 1 found them acceptable except for the four items discussed below.
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The Millstone Interim Reliability Evaluation Program (IREP) considered failure of safety systems resulting from inadequate ventilation and did not identify any systems where ventilation was a concern.
However, additional information was necessary in some cases to confirm the assumptions and judgments in the Probabilistic Risk Assessment (PRA).
The licensee responded to the identified ventilation issues by letter dated April 18, 1983 and the staff issued its SER on July 5, 1983.
2.18.1 Core Spray and LPCI Systems Ventilation Systems This issue is identical to that discussed in Section 2.15 in this supplement.
Conclusions given in that section also apply here.
2.18.2 Reinitiation of Ventilation Following Loss of Offsite Power In IPSAR Section 4.32.2, the staff required that the licensee define the maximum period the turbine building ventilation system could be inoperative and still assure that the equipment serviced would continue to function, as necessary.
In addition, the staff concluded that the licensee should demonstrate that the amount of hydrogen generated as a result of battery charging during that period would not exceed the minimum combustion limit.
The results of the licensee's evaluation indicated that the components which contribute the largest heat loads in the turbine building during a loss of offsite power event are the 480 volt load centers 1, 2 and 2A, and the load center and switchgear heaters.
The licensee's analysis of the heat gain in the area of this equipment, during a summer period indicated that the ambient temperature would increase to 104 F (design temperature) in approximately 30 minutes.
Of these heat sources, all of the load center and switchgear heaters and 480 volt load centers 1 and 2 receive power from the gas turbine generator.
Only the 480 volt load center 2A is powered by the diesel generator.
The evaluation that was performed for this area assumed that both the diesel and the gas turbine start and run.
However, the licensee's evaluation found that, in the event the gas turbine is not started, the heat loads in this area would be significantly reduced (i.e.,
loss of heaters and load centers 1 and 2) to the point where ventilation in the area would not be required.
Therefore, since ventilation in this area is required only when the gas turbine generator is operating, and sufficient time (a minimum of 30 minutes) is available for operator action, the licensee committed to revise the plant operating procedures to instruct the operator to start both the supply and exhaust ventilation fans in the switchgear area if the gas turbine is started during a loss of offsite power event.
With respect to the accumulation of hydrogen, the licensee calculated the
. maximum rate of hydrogen production in each battery room and concluded that it would require considerable time for hydrogen concentration to reach a combustible concentration.
Due to the low hydrogen production rate and the presence of the combustible gas detector systems, the staff concludes that sufficient time (i.e., aays) is available for the operator to either restore offsite power or to successfully start the diesel generator to provide ventilation before the limit of combustion is reached.
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Based on the staff's review of this information and the licensee's commitment to modify the plant operating procedures, the staff considers these issues acceptably resolved.
2.18.3 FWCI and Diesel Generator Area Coolers In IPSAR Section 4.32.3, the staff indicated that insufficient information was provided on the design and operation of the area space coolers for the feedwater coolant injection (FWCI) and diesel generator areas.
Subsequently, the licensee evaluated the design and operation of both area cooling systems.
As a result of the licensee's review of the space coolers in the diesel generator area, it was determined that both area coolers are automatically sequenced into the diesel generator.
Each area cooler has a heat removal capacity of 500,000 BTU /hr, which is sufficient to cool this area without requiring the ventilation system to be operable.
Therefore, this issue is resolved.
The licensee's evaluation of the heat gain in the vicinity of the FWCI components, during the summer indicated that the area coolers (HVH 3, 3A, 4, 4A, 5, and 5A) must be operable to maintain these areas below the 104 F design temperature.
The cooling capacity of these area coolers is sufficient even with the loss of one cooler.
If the area coolers are operating, the ventilation system is not essential; however, the heat removal capacity of the ventilation system alone is not sufficient to cool these areas without operation of the area coolers.
The FWCI space coolers do not presently receive emergency power; consequently, the licensee committed, by letter dated April 18, 1983, to modify the units so that the area coolers will be automatically sequenced onto the gas turbine generator.
The proposed modification will resolve this issue; implementation will be scheduled in accordance with the ISAP review (issue 1.05).
2.18.4 Intake Structure Ventilation System In IPSAR Section 4.32.4, the staff indicated that sufficient information was not provided to support the licensee's position that active ventilation is not required for the intake structure.
The licensee subsequently evaluated the heat buildup in the intake structure during the summer due to operation of the service water and emergency service water pumps and has determined that starting one of the 20,000 cfm exhaust fans will provide sufficient cooling. Operation of only one fan will provide a complete air change in the affected area in approximately four (4) minutes.
The two 20,000 cfm exhaust fans (HVR 8 and 9) do not presently receive power from an emergency bus.
The licensee committed in the April 18, 1983 letter to modify.the exhaust fans so that one unit is automatically sequenced onto the diesel generator and the other unit is automatically sequenced onto the gas turbine generator.
The modification will resolve this issue; implcx.entation will be scheduled in accordance with the ISAP review (issue 1.05).
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Table 2.1 Summary of IPSAR and Supplement Evaluations SEP IPSAR Supplement Requirements Topic Section Requirements Section From No.
No.
Title From IPSAR No.
Supplement 11-3.8, 4.1.1 Flooding Elevation Determine the effects of 2.1.1 Assure structural adequacy of 11-3.B.1, probable maximum hurricane masonry walls in fire pumphouse.
II-3.C (PMH) wave inleakage and (ISAP Issue 1.19) identify any necessary corrective actions.
Provide analysis of PMi None wave structural effects.
(See Sections 4.6 and 4.12.)
4.1.2 Intake Structure Mone 4.1.3 Local Flooding None 4.1.4 Gas Turbine Building Nonc 4.1.5 Diesel Fuel 011 None (See Section 4.1.6.)
4.1. 6 Emergency Procedures Revise the flood emergency 4.0 Procedural change complete.
procedures to address the topic concerns and implement the revised procedures.
4.1.7 Roofs Determine the adequacy of 2.1.2 Install scuppers on selected ro roofs subjected to ponding roofs. (1985 Refueling Outage)
L resulting from the local m
probable maximum precipitation (PMP). (See Section 4.12.)
II-4.F 4.2.1 Turbine Bu11dina Evaluate structural capability 2.2.1 Under staf f review. (ISAP issue of the piles supporting the 1.19) building. (See Section 4.12.)
4.2.2
. Gas Turbine Generator Evaluate structural capability 2.2.2 under staff review. (ISAP Issue Building of the piles supporting the 1.19) building. (See Section 4.12.)
Tab 13 2.1 Summary att IPSAR and Supplement Evaluations SEP IPSAR Supplement Requirements Topic Section Requirements Section From No.
No.
Title From IPSAR No.
Supplement 4.2.3 Buried Pipelines Conduct soll investigation in 2.2.3 Awaiting information from area of the safety-related Ilcensee. (ISAP Issue 1.19) water pipelines where they may be underlain by peat. (See Section 4.12.)
III-1 4.3.1 Radiography Requirements Perform a volumetric inspection 2.3 Same (ISAP Issue 1.15) of all Class 1 and 2 piping, and valves and Class 2 vessels not volumetrically inspected previously. Document in FSAR update.
4.3.2 Fracture Toughness Identify and replace, if 2.3 Same (ISAP Issue 1.15) necessary, the components that do not meet fracture toughness requirements.
Document in FSAR update.
'4.3.3 Valves Evaluate design of Class 1, 2.3 Same (ISAP Issue 1.15) 2, and 3 valves on a sampling basis; upgrade if necessary.
N Document in FSAR update.
a e
4.3.4 Pumps' Analyze the design safety 2.3 Same (ISAP Issue 1.15) margins of the specified pumps. Document in FSAR update.
4.3.5 Storage Tanks Evaluate design of specified 2.3 Same (ISAP Issue 1.15) tanks. Document in FSAR update.
t 4
Tabla 2.1 Summary of IPSAR and Supplement Evaluations SEP-IPSAR Supplement Requirements Topic Section Requirements Section From No.
No.
Title From IPSAR No.
Supplement III-2 4.4.1 Reactor Building Steel Analyze the specified-2.4 Provide protected safe shutdown Structures Above the structures capabilities to train and assure adequacy of CST Operating Floor resist tornado loads and and firewater tank anchor bolts.
propose corrective actions, (ISAP Issue 1.02) if necessary. (See Section 4.12.)
4.4.2 Ventilation Stack Submit analyses demonstrating 2.4 Same as 4.4.1 capability to achieve and maintain safe shutdown of Units 1 and 2 case of a tornado-induced failure of the stack. (Analysis submitted--
under staff review.)
4.4.3 Effects of Failure of Provide an analysis of the 2.4 Same as 4.4.1 Nonqualified Structures effects and any corrective actions that may be necessary.
4.4.4 Components Not Enclosed Determine the adequacy of the 2.4 Same as 4.4.1 ro in Qualified Structures components and identify any
/o corrective actions that may O
be necessary.
4.4.5' Roofs Determine the adequacy of roofs 2.4 Same as 4.4.1 of Category I structures. (See Section 4.12.)
4.4.6 Load Combinations Demonstrate that wind loads
- 2. 4 See Section 2.10.
were properly combined with (ISAP Issue 1.19) other specified loads or identify any necessary corrective action.
(See Section 4.12.)
t i
l Table 2.1 Summary of IPSAR and Supplement Evaluations i
SEP IPSAR Supplement Requirements Topic Section Requirements Section Froe No.
No.
Title From IPSAR No.
Supplement l
Ill-3.A 4.5.1 Flood Elevation Provide analysis of PMH wave 2.5 See Section 2.1.1 structural effects and identify any necessary corrective actions. (See Sections 4.1.1 and 4.12.)
4.5.2 Groundwater Demonstrate appropriate 2.5 Provide additiona justification consideration of hydrostatic of margin (ISAP !ssue 1.19) forcas on a sampling basis.
(See Section 4.12.)
III-3.C 4.6.1 Deficiencies Noted Determine the adequacy of roofs 2.1.2 Install scuppers on selected roofs.
subjected to ponding resulting (To be installed during Fall 1985 from the local PMP. (See Section refueling outage) 4.12 and 4.1.7.)
4.6.2 Structures and Components Revise procedure to include 4.0 Procedural change complete.
Requiring Inspection inspection of floodwalls.
i l
(See Section 4.6.3.)
- 4. 6. 3 -
Inspection Program Develop and submit an improved 4.0 Inspection program complete.
inspection program for water y
a control structures.
N
~
III-4.A 4.7 Tornado Missiles Evaluate protection of systems 2.4 Provide safe shutdown train.
.and components to ensure the capability to safely shutdown the plant via a tornado-missile-protected path.
Modify as necessary to provide 2.4 Provide safe shutdown train.
a tornado-missile protected See Section 2.4 (ISAP Issue 1.02) shutdown path.
III-4.8 4.8 Turbine Missiles Inspect turbine and propose 2.6 None frequency based on results.
m l
l Tab 12 2.1 Summary tf IPSAR and Supplement Evaluations SEP IPSAR Supplement' Requirements Topic
.Section Requirements Section From No.
No.
Title From IPSAR No.
Supplement III-4.8 4.8 Turbine Missiles
. Evaluate the improvement in 2.6 Awaiting further information control valve availability from turbine manufacturer.
associated with full closure testing and feasibility of conducting such tests.
III-5.A 4.9.1 Cascading Pipe Breaks Subett an analysis of cascading 2.7.1 Mone pipe breaks.
4.9.2 Jet Impingement Provide information specified.
2.7.2 None 4.9.3 Pipe Wip Provide an analysis of the 2.7.3 None potential for and consequences of pipes whipping into the drywell liner.
III-5.8 4.10.1 Moderate-Energy Piping None 4.10.2 Jet !apingement Subelt a review of affected jet 2.8 Mone impingement analysis, p
4.10.3 Unisolable Breaks None N
III-6 4.11.1 Pile Foundations Evaluate structural capability 2.9.1 None (See Section 2.2.1 and 2.2.2) of piles supporting the turbine (ISAP Issue 1.19) and gas turbine buildings.
(See Sections 4.2.1, 4.2.2, and 4.12.)
4.11.2 Motor-Operated Valves Demonstrate valve structural 2.9.2 Under staff review, integrity.
(ISAP !ssue 1.06) t 4.11.3 tow-Pressure Coolant Mone Injection / Containment Spray Heat Exchangers
Table 2.1 Summary of IPSAR and Supplement Evaluations SEP IPSAR Supplement Requirements Topic section Requirements section From No.
No.
Title From IPSAR No.
Supplement III-6 4.11.4 Transformer and Control Mone (staff is reviewing).
2.9.3 Under staff review.
Room Panels (ISAP issue 1.06) 4.11.5 Ability of Safety-Related None Elactrical Equipment to Function.
4.11.6 Qualification of Cable Provide plan to loplement 2.9.4 None Trays results of SEP Dwners Group Program.
4.11.7 Recirculation Pump Supports None 4.11.8 Reactor Vessel Internals Provide a seismic analysis of 2.9.5 Under staf f review.
the reactor vessel internals.
(ISAP Issue 1.06) 111-7.8 4.12 Design Codes, Design Evaluate adequacy of original 2.10 Under staf f review.
Criteria, Load Combinations design criteria on a sampling (15AP Issue 1.19) and Reactor Cavity Design basis for specified structural Criteria elements; provide information requested in Topics 11-3.8, as II-4.F. III-2. III-3.A. and
[a 111-6 that has been deferred to LJ this topic.
III-8.A 4.13 Loose-Parts Monitoring and None Core Barrel Vibration Monitoring III-10.A 4.14 Thereal-Overload Protection Demonstrate proper setting 2.11 None for Motors of Motor-Operated of thermal-overload trip Valves setpoints and discuss operating experience of specified valves.
Implement modifications found to be necessary.
Table 2.1 Summary of IPSAR and Supplement Evaluations SEP IPSAR Supplement Requirements Topic Section Requirements Section From No.
No.
Title From IPSAR No.
Supplement IV-2 4.15 Reactivity Control Systees,.
None Including Functional Design ahd Protection Against Single Failures V-5 4.16.1 Systems Currently Available Provide at least one leakage 3.1 Technical Specification change at Mllistone Unit I detection method that is complete (Amendment 97).
qualified to a safe shutdown earthquake or provide procedures that specify actions to be taken for a satselc event and failure of the leakage detection equip-ment (e.g., plant shutdown).
The leakage detection method should be testable during operation.
Evaluate sensitivity in conjunc-3.1 None tion with Topic III-5.A.
(See Section 4.9.1.)
4.16.2 Intersystem Leakage None V-10.B 4.17 Residual Heat Removal Review and implement emergency 4.0 Procedural change complete System Reliability procedures, including steps to proceed to a cold shutdown condition from outside the control roce.
V-11.A 4.18 Requirements for Isolation Install an independent pressure 4.0 Region I outstanding Ites No.
of High-and Low-Pressure interlock for the reactor water 245/84-27-05. (ISAP Issue 1.04)
Systees cleanup (RWCU) system inboard suction isolation valve.
Table 2.1 Summary of IPSAR and Supplement Evaluations SEP IPSAR Supplement Requirements Topic Section Requirements Section From No.
No.
Title From IPSAR No.
Supplement V-12.A 4.19.1 Water Chemistry Limits Revise Technical Specifications 3.2 Technical Specification change to incorporate RG 1.56 Ilmits complete (Amendment 99).
for chlorides and conductivity limits, or provide justification for not doing so.
4.19.2 Limiting Condittuns for Mone Operation VI-4 4.20.1 Locked-Closed Valves Install administratively 4.1 Complete including revisions controlled mechanical locking to IPSAR requirements noted in devices in the specified valves.
Section 4.0.
4.20.2 Lines Requiring a Second Install a second valve and 4.2 Complete except for Region I Valve and Both Locked administrative 1y controlled outstanding Item No. 245/84-27-08.
Closed locking devices on both, on the specified lines.
4.20.3 Remote Manual Valves Demonstrate leakage detection, 2.12.1 None locate operating stations in y
accessible areas, and develop procedures for isolation of the specified valves.
4.20.4 Valve Location None 4.20.5 Instrument Lines tione 4.20.6 Valve Location and Type None 4.20.7 Lack of Information Review isolation capability of 2.12.2 Relocate branch line connection two lines and implement points. (ISAP Issue 1.03) modifications, if necesary.
Tabis 2.1 Suemary af IPSAR and Supplement Er21uations SEP
-IPSAR Supplement Requirements Topic Section Requirements Section From No.
No Title From IPSAR
- No.
Supplement VI-7.A.3-4.21.1 Testing of Space Coolers Demonstrate that the space 2.13 None coolers are not essential.
(Analysis subeltted--under staff review.)
4.21.2 Testing of the ESWS Mone VI-7.A.4 4.22 Core Spray Moz21e None Effectiveness VI-7.C.1 4.23.1 Automatic Bus Transfers Evaluate the existing automatic 2.14.1 Perfore circuit breaker bus transfers and identify coordination analysis corrective actions to ensure including need for faulted loads would not be instrument ac bus (ISAP transferred.
Issues 1.21 and 2.17) 4.23.2 Manual Bus Transfers Install appropriate interlocks 2.14.2 Same as above.
or provide justification for not (ISAP !ssue 2.19) doing so.
VI-10.A 4.24.1 Surveillance Frequency Evaluate the surveillance 4.0 None frequency of the specified 7
- channels, ro C5 4.24.2 Channel Functional Test None Frequency 4.24.3 Response-Time Testing Mone VIII-1.A 4.25.1 Isolation Devices Between Conduct test to determine if 2.15 (ISAP Issue 1.22)
Reactor Protection Systen existing isolation is adequate.
(RPS) and Monitoring Propose corrective actions if Systems necessary.
4.25.2 Isolation Devices Between Provide adequate isolation.
4.0 Mone the RPS and its Power Supply
.m.
i_
9 Table 2.1 Summar:y of IPSAR and Supplement Evaluations
.SEP IPSAR Supplement Requirements Topic Section Requirements Section From No.
. No.
Title From IPSAR No.
Supplement VII-3 4.26 Systees Required for Safe None Shutdown VIII-1.A 4.27 Potential Equipment Failures Develop and ig iement procedures 4.0 None Associated with Degraded to protect Class if systems of a Grid Voltage degraded grid volt.ge condition.
7 VIII-2 4.28.1-
.Startup Trips Bypass light-off speed and 4.3 Region I outstanding Ites No.
ro generator excitation speed 245/84-27-07. (ISAP Issue 1.01) trips under accident conditions.
4.28.2
_0perational Trips Bypass high lube oli temperature 4.3 Region I outstanding Item No.
trip under accident conditions.
245/84-27-07. (ISAP !ssue 1.01) 4.28.3 Gas Turbine Preventive laplement a preventive mainten-2.16 None Maintenance Program ance program, improve existing
/
one, or provide justification for y.
not doing so,
'4.28.4 Generator Trips Bypass specified trips under 4.3 Region I outstanding Ites No.
p accident conditions.
245/84-27-08. (ISAP Issue 1.01) 4.28.5 Annunciators None VIII-3.A 4.29
. Station Battery Test Revise Technical Specifications 3.3 Technical Specification change Requirements to require battery service and complete (Amendment 99),
discharge tests.
VIII-3.8 4.30.1 Battery Status Alarms and Install the specified battery 2.17 Under staf f review.
Indications.
status alarms or provide justification for not doing so.
T
+..
~
~
Tabi) 2.1 Summary cf IPSAR and Supplement Eviluations SEP IPSAR Supplement Requirements Topic Section Requirements Section From peo.
No.
Title from IPSAR No.
Supplement VIII-3.8 4.30.2 Battery Outage Limits Revise Technical Specifications 3.4 Proposed Technical Specifications to reduce battery outage limits not received from Itcensee.
or provide justification for present lielts.
IX-3 4.31 Station Service and Cooling leone (pending results of Topic Water Systems II-4.F review).
IX-5 4.32.1 Core Spray and LPCI Systees Demonstrate that the space coolers 2.18.1 See Section 2.15.
Ventilation are not essential. (See Section 4.21.1.) (Analysis submitted--
under staff review.)
4.32.2 Reinitiation of Venttiation Demonstrate that the equipment 2.18.2 peodify operating procedures After a Loss-of-Offsite-serviced is unaffected by the to start exhaust fans.
Power Event lack of ventilation and that the hydrogen combustion limit in the battery rooms will not be reached.
m 4.32.3 Lack of Information Provide information on the space 2.18.3 (ISAP Issue 1.05) coolers for the feedwater S!
coolant injection and diesel generator areas.
4.32.4 Intake Structure Ventilation Demonstrate that sufficient 2.18.4 (ISAP Issue 1.05)
System ventilation can be provided in a timely eenner.
XV-1 4.33 Decrease in Feedwater leone currently; surveillance of Temperature Increase in turbine bypass valves and limits Flow, Increase in Steam for reactor power if the turbine Flow, and Inadvertent bypass is inoperable will be Opening of a Steam Generator required if credit is taken for Relief or Safety Valve the turbine bypass in the reload analysis.
t
-~
Table 2.1 Sumary of IPSAR and supplement Evaluations
(
SEP IPSAR Supplement Requirements
. Topic Section Requirements Section From i
No.
No.
Title From IPSAR No.
Supplement l
Mw-3 4.34 Loss of External Load None Turbine Trip, Loss of l
Cendenser Vacuum, Closure I
of Main Steam Isolation Valve (BWR), and Steam
. Pressure Regulator Failure (Closed)
MV-16 4.35 Radiological Consequences laplement BWR Standard 3.5 Technical Specification change of Failure of Small Lines Technical Specification limit complete (Amendment 99).
Carrying Primary Coolant for primary coolant activity Outside Containment and propose, for staff review, associated action statements if these 11mits are exceeded.
XV-18 4.36 Radiological Consequences See Section 4.35.
3.6 See Section 3.5 of a Main Steam Line Failure Outside Containment.
7 m
W
3 IPSAR TOPICS RESOLVED BY CHANGES TO PLANT TECHNICAL SPECIFICATIONS During the Integrated Assessment for Millstone 1, a number of issues were resolved by commitments from the licensee to perform evaluations in order to determine whether modifications to plant Technical Specifications are warranted.
In other cases, the licensee committed to adopt the corresponding requirements of the Standard Technical Specifications and the issue was considered resolved during the Integrated Assessment.
This section describes the actions taken regarding resolution of IPSAR issues involving Technical Specification changes.
A summary of all changes to Technical Specifications and the associated license amendment is given in Table 3.1.
3.1 Topic V-5, Reactor Coolant Pressure Boundary Leakage Detection (NUREG-0824 Section 4.16) 10 CFR 50 (GDC 30) as implemented by Regulatory Guide 1.45 and SRP Section 5.2.5, prescribes the types and sensitivity of systems and their seismic, indication, and testability criteria necessary to detect leakage of primary reactor coolant to the containment or other interconnected systems.
In IPSAR Section 4.16, the staff concluded that:
(1) The licensee should provide a seismically qualified method for determining RCPB 1eakage or provide procedures that specify actions to be taken in the case of a seismic event and failure of the leakage detection equipment (e.g., plant shutdown).
(2) The method should be testable during operation.
(3) The licensee should evaluate leakage detection sensitivity requirements in conjunction with the resolution of Topic III-5.A for the purpose of establishing appropriate limiting conditions for operation.
By letter dated November 1, 1983, the licensee proposed changes to the plant Technical Specifications to address the matter.
The staff approved the Technical Specifications by letter dated May 23, 1984, which transmitted Amendment 97 to the licensee.
The Technical Specification change requires initiation of an orderly plant shutdown if leak rate cannot be determined as required by item (1) and (2) above.
Since the pipe break issue was resolved without reliance on leakage detection provisions, there was no need to determine leakage detection sensitivity as described in (3) above.
The staff's SER on pipe breaks, dated June 29, 1983, elaborates on this matter.
3.2 Topic V-12.A, Water Chemistry Limits (NUREG-0824 Section 4.19) 10 CFR 50 (GDC 14) as implemented by Regulatory Guide 1.56, requires that the reactor coolant pressure boundary have minimal probability of rapidly propagating failure.
This includes corrosion induced failures from impurities in the reactor coolant system.
IPSAR Section 4.19.1 requires that the licensee revise their Technical Specifica-tions for chlorides and conductivity to be consistent with Regulatory Guide
'1.56 or provide justification for not doing so.
3-1
The licensee proposed Technical Specification changes by letter dated December. 28, 1983.
The staff approved the Technical Specifications by letter dated June 21, 1984, which transmitted Amendment 99 to the license.
The staff concluded that the Technical Specifications regarding reactor water conductivity and chloride concentration meet Regulatory Guide 1.56 limits, for the most part, and that the minor differences meet the intent of the guide.
Therefore, the modified Technical Specifications will provide adequate protection for the primary coolant pressure boundary.
3.3 Topic VIII-3. A, Station Battery Test Requirements (NUREG-0824 Section 4.29) 10 CFR 50 (GDC 18), as implemented by Regulatory Guide 1.129, requires periodic testing to determine battery capacity and to demonstrate that the batteries will provide sufficient power under accident conditions.
The Millstone 1 battery surveillance requirements are included in Section 4.9.8 of the station Technical Specifications.
These specifications require a battery discharge test at each refueling outage or at least every 18 months.
The current licensing requirement for a discharge test is once every 60 months; however, there is no battery service test in the station Technical Specifications.
(Current requirements are for a battery service test every 18 months).
The staff proposed in IPSAR Section 4.29 that the testing of the batteries be in accordance with IEEE Std. 450-1975, IEEE Std. 308-1974, and the " Standard Technical Specifications for General Electric Boiling Water Reactors" (NUREG-0123).
The associated tests are as follows:
(1) At least once every 18 months, during shutdown, a battery service test should be performed to verify that the battery capacity is adequate to supply and maintain in operable status all of the actual emergency loads for two hours.
(2) At least once every 60 months, during shutdown, a battery discharge test should be performed to verify that the battery capacity is at least 80%
of the manufacturer's rating.
The licensee proposed these changes by letter dated December 13, 1983, and the staff approved the changes in Amendment 99 to the license, transmitted to the licensee by letter dated June 21, 1984.
3.4 Topic VIII-3.B, DC Power System Bus Voltage Monitoring and Annunciation (NUREG-0824 Section 4.30) 10 CFR 50.55a(h), through IEEE Std. 279-1971, and 10 CFR 50 (GDC 2, 4, 5, 17, 18, and 19), as implemented by SRP Section 8.3.2, Regulatory Guides 1.6, 1.32, 1.47, 1.75, 1.118, and 1.129, and BTP ICSB 21, require that the control room operator be given timely indication of the status of the station batteries.
In IPSAR Section 4.30.2, the staff noted that the PRA results from the Interim Reliability Evaluation Program (IREP) indicated that the current outage times allowed by Technical Specifications for the station batteries (7 days) are too long. The staff therefore, requested the licensee revise the TS or justify the existing limits.
The licensee is preparing a Technical Specification change to propose a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> outage limit.
3-2 1.
i 3.5 Topic XV-16, Radiological Consequences of Failure of Small Lines Carrying Primary Coolant Outside Containment (NUREG-0824 Section 4.35) 10 CFR 100, as implemented by SRP Section 15.6.2, requires that the radiological consequences of failure of small lines carrying primary coolant outside containment be limited to small fractions of the exposure guidelines of 10 CFR 100.
During the topic review, the staff determined that Millstone Unit 1 did not comply with current licensing criteria.
Based on the existing Technical Specifi-cation limits for primary coolant activity, the potential offsite doses would substantially exceed the applicable dose limits.
Therefore, the staff position in the IPSAR was that reactor coolant activity limits should be maintained within the limits imposed on new operating reactors; that is, within the limits of the Standard Technical Specifications for General Electric Boiling Water Reactors (NUREG-0123). This is necessary to limit plant operation with potentially significant amounts' of failed fuel so that the radiological consequences of events that do not damage fuel but do involve a release of reactor coolant to the environment will be low.
The licensee responded by letter dated December 28, 1983, which proposed to incorporate the primary activity limits given in the GE Standard Technical Specifications into the Millstone 1 Technical Specifications.
Although the proposed action statement associated with these limits do not conform exactly with those in the Standard Technical Specifications, the staff concluded in a letter dated June 21, 1984, that they accomplish the intended objective and, therefore, approved the licensee's proposed Technical Specifications in Amendment 99 to the license.
3.6 Topic XV-18, Radiological Consequences of a Main Steam Line Failure Outside Containment 10 CFR 100, as implemented by SRP Section 15.6.4, requires that the radiological consequences of failure of a main steam line outside containment be limited to small fractions of the exposure guidelines of 10 CFR 100. On the basis of an independent assessment of the radiological consequences of a main steam line failure outside containment, the staff determined that Millstone Unit 1 did not meet the current acceptance criteria for this topic.
If the existing Technical Specification Ifmits for primary coolant activity were used, the potential offsite doses would substantially exceed the applicable dose limits.
The staff's evaluation showed that small-line failure is more limiting than the main. steam line failure.
Therefore, the conclusions presented in Section 3.5 apply to this section as well.
i 3-3
1 TABLE 3.1 MODIFICATIONS TO MILLSTONE 1 TECHNICAL SPECIFICATIONS AS A RESULT OF SEP Supplement IPSAR-Amendment Section Section Title No.
3.1 4.16 Reactor Coolant Pressure Boundary 97 Leakage Detection 3.2 4.19 Water Chemistry Limits 99 3.3 4.29 Station Battery Test Requirements 99 3.4 4.30 Battery Outage Limits (Pending)
None 3.5 4.35 Radiological Consequences of Failure 99 of Small Lines Carrying Primary Coolant Outside Containment 3.6 4.36 Radiological Consequences of a Main 99 Steam Line Failure Outside Containment G
i f
3-4 i
't 4 IPSAR TOPIC RESOLUTIONS CONFIRMED BY NRC REGION I 0FFICE During the Integrated Assessment for Millstone 1, a number of issues were resolved by commitments made by the licensee for specific plant modifications or procedural changes.
Subsequent to issuance of the IPSAR for Millstone 1, Region I was requested through Task Interface Agreement 83-41, to verify that plant modifications had been implemented and to review changes to plant operating procedures made by the licensee.
Table 4.1 provides a list of IPSAR action where confirmation by Region I was requested.
Region I conducted onsite inspections for each item identified in Table 4.1.
The inspections consisted of examinations of installed equipment as well as a review of supporting procedures and other documentation.
Region I concluded that the licensee had met the commitments documented in the IPSAR for the items in Table 4.1 except for the discrepancies discussed below.
Each discrepancy was discussed with licensee management, who agreed to implement the necessary corrective action or to initiate additional correspondence if an alternative method of implementation was to be proposed.
Inspection findings with the results of this review are documented in Inspection Report 50-245/84-27.
b I
a 4-1
=
TABLE 4.1 ITEMS FOR CONFIRMATION BY NRC REGION I 0FFICE ITEM IPSAR NO.
DESCRIPTION REFERENCE 1.
Revise flood emergency procedures as described in 4.1.6 letter (Counsil (NNECO) to Crutchfield (NRC)), dated January 13, 1983.
These revisions are to include:
a)
Keeping the large flood gate on the gas turbine building closed.
b)
Initiation of flood protection actions at a flood level of 8.0 msl.
2.
Revise inspection program to include 4.6.2 inspection of floodwalls.
3.
Formalize inspection program for water 4.6.3 control structures.
4.
Review and implement generic emergency procedures, 4.17 including steps to proceed to a cold shutdown from outside the control room.
5.
Install an independent pressure interlock for RWCU 4.18 system isolation.
6.
Install administrative 1y controlled mechanical 4.20.1 locking devices on specific valves.
7.
On specified lines, install a second isolation 4.20.2 valve and administratively controlled locking devices on both valves.
8.
Revise operating procedures to include APRM and 4.24.1 IRM channel checks every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
9.
Provide isolation between the reactor 4.25.2 protection system and its power supply.
10.
Procedures to protect Class IE systems from a 4.27 degraded grid voltage condition under non-accident conditions.
11.
Bypass light off speed and generator excitation 4.28.1 speed trips under accident conditions.
4-2
i ITEM IPSAR N0.
DESCRIPTION REFERENCE 12.
Bypass high lube oil temperature trip under 4.28.2 j
accident conditions.
13.
Improve preventive maintenance program for 4.28.3 gas turbine generator.
14.
Bypass specified generator trips under accident 4.28.4 conditions.
4-3
Discrepancies identified by Region I are as follows:
4.1 Install administrative 1y controlled mechanical locking devices on manual containment isolation valves (NUREG-0824).
Four valves from the listing of nineteen in IPSAR Section 4.20.1 were found to be without mechanical locking devices.
The licensee has taken action to lock those valves, none of which were mispositioned.
The licensee had revised system valve alignment procedures to include the requirements for locking closed the valves listed in this section of the IPSAR. However, three exceptions were found. These valves are associated with piping at containment penetrations X-2108 and X-211A and had not been identified by station equipment number in the IPSAR.
This may have contributed to the lack of locking devices for these valves, which were in the proper position.
The fourth valve, 1-CS-35B, was included on lists of valves to be locked, and was in the closed position but the inspector found the valve unlocked.
A subsequent investigation by the licensee found that an incorrect valve had been locked.
The locking of the wrong valve (which was properly positioned) presented no safety problem because both the locked and unlocked valves are core spray test line manual isolation valves in series with each other.
Based on these findings, plant equipment and system drawings were reviewed by the inspector for additional valves which would meet the criteria of IPSAR Sections 4.20.1 and 4.20.2.
The inspector identified six additional valves associated with the low pressure coolant injection (LPCI) and Core Spray systems.
The licensee committed to install locks on these valves and to include them in the administrative procedures for locked valves.
They are:
Containment Valve Penetration Vents on Torus Test Line 3/4" CC-25; 1-LP-77A X-211A 1-LP-78A X-211A Drains on Containment Spray Line; 1-LP-698 X-398 1-LP-70B X-39B Core Spray Line Local Leak Rate Test Connections; 1-CS-38A X-16A 1-CS-38B X-16B I
4-4
.f -
Because of these findings, the licensee performed a full verification of all locked valves for both SEP commitments and for 10 CFR 50 Appendix J requirements and found no additional discrepancies.
This item will be followed up under Region I Outstanding Item No. 245/84-27-02.
4.2 Install a second isolation valve on lines with one manual containment i
isolation valve and lock closed both manual valves (NUREG-0824 Section 4.20.2).
4 In IPSAR Section 4.20.2 the staff identified three lines for modification.
None have been modified to include a second valve; no valves were locked, and two of the valves, the torus drain and the LPCI/CS pump suction ring header drain valves, did not appear in valve alignment procedures.
The third valve, 1-LP-67A, was not required to be locked closed.
All these
. valves were in the proper position.
The licensee committed to lock closed all three valves and to revise the applicable procedures.
)
Each line was found with a pipe cap or blank flange in addition to the one manual valve.
The torus-drain'line is equipped with a blank flange.
Local. leak rate testing is performed at that fitting by pressurizing between double "0" rings with the blank flange installed.
l, The licensee committed to review this finding and to update their previous commitments, contained in a December 27, 1982 letter to the NRC, if they intend to use pipe caps or blanked flanges instead of installing an additional valve.
This item will be followed up under Region I Outstanding Item No. 245/84-27-03.
4.3 Integrated Safety Assessment Program (ISAP) f In relation to the ISAP, the licensee has identified associated plant modifications and has proposed deferring certain corrective actions l
beyond the original scheduled completion dates.
Acceptance of this was addressed in an NRC letter to the licensee dated April 5, 1984.
i 1
Licensee programs to develop integrated implementation schedules, which I
.would include these modifications, are presently outstanding with the NRC l
Office of Nuclear Reactor Regulation as ~part of ISAP.
These items are identified as follows:
j-Item.
IPSAR Ref.
Outstanding Item No.
ISAP Issue RWCU Interlock 4.18 245/84-27-05 1.04 EGTG Startup Trip 4.28.1 245/84-27-06 1.01 EGTG Operational Trip 4.28.2 245/84-27-07 1.01 Generator. Trips 4.28.4 245/84-27-08 1.01 i
i i
4-5 1
f 5 REFERENCES Code of Federal Regulations, Title 10, " Energy" (includes General Design 3"
Criteria) l Letter dated September 29, 1982 from Counsil (NNEco) to Crutchfield (NRC),
Subject:
SEP Topic III-4.8, Turbine Missiles.
--, September 29, 1982 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic III-6, Seismic Design Considerations, III-11, Component Integrity.
--, December 3,1982 from Counsil (NNECo).to Crutchfield (NRC),
Subject:
SEP Topic VI-7.A.3, ECCS Actuation System; SEP Topic IX-5, Ventilation Systems.
--, December 27, 1982 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP Integrated Assessment Supplemental Information.
--, February 23, 1983 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic VI-4, Containment Isolation System.
--, March 24, 1983 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic 111-5.B, Pipe Break Outside Containment.
--, March 24, 1983 from Counsil (NNEco) to Crutchfield (NRC),
Subject:
SEP i
Topic III-6, Seisnic Design Considerations.
I
--, April 12, 1983 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic VI-7.C.1, Independence of Redundant Onsite Power Systems.
4
--, April 15, 1983 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic III-5.A, High Energy Pipe Break Inside Containment.
--, April 18, 1983 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic IX-5, Ventilation Systems.
--, May 10, 1983 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic VIII-2, Onsite Emergency Power System.
--, June 13, 1983, from Counsil (NNECO) to Dircks (NRC),
Subject:
Integrated Assessment of Regulatory Requirements.
1
--, June 20, 1983 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic VIII-2, Onsite Emergency Power System.
--, June 22, 1983 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic III-4.8, Turbine Missiles.
--, June 22, 1983 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic VI-4, Containment Isolation System.
--, July 20, 1983 from Counsil. (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic III-2, Wind and Tornado Loadings.
--, September 14, 1983 from Council (NNEco) to Eisenhut (NRC),
Subject:
Integrated Safety Assessment Program.
5-1
3 2 <
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~
--, November 1,1963 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
ProposedTechnicalSpecificationCharygeLeakageDetection.
--, recember 2, 1983 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic III-4.A, Tornado Missiles.
--, December 6,1983 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
._ Topic VI-7.C.1, Independence of Redundant Onsite Power Systems, r-
, December 13,' 1983 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
Proposed Technical Specification Change Station Batteries.
i
--, December 28, 1983 from Counsil (NNECo) to Eisenhut (NRC),
Subject:
Integrated Assessment of Regulatory Requirements.
- , December 28, 1983 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
Proposed Technical Specification Change Primary Coolant Chemistry.
--i, January 23, 1984 froni Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP Topi; MI-6, Seismic Design Considerations.
~
)
--, January 31, 1984 from Counsii (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic VII-1. A, Isolation of Reacter Protection System From Non-Safety Systems, Including Qualifications of Isolation Devices.
[
--, February 2,1984 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP t
lo~pic II-3.P4 Flooding Potentia) and Protection Requirements, SEP Topic III-2, Wind and Tornado Loadings, SEP Topic III-3.A,-Effects of High Water Level on Structures, SEP Topic III-7.8, Design Ccdes, Design Criteria and Load Combinations.
--, March 16c1984 from Counsil (NNECo) to Crutchfield (NRC),
Subject:
SEP s.,
Topic II-3.8, Flooding Potential and Protection Requirements, SEP Topic II-4.F, Settlement ofs(oindations and Buried Equipment, SEP Topic III-2, Wind and Tornado Loadingc, SEP Topic III-3,A, Effects of High Water Level on Structures, j
and SEP Topic III-6, Seismic. Design Considerations.
--, May 17, 1984 from Counsfl (NNECo) to Crutchfield (NRC),
Subject:
SEP Topic III-6, Seismic Design Considerations.
x
.3
.s 4
--, June 12, 1984 from Coun,sil (NNEco) to Crutchfield (NRC),
Subject:
SEP Topic,VIII-2, Onsite Emergency Power System.
T, m
--, December 50984 from Chunsil (NNEco) to Zwolinski (NRC),
Subject:
+,
fopic III-4.B,9urbine Missiles.
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p i
--, M5rch 14,- 1985 from Counsil (NNECo) to"Zwolinski (NRC),
Subject:
SEP j
Topic III-3.C, Inservice Inspection of Water Control Structures.
m
-r, Ai.rli $7,1985 from Zwolinski (NRC) to Counsil (NNECo),
Subject:
t 8xemption fronCSibmittal~0aterfor Updated Final Safety Analysis Report (FSAR)
--, July 31', 1985 from Thompson (NRC) to Opska (NNECo),
Subject:
Integrated
.';3 Safety AssefsmentcProgram.
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--, August 13, 1985 from Opeka (NNECo) to Grimes (NRC),
Subject:
Integrated Safety Assessment Program.
--, September 16, 1985 from Zwolinski (NRC) to Opeka (NNECo),
Subject:
ISAP Issue 1.20, M0V Interlocks.
--, October 7,1985 from Zwolinski (NRC) to Opeka (NNECo),
Subject:
IPSAR Sections 4.1.1, 4.1.7, 4.5.1, 4.5.2 and 4.6.1.
--, November 25, 1985 from Grimes (NRC) to Opeka (NNECo),
Subject:
IPSAR Sections 4.4 Wind and Tornado Loadings and 4.7 Tornado Missiles.
U.S. Nuclear Regulatory Commission, NUREG-75/087, See NUREG-0800.
--, NUREG-0123,Rev. 3, " Standard Technical Specifications for General Electric Boiling Water Reactors," December 1980.
--, NUREG-0800 (formerly NUREG-75/087) " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants - LWR Edition,"
December 1975.
--, NUREG-0824, " Integrated Plant Safety Assessment Report (IPSAR), Millstone Nuclear Power Station, Unit 1, Final Report," February 1983.
--, NUREG/CR-0098 - Development of Criteria for Seismic Review of Selected Nuclear Power Plants by N. M. Newmark and W. J. Hall, June 1978.
--, Regulatory Guide (RG) 1.6, " Independence Between Redundant Standby (Onsite) Power Sources and Between Their Distribution Systems."
--, RG 1.11 " Instrument Lines Penetrating Primary Reactor Containment."
--, RG 1.22, " Periodic Testing of Protection System Actuation Functions."
--, RG 1.26, " Quality Group Classifications and Standards for Water, Steam,
and Radioactive-Waste-Containing Components of Nuclear Power Plants."
--, RG 1.32, " Criteria for Safety-Related Electric Power Systems for Nuclear Power Plants."
--, RG 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems."
--, RG 1.47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems."
--, RG 1.56, " Maintenance of Water Purity in Boiling Water Reactors."
--, RG 1.59, " Design Basis Floods for Nuclear Power Plants."
--, RG 1.75, " Physical Independence of Electric Systems."
--, RG 1.76, " Design Basis Tornado for Nuclear Power Plants."
--, RG 1.106, " Thermal Overload Protection for Electric Motors on Motor-Operated Valves."
5-3
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--, RG 1.115, " Protection Against Low-Trajectory Turbine Missiles."
--, RG 1.117, " Tornado Design Classification."
--, RG 1.118, " Periodic Testing of Electric Power and Protection Systems."
, -, RG 1.129, " Maintenance, Testing and Replacement of Large Lead Storage Batteries for Nuclear Power Plants."-
i RG 1.132, " Site Investigations for' Foundations of Nuclear Power Plants."
--, RG 1.141, " Containment Isolation Provisions for Fluid Systems."
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APPENDIX A REFERENCES TO STAFF SERS FOR EACH TOPIC EVALUATED IN THE SUPPLEMENT
IPSAR Supplement No.
Date Reference 2.1 October 7, 1985 Letter from Zwolinski (NRC) to Opeka (NNECo)
Subject:
IPSAR Sections 4.1.1, Flooding Elevation, 4.1.7, Roofs; 4.5.1, Flood Elevation, 4.5.2, Groundwater; 4.6.1, l.
Deficiencies Noted During Site Visit.
2.
November 25, 1985 Letter from Grimes (NRC) to Opeka (NNECo)
Subject:
IPSAR Sections 4.4 Wind end Tornado Loading and 4.7 Tornado Missiles.
2.7 June 29, 1983 Letter from Shea (NRC) to Counsil (NNECo) 2.8
Subject:
IPSAR Sections 4.9, Effects of Pipe Break On Structures, Systems and Components Inside Containment; 4.10.2, Jet Impingement Criteria - Outside Contain-ment; and 4.16.1, RCPB Leakage Detection Operability Requirements.
2.11 September 16, 1985 Letter from Zwolinski (NRC) to Opeka (NNECo)
Subject:
ISAP Issue 1.20, M0V Interlocks.
2.12 July 7, 1983 Letter from Shea (NRC) to Counsil (NNECo),
Subject:
IPSAR Section 4.20, Containment Isolation.
2.13 July 5, 1983 Letter from Shea (NRC) to Counsil (NNECo),
Subject:
IPSAR Section 4.32, Ventilation Systems.
2.16 August 16, 1984 Letter from Shea (NRC) to Counsil (NNECo),
Subject:
IPSAR Section 4.28.3,. Gas Turbine Preventive Maintenance Program.
2.18 July 5, 1983 Letter from Shea (NRC) to Counsil (NNECo),
Subject:
IPSAR Section 4.32, Ventilation Systems.
3.1 May 23, 1984 Letter from Crutchfield (NRC) to Counsil (NNECo),
Subject:
Redefinition of the Word " Operable" - Coolant Leakage Monitoring.
3.2 June 21, 1984 Letter.from Crutchfield (NRC) to Counsil
-(NNECo), -
Subject:
Technical Specification Changes to Reactor Coolant Chemistry Limits and Station Battery Testing Requirements.
3.3 June 21, 1984 See Reference for 3.2.
3.5 June 21,.1984 See Reference for 3.2.
3.6 June 21, 1984 See Reference for 3.2.
A-1
APPENDIX 8 NRC STAFF CONTRIBUTORS AND CONSULTANTS
This Safety Evaluation Report Supplement is a product of the NRC staff and its consultants.
The NRC staff members listed below were principal contributors to this report.
NRC Staff Title Branch i
C. Grimes Branch Chief Systematic Evaluation Program M. Boyle Integrated Assessment Project Manager Systematic Evaluation Program S. Brown Reactor Systems Engineer Technical Specifications Review Group D. Chery Section Leader Environmental and Hydrologic Engineering Branch T. Michaels Sr. Integrated Assessment Project Systematic Evaluation Program Manager E. McKenna Sr. Project Manager Systematic Evaluation Program D. Persinko Maintenance and Surveillance Engineer Licensee Qualifications T. Cheng Sr. Structural Engineer Systematic Evaluation Program P. Chen Sr. Mechanical Engineer Systematic Evaluation Program J. Shea Sr. Project Manager Operating Reactors Branch #5 1
R. Scholl Sr. Reactor Systems Engineer Operating Reactors Assessment Branch J. Chen Geotechnical Engineer Structural and Geotechnical Engineering Branch Y. Li Mechanical Engineer Mechanical Engineering Branch Consultants Name Affiliation T. Stilwell Franklin Research Center D. Barrett Franklin Research Center B-1
O y
U.S. NUCLE AR REGUL ATORY COMMISSION BIBLIOGRAPHIC DATA SHEET NUREG-0824, Supplement No.1 4 TlTLE AND SUBTITLE (Add Volume Na, *rmermvaref 2.rtrasc ote ns Int: grated Plant Safety Assessment Report, Systematic Evalua-tion Program Millstone Nuclear Power Station, Unit 1 3 RECIPIENT S ACCESSION NO.
Northeast Nuc ar Energy Company - Docket No.: 50-245 j
- 7. AUTHOR (S)
- 5. DATE REPORT [MPLETED MONTH
/
l YEAR Novemberf 1985
- 9. PERFORMING ORGANit TION N AME AND MAILING ADDHESS IInclude lip Codel DATE REP [H f ISSUED Division *of Licen 'ng uONTw lvEAR Office of Nuclear actor Regulation Nov er 1985 U.S. Nuclear Regula ry Commission s (te,,
u,nai Washington, D.C.
- 20. 5 a (ife si ns:
12 SPONSORING ORGANIZATION N AE AND MAILING ADDRESS (factuae Zip Codel 77 g
g Same as #9 above bl. CONTRACT NO-13 TYPE OF REPORT PE RIOD C E RE D (Inctus.ve dates)
(Final Report)
Technical Evaluation
- 15. SUPPLEMENTARY NOTES 1
/
14 (Leave o/ anal Pertains to Docket No. 50-245
\\
/
- 16. ABSTR ACT 200 words or less/
The Nuclear Regulatory Comission (NR has publish d its Supplement No. I to the Final Integrated Plant Safety Assessment Repo (IPSAR)
UREG-0824) under the scope of the Systematic Evaluation Program (SEP), for ortheas Nuclear Eilergy Company's Millstone 1 Plant located in Waterford, New London Co ty, C necticut.
The SEP was initiated by the NRC to review the design of older oper ion luclear power plants to reconfirm and document their safety. This report documen e review completed under the SEP for those issues that required refined engineerin evaluations or the continuation of ongoing evaluations after the Final IPSAR for the Mi' one 1 Plant was issued. The review has provided for (1) an assessment of the signi ca of differences between current technical positions on selected safer.y issues and th e th existed when the Millstone 1 Plant was licensed, (2) a basis for deciding on how hese di erences should be resolved in an integrated plant review, and (3) a docu ted evalu ion of plant safety when the supplcment to the Final IPSAR and the S ety Evaluat-Report for converting the license from a provisional to a full-term lice e have been i ed.
The Final IPSAR and its supplements will form part of the base for considerins he conversion of the license.
- 17. KEY WORDS AND DOCUMENT AN ALYSIS 17a DESCRIP RS Systematic Evaluation Program 17b. IDENTIFIE RS'OPEN ENDE D TERMS 18 AVAILI.BILITY STATEMENT 19 SECURITY CLA,SS (Th,s report!
21 NO. OF PAGE S Unlimited Unclassified 20 SECURITY CLASS (Te,s paget 22 PRICE Unclassified s
NEC FORM 335 47 77)
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