ML20141D899
| ML20141D899 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 12/31/1985 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-1048, NUREG-1048-S04, NUREG-1048-S4, NUDOCS 8601070492 | |
| Download: ML20141D899 (53) | |
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NUREG-1048 Supplement No. 4 l
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- Safety Evaluation Report 1 related to the operation of Hope Creek Generating Station
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Docket No. 50-354 Public Service Electric and Gas Company Atlantic City Electric Company i
U.S. Nuclear Regulatory Commission L ] Office of Nuclear Reactor Regulation b -
1 December 1985 l i q
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1 NOTICE I Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:
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I
i ABSTRACT Supplement No. 4 to the Safety Evaluation Report on the application filed by Public Service Electric and Gas Company on its own behalf as co-owner and as agent for the other co-owner, the Atlantic' City Electric Company, for a license to operate Hope Creek Generating Station has been prepared by the Office of Nu-clear Reactor Regulation of the U.S. Nuclear Regulatory Commission. The facil-ity is located in Lower A110 ways Creek Township in Salem County, New Jersey.
This supplement reports the status of certain items that had not been resolved at the time of publication of the Safety Evaluation Report.
l Hope Creek SSER 4 iii
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TABLE OF CONTENTS P_ age ABSTRACT ............................................................. iii 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT.................... 1-1 1.1 Introduction................................................
1.7 Outstanding 1-1 Issues.......................................... 1-2 1.8 Co n f i rma to ry I s s ue s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.9 License Condition Items..................................... 1-2 2 SITE CHARACTERISTICS............................................. 2-1 2.2 Nearby Industrial, Transportation, and Military s
Facilities.................................................. 2-1 2.2.2 River Transportation................................. 2-1 3 . DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, EQUIPMENT, AND COMP 0NENTS....................................................... 3-1 3.4 Water Level (Flood) Design.................................. 3-1 3.4.1 Flood Protection..................................... 3-1 3.9 Mechanical Systems and Components........................... 3-5 3.9.3 ASME Code, Class 1, 2, and 3 Components, Component Supports, and Core Support Structures................ . 3-5 3.9.3.1 Loading Combinations, Design Transients, and Stress Limits........................... 3-5 3.9.6 Inservice Testing of-Pumps and Valves................ 3-6 3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment......................... 3-7 3.10.1 Seismic and Dynamic Qualification.................... 3-7 3.10.1.1 Introduction............................... 3-7 3.10.1~2. Discussion................................. 3-7 3.10.1.3 Generic Item............................... 3-8 3.10.1.4 Equipment-Specific Item.................... 3-8 c 3.10.1.5 Confi rmato ry Is s ues. . . . . . . . . . . . . . . . . . . . . . . . 3-8 3.10.1. 6 S umma ry. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-9 3.10.3 TMI' Action Plan Item II.K.3.28....................... 3-10 Hope Creek SSER 4 v l
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1 TABLE OF CONTENTS (Continued)
Page 7 INSTRUMENTATION AND CONTR0LS..................................... 7-1 7.2 Reactor Protection (Trip) System ........................... 7-1 7.2.2 Specific Findings.................................... 7-1 7.2.2.3 Testability of Plant Protection Systems at Power............................ 7-1 7.2.2.5 Instrumentation Setpoints................... 7-2 7.5 Safety-Related Display Instrumentation...................... 7-4 7.5.2 Specific Findings.................................... 7-4 7.5.2.4 Bypassed and Inoperable Status Indication.................................. 7-4 7.6 Interlock Systems Important to Safety....................... 7-5 7.6.2 Specific Findings.................................... 7-5 7.6.2.1 Isolation of Low-Pressure System From the High-Pressure Reactor Coolant System.............................. 7-5 11 RADI0 ACTIVE WASTE MANAGEMENT..................................... 11-1 11.4 Solid Waste Management System.............................. 11-1 14 INITIAL TEST PR0 GRAM............................................. 14-1 14.2 Initial Plant Test Program - Final Safety Analysis Report............................................ 14-1 APPENDICES APPENDIX A CONTINUATION OF CHRONOLOGY APPENDIX B BIBLIOGRAPHY ePPENDIX D ACRONYMS AND INITIALISMS APPENDIX E PRINCIPAL STAFF CONTRIBUTORS AND CONSULTANTS Hope Creek SSER 4 vi
LIST OF TABLES Pa!Le 1.1x Outstanding Issues (Revised Table 1.1 From Supplement No. 3)..... 1-3 1.2 Confirmatory Issues (Revised Table 1.2 From Supplement No. 3).... 1-5 1.3 License Conditions (Revised Table 1.3 From Supplement No. 3)..... 1-7 3.1 Water Depth and Duration of Flooding at Plant Grade and Windspeed and Direction at Hope Creek Site....................... 3-13 3.2 Potential Number of Floating Missiles at Hope Creek Site......... 3-13 3.3 Equipment Audited................................................ 3-14 i
i Hope Creek SSER 4 vii l
1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 Introduction In October 1984, the U.S. Nu: lear Regulatory Commission (NRC) staff issued its Safety Evaluation Report (SER) (NUREG-1048) on the application filed by Public Service Electric and Gas Cer.vany (PSE&G) (applicant) on its own behalf as co-owner and as agent for the other co-owner, the Atlantic City Electric Company, for a license to cperate the Hope Creek Generating Station (Docket No. 50-354).
At that time, th staff identified items that were not yet resolved with the applicant. Supplement Nos. 1, 2, and 3 to the SER were issued in March 1985, August 1985, and October 1985, respectively. The purpose of this supplement to the SER is to provide the staff evaluation of open items that have been re-solved and to report on the status of all open items.
During its 296th meeting on December 13-15, 1984, the Advisory Committee on Reactor Safeguards reviewed the operating license application filed by the appli-cant. The Committee, in a December 18, 1984, letter from Chairman Jesse C.
Ebersole_to NRC Chairman Nunzio J. Palladino (reproduced-as Appendix H in Supple-ment No.1), concluded that subject to the resolution of open items identified by the staff in the SER and the items noted in the above referenced letter and satisfactory completion of construction, staffing, and preoperational testing, there is reasonable assurance that Hope Creek can be operated at power levels up to 3,293 megawatts-thermal (100% power) without undue' risk to the health and safety of the public.
Each of the following sections or appendices of this SER supplement is numbered the same as the corresponding SER section or appendix that is being updated.
Appendix A is a continuation of the chronology of the staff's actions related to the processing of the Hope Creek application and lists letters between the NRC staff and the applicant in chronological order. Appendix B is a list of references cited in this report.* Appendix 0 is a list of acronyms used herein.
Appendix E identifies principal contributors to this SER supplement.
Copies of this SER supplement are available for inspection at the'NRC Public Document Room at 1717 H Street, N.W., Washington, D.C., and at the Pennsville Public Library,190 South Broadway, Pennsville, New Jersey. They are also available for purchase from the sources indicated on the inside front cover of this report.
The NRC Project Manager assigned to the operating license application for Hope Creek is Mr. David H. Wagner. Mr. Wagner may be contacted by writing to Mr. David H. Wagner Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555
- Availability of all material cited is described on the inside front cover of this report.
Hope Creek SSER 4 1-1 l
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- 1. 7 Outstanding Issues The staff identified certain outstanding issues in the SER that had not been re-solved with the applicant. The status of these issues is listed in Table 1.1 1
and discussed further in the indicated sections of this report. If the staff review is completed for an issue, it is indicated as " closed." The staff will complete its review of outstanding issues before the operating license is issued.
1.8 Confirmatory Issues The staff identified confirmatory items in the SER that required additional in-formation to confirm preliminary conclusions. The status of these items is listed in Table 1.2 and discussed further in the indicated sections of this report. If the staff review is completed for an item, it is identified as " closed."
1.9 License Condition Items There are certain issues for which a license condition may be desirable to en-1 sure that' staff requirements are met by a specified date (Table 1.3). These conditions will be in the form of a condition in the body of the operating license.
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Hope Creek SSER 4 1-2
Table 1.1 Outstanding' issues (revised Table 1.1 from Supplement No. 3)
Issue Status
- SER section(s)
(1) Riverborne missiles Closed 3.4.1 1 (2) Equipment qualification Partial closure 3.10 (also 3.10, Supplements 2 and 3)
(3) Preservice inspection program Confirmatory 5.2.4,'6.6 (Supplement 3)
(O GDC 51 compliance Closed 6.2.7 ~
(Supplement 2)
(5) Solid-state logic modules Under review
, (6) Postaccident monitoring Closed 7.5.2.3 i instrumentation (Supplement 2)
(7) Minimum separation'between Closed 8.3.3.3.3 non-Class 1E conduit and (Supplement 3)
Class 1E cable trays (8) Control of heavy loads Closed 9.1.5 (Supplement 1)
! (9) Alternate and safe shutdown Partial closure 9.5.1.4 (Supplement 2)
(10) Delivery of diesel generator Closed 9.5.4.2 fuel. oil and lube oil (Supplement 1)
(11) Filling of key management Under review positwns (12) Training program items (a) Initial training programs Closed 13.2.1.1 (Supplement 2)-
(b) Requalification training ~ Closed 13.2.1.2 programs (Supplement 2)
(c) Replacement training -C1osed 13.2.1.3 l programs (Supplement 2) i
- (d) .TMI issues I.A.2.1, I.A.3.1,. Closed 13.2.1.4 and-II.B.4 (Supplement 2) 1 (e) Nonlicensed training Closed ~ 13. 2. 2
, programs. (Supplement 2)
Hope Creek SSER 4 1-3
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Table 1.1 (Continued)
Issue Status SER section(s)
J (13) Emergency dose assessment Under review computer model (14) Procedures generation package Awaiting information (15) Human factors engineering Awaiting information
[ Hope Creek SSER 4 1-4
Table 1.2 Confirmatory issues (revised Table 1.2 from Supplement No. 3)
Issue Status SER section(s)
(1) Feedwater isolation check valve Closed 3.6.2 analysis (Supplement 3)
(2) Plant unique analysis report Closed 3.9.3.1, 6.2.1.7 (also Supple-rent 3)
(3) Inservice testing of pumps and Closed 3.9.6 valves ,
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(4) Fuel assembly accelerations Closed 4.2 (Supplement 2)
(5) Fuel assembly liftoff Closed 4.2 (Supplement 2)
(6) Review of stress report Closed 5.2.1.1 (Supplement 3)
(7) Use of Code cases Closed 5.2.1.2 (Supplement 2)
(8) Reactor vessel studs and Closed 5.3.1.5 fasteners (Supplement 3)
(9) Containment depressurization Under review
, analysis (10) Reactor pressure vessel shield Closed 6.2.1.5 annulus analysis (Supplement 3)
(11) Drywell head region pressure Closed 6.2.1.5 response analysis (Supplement 3)
(12) Drywell-to-wetwell vacuum Closed 6. 2.1. 7 breaker loads (Supplement 3)
(13) Short-term feedwater system Closed 6.2.3 analysis (Supplement 3)
(14) Loss-of-coolant-accident Closed 6.3.5, 15.9.3 analysis (Supplement 2)
(15) Balance-of plant testability Closed 7.2.2.3 analysis (16) Instrumentation setpoints Closed 7.2.2.5 (17) Isolation devices Under review Hope Creek SSER 4 1-5
Table 1.2 (Continued)
Issue Status SER section(s)
(18) Regulatory Guide 1.75 Under review (19) Reactor mode switch Closed 7.2.2.9 (Supplement 3)
(20) Engineered safety features Under review reset controls (21) High pressure coolant injection Closed 7.3.2.9 initiation (Supplement 3)
(22) IE Bulletin 79-27 Closed 7.4.2.1 (Supplement 3)
(23) Bypassed and inoperable status Closed 7.5.2.4 indication (24) Logic for high pressure coolant Closed 7.6.2.1 injection interlock circuitry (25) End-of-cycle recirculation pump Closed 7.6.2.4 trip (Supplement 3)
(26) Multiple control system failures Closed 7.7.2.1 (Supplement 3)
(27) Relief function.of safety / relief Closed 7.7.2.2 valves (Supplement 3)
(28) Main steam tunnel flooding Closed 8.3.3.1.4 analysis (Supplement 3)
(29) Cable tray separation testing Closed 8.3.3.3.2 (Supplement 3)
(30) Use of inverter as isolation Closed 8.3.3.3.4 device (Supplement 3)
(31) Core damage estimate procedure Closed 9.3.2 (Supplement 3)
(32) Continuous airborne particulate Closed 12.3.4.2 monitors (Supplement 3)
(33) Qualifications of senior Closed 12.5.1 radiation protection engineer (Supplement 2)
(34) Onsite instrument information Closed 12.5.2 (Supplement 3)
Hope Creek SSER 4 1-6
t Table 1.2 (Continued)
Issue Status SER section(s)
(35) Airborne iodine concentration ~ Closed 12.5.2 instruments (Supplement 3)
(36) Emergency Plan items Under. review (37) TMI Item II.K.3.18 Closed 15.9.3 (Supplement 2)
Table 1.3 License _ conditions (revised Table 1.3 from Supplement No. 3)
License condition Status SER section (1) Turbine system maintenance program
, (2) NUREG-0803 implementation (3) Inservice inspection
. (4) Postaccident sampling system Removed 9.3.2 l (Supplement 3)
(5) Solid waste process control Revised 11.4 program (6) Partial feedwater heating (7) Cask drop accident Removed, 15.7.5 (Supplement 3)
Hope Creek SSER 4 1-7 a
2 SITE CHARACTERISTICS 2.2 Nearby Industrial, Transportation, and Military Facilities 2.2.2 River Transportation At the construction permit review stage, the applicant indicated that the size of ships that could conceivably ram into the seismic Category I intake struc-ture was limited by tidal ccnditions. Under normal tidal water elevations, ships weighing more than 8,600 tons would ground on the shallow water shoal areas out-side the river channels before they could reach the intake structure. The kinetic energy levels associated with a postulated ramming of the.. intake.struc-ture would be of the same order of magnitude as those from past major ship col-lisions. From historical data on ship collisions, it was determined, in this case, that the expected structural damage would be to the impacting ship and not to the massive concrete intake structure.
At the operating license review stage,-the applicant studied the most recent river bathymetry and National Oceanic and Atmospheric Administration charts and determined that a smaller 2,000-ton displacement ship with a draft of 12 ft would be grounded 400 ft before it reached the intake structure. The applicant reconfirmed that the major damage would occur to the impacting vessel rather than to the intake structure. The staff concurs with the applicant's analysis.
On the basi's of past historical collision data on the Delaware River, the staff estimates that the probability of a ship or barge impacting the intake struc-ture during normal water levels is less than 10 7 per year.
The staff's evaluation of waterborne missiles on the intake structure during storm or flood conditions is given in Section 3.4.1 of this supplement.
Hope Creek SSER 4 2-1
l 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, EQUIPMENT, AND COMPONENTS 3.4 Water Level (Flood) Design 3.4.1 Flood Protection In the SER, the staff stated that the applicant had not addressed its concerns about the structural integrity of the safety-related structures, the design of the service water pump seals.during the design-basis flood, and the effects of floating missiles. During hurricanes, the water level at the site will rise to the power block structures and can exceed the top of the service water intake structure. The staff expressed concern t. hat the high water level could transport floating missiles onto the site and impact the power block structures resulting in the loss of watertight integrity. The resulting inleakage of water could cause flooding of safety-related components or systems within the structures.
.In response to staff concerns, the applicant submitted additional information on July 27 and September 17, 1984, and January 31, February 22, May 8, and September 16, 1985. In these submittals, the applicant used probabilities of occurrence to demonstrate that the probability of floating missiles causing significant damage to the plant is insignificant and need not be considered.
Additionally, the staff performed an independent evaluation of the design-basis flood resulting from the probable maximum hurricane (PMH). Its analysis indi-cates that the PMH generates water 12.3 ft deep at the power block. Hurricanes of lesser force than the PMH would produce less water at the power block and, therefore, are bounded by the analysis of the PMH. The applicant had previ-ously stated that neither the probable maximum flood nor the upstream dam failure imposes a flooding concern at plant grade around the power block.
The staff's. evaluation of the effect of the PMH on the Hope Creek site consid-ered the storm approaching the Atlantic coastline in the vicinity of the en-trance to the Delaware Bay along the most critical path. The hurricane winds were considered to be blowing witn increasing windspeeds normal to (across) the axis of the lower bay. When the eye of the hurricane makes landfall in the vicinity of Ocean City, Maryland, the cyclonic (counterclockwise) winds asso-ciated with the forward quadrant of the hurricane will have windspeeds of about 90 to 100 miles per hour and will blow in a southwesterly direction or perpen-dicular to the axis of. Delaware Bay. Waves in the bay will approach a maximum height of 15 ft (traveling in the same direction as the wind).
As the hurricane moves inland, the surge associated with the hurricane progres-ses up the bay. The combined effects of the tidal flood currents and the surge will produce upstream current velocities from 5 knots to a maximum of 15 knots (during the passing of the crest of the surge). Although the winds and asso-ciated waves provide a-small additional component to the current ulocity, the winds have a primary effect on the direction of travel of any riverborne object (missile) located in the bay area. The extent of this effect and the direction of travel of a riverborne object will depend on the object's above-wat'er surface area exposed to the hurricane winds and below-water surface area influenced by Hope Creek SSER 4 3-1
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the water current velocities. During the period of crosswinds and waves and the relatively low flood current velocities, riverborne missiles will be driven '
ashore onto the Delaware side of the bay. At this time, the wind direction at l- the plant site will be blowing from the plant toward the river. It is only as the peak or crest of the surge occurs at the site that the wind-will change direction to blow from the river toward the plant power block. i As stated above, during the early stages of the hurricane, the crosswinds will
! transport riverborne missiles on the bay onto the Delaware shore. Large run-i away riverborne missiles (barges and pleasure craft) will be impacted by both j
wind and wave (surf) action with the result that the more fragile floating craft l may' sink or break up. Those that maintain their watertightness or buoyancy may be grounded during the earlier stages of the storm but can be refloated as the
] crest of the surge passes. They can then be transported up the bay by the
. hurricane winds. It is only during the latter stages of the hurricane, as the >
1 surge at the site begins to recede, that the potential increases for riverborne '
missiles to impact-the power block. The water depth at plant grade, the dura-tion of flooding at plant grade, and the windspeed and direction at the site are summarized in Table 3.1. -
On the basis of its independent analysis, the staff concludes that the ~poten-tial for riverborne missiles being transported up the Delaware Bay during the PMH event onto the Hope Creek site and impacting safety-related structures ia extremely small. The potential is further limited by the combination of (1) the
} very short time during this event that the water depths at plant grade are suf-ficient to float large waterborne missiles and (2) the wind blowing from the appropriate direction to transport riverborne missiles toward the plant's <
safety-related facilities in the power block. Because the water depths adja-cent to the service wa_ter. intake structure are greater than those at plant grade during and following the PMH, the service water intake structure is the most vulnerable of the safety-rehted facilities.
Although the staff's analysis indicates that the potential for waterborne mis-siles being transported onto the Hope Creek-site and endangering the power block safety-related structures is acceptably low, the service water intake structure, l because of its location on the bank of the Delaware River, is exposed to river-l ' borne missiles. In evaluating the risk of-riverborne missiles striking the ser-i vice water intake structure, the staff considered the following scenarios:
i (1) riverborne missiles resulting from upstream dam failures (2) -riverborne missiles resulting from severe storms i With regard to riverborne missiles resulting from upstream dam failures, the
- staff evaluated the effects from failure of the Pepacton Reservoir Dam (Downs-ville), the Cannonsville Dam, and the F. E. Walter Dam. The applicant stated that "since the dams are sited in locations influenced by different tectonic sources, the seismically-induced failures of the. individual dams are considered
- .to be independent events..."; therefore, multiple dam failures were not evalu-l ated. The applicant performed an analysis that demonstrated that floating mis-siles, with an impact force less than a 52,000-lb boat traveling at 16 ft per second and impacting the intake structure over an impact area that is. assumed to be 10 in. in diameter, would not damage the traveling screen or pump seals.
Therefore, the potential' floating missiles, based on the applicant's site sur-veys, that were considered in the evaluation are given in Table 3.2.
- Hope Creek SSER 4 3-2 J
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By combining the probability of dam failure with the number of missiles and the probability of impacting the intake structure (120 ft/100 mi) given a missile, the staff can determine the probability of a riverborne missile damaging the intake structure as a result of a dam failure.
Downsville (0.5 x 10 5)(432)(2.3 x 10 4) = 496.8 x 10 9 impact / year Cannonsville (0.2 x 10 5)(432)(2.3 x 10 4) = 198.7 x 10 9 impact / year Walter (0.8 x 10 8)(432)(2.3 x 10 4) = 79.4 x 10 9 impact / year i
Total 774.9 x 10 " impact / year i
Therefore, the probability of riverborne missiles resulting from upstream dam failures damaging.the intake structure is 7.7 x 10 7 impact per year. This probability value is conservative.
In considering riverborne missiles resulting from severe storms, the staff evaluated the effects of a PMH, a model hurricane, and extreme wind events 4
(EWES). The probabilities of these storms were based on information provided by the applicant.
estimated to.be in TheprobabilityofaPMHstrikingthesiteinanyyearwas the range of 1 x 10 3 to 1 x 10 . This prcbability was multiplied by the probability that the PMH would approach the site along the most critical path as discussed earlier. The staff used a ratio of 5/39, which is based on those hurricanes that would have produced significant tidal surges in Delaware Bay of all the hurricanes occuring north of Cape Hatteras between 1683 and 1869.
By combining the probabilities of the storms with the number of self propelled vessels that are potential runaways (Table'3.2) and the probability of impacting the 120-ft intake structure given the vessel loses power and steering, the staff can determine the probability of self propelled vessels damaging the traveling screen or pump seals as a result of severe storms.
l PMH (1.3 x 10 4 to 1.3 x 10 5)(2)(2.1 x 10 8) = 5.5 x 10 10, 11 Model hurricane (1 x 10 2)(2)(2.1 x 10 s) = 4.2 x 10 8 EWE (2 x 10 3)(2)(2.1 x 10 8) = 8.4 x 10 9 Total 5.095 x 10 8 By using the same storm probabilities with the number of barges and the proba-i bility of a runaway barge striking the 120-ft intake structure, the staff can determine the probability ~of non-self propelled vessels damaging the traveling screen or pump seals as a result of severe storms.
4 PMH (1.3 x 10 4 to 1.3 x 10.s)(2)(1.2 x 10.s) = 3.1 x 10 10, 11 Model hurricane (1 x 10 2)(2)(1.2 x 10 s) = 2.4 x 10 8 EWE- (2 x 10 3)(2)(1.2 x 10 s) = 4.8 x 10 9 !
Total 2.911 x 10 8 i Therefore, the probability of riverborne missiles resulting from the severe storms damaging the intake structure is 8 x 10 s. impact per year. By combining this probability with the probability of~ damaging the intake structure as a re-sult of. upstream dam failures, the staff finds that the overall risk of river-borne missiles striking the service water intake structure is about 8.5 x 10 7 impact per year.
i ~ Hope Creek SSER 4 3-3 j
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1 In summary, the staff has considered the potential for riverborne missiles en-tering the Hope Creek site and impacting the power block during PMH conditions.
Its analysis of this event has led the staff to conclude that this potential is extremely small. Additionally, the staff analyzed the potential for river-L i
bo'rne missiles striking the service water intake structure and determined that the probability of damage from riverborne missiles is acceptably small.
In the SER, the staff also stated that the applicant had not provided the re-sults of an analysis that shows that all penetrations, including the submarine i doors and doors from safety-related structures to non-safety-related structures, will maintain their leaktightness against the static and dynamic effects of the :
probable maximum flood, including wave runup. The probable maximum flood is l caused by the PMH.-
By Amendment 8 to the Final Safety Analysis Report (FSAR),
the applicant ~ stated that all exterior doors in seismic Category I structures
! are designed to withstand the static and dynamic effects, including wave runup, from postulated flooding events. By Amendment 9, the applicant identified in Table 3.4-2 the various means of preventing flooding from non-safety-related structures to safety related structures. These methods include water and i pressure-tight doors, hatches, and manhole covers, waterstops, boots, and em-4 bedded collars.
l The only opening that is not watertight is the air intake for the service water
, pumps in the intake structure.
basic elevations that run the length The of roof of structure.
the: the intake structure is at three the intake structure, the roof elevation is 128 ft. During At the river side of overtops this part of the structure. The middle section is the PMH, the water lower, at elevation 122 ft. This section has the access hatches for the service water pumps and the air intakes for the intake structure ventilation. The water enters this section of the roof when the wave overtops the 128-ft-elevation section of roof or directly from the open-end of the building. The air intake consists, in part, of an inlet louver with a bottom elevation of 122 ft and with a concrete e missile barrier behind the louver. This barrier forms a labyrinyth to protect the ventilation system from tornado generated missiles. By Amendment 12, the applicant identified the top of this concrete barrier to be at elevation 128 ft 6 in. , which is higher than the river-side section of the roof. Therefore, the effects of the waves that would overtop the 120-ft section of the roof would be dampened by the intake louvers before hitting the tornado missile barrier.
' Thus, the water would be prevented from entering the ventilation. system. Any spray or mist that might be drawn into the ventilation system would be reduced by the intake control damper located in the inlet to the intake fan. The appli-cant considers that any remaining mist would not adversely affect the safety-l related equipment in the intake structure.
i Therefore, the staff concludes, on the basis of its previous safety evaluation 3
in the SER, its analysis of-the PMH, a probabilistic risk assessment, and its
' evaluation of the leaktightness capability, that conformance with the require-ments of General Design Criteria (GDC) 2 and 4 (Appendix A to Title 10, Part 50, of the Code of Federal Regulations (10 CFR 50)) has been demonstrated with regard to protection against natural phenomena and missiles and with the guide-lines of Positions ~ C.1 and C.2 of Regulatory Guide (RG) 1.59 and Position C.1 i
of RG 1.102, concerning design-basis floods and flood protection and is,
- therefore, acceptable. The design of the facility for flood protection meets i the intent of the acceptance criteria of Section 3.4.1 of NUREG-0800.
l Hope Creek SSER 4 3-4 l
- - - . . - -_. - .. . - = -
M r l
, 3.9 Mechanical Systems-and Components 1
1 3.9.3 ASME Code, Class 1, 2, and 3 Components, Component Supports, and Core 3
Support Structures
, 3.9.3.1 Loading Combinations, Design Transients, and Stress Limits i 'In the SER, the staff stated that the plant-unique analyses for the Hope Creek i boiling water reactor (BWR) Mark I safety / relief valve (SRV) and loss-of-coolant-accident (LOCA) hydrodynamic loads were being reviewed. The capability of the 3 BWR Mark I containment structures and piping systems to withstand the effect of
- hydrodynamic loads resulting from a LOCA and/or an SRV discharge was not consid-ered in the original design of the structures. The resolution of this issue was j divided into a short-term and a long-term program.
I I
On.the basis of the results of the short-term program, which verified that each Mark ~I containment would maintain its integrity and functional capability.when subjected to the loads induced by a design-basis LOCA, the NRC staff granted an j exemption relating to the structural safety requirements of 10 CFR 50.55(a). -
The study was reported in NUREG-0408,." Mark I Containment Short Term Program."
The objective of the long-term program was to maintain a margin of safety when the Mark I containment structures and piping systems are subjected to the additional hydrodynamic loads. The detailed guidance of the long-term program
, is contained in NUREG-0661, " Mark I Containment Long-Term Program," which de- ,
scrib.es the generic hydrodynamic load definition and structural acceptance cri-j teria consistent with the requirements of the applicable codes and scandards.
j To fulfill the objective of the long-term program, the applicant has completed
- all modifications on the containment and torus-attached piping. The adequacy of 4
these modifications was documented in the report prepared by the Nutech Engi-
! neers, Inc. (Nutech), BPC-01-300-1-through -6, " Hope Creek Generating Station -
} Plant Unique Analysis Report" (PUAR). Under contract to the NRC, the Brookhaven l National Laboratory (BNL) reviewed and approved the loads and loading combina-i tions part of the PUAR. These approved loads were used as input to the struc-f tural analyses. The Franklin Research Center (FRC), under contract to the NRC, j reviewed these analyses and determined their adequacy. :
I
- j. The Mark I long-term program of the Hope Creek plant was described in the PUAR 1 prepared by Nutech. This. report' describes the modifications performed on con- {
l
! tainment structures and torus-attached piping in the Hope Creek design. Areas !
! covered by the PUAR include the torus shell, external support system, vent q header system, internal structures, torus ittached piping, all SRV lines, and
~
i vent pipe penetrations. The materials, design, and fabrication requirements of ;
i i the modifications were in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Division 1,Section III, i with Addenda through Summer 1978. Inspections of modification were performed I in accordance with the requirements of Section XI_of the ASME Code with Addenda
- through Summer 1978. To determine the appropriate' Code-allowable service limits for the'specified loading combinations, the PUAR follawed the guidelines set by i NUREG-0661 and the General Electric (GE) report, NEDO-24583-1, " Mark I Contain-i ment Program Structural Acceptance Criteria' Plant Unique Analysis Application Guide." The portion of the Nutech report applicable to loadings and loading combinations was audited by BNL and found' acceptable. Results of this audit
- were discussed in Section 6.2.1.7 of Supplement No. 3 to the SER.
l Hope Creek SSER 4 3-5 i
--,..,.---,c -.,,--,, ~ n n, - - - , , . - _ . - - . ~ . - - . , . - ~ _ , _ - - - - . _ . , . n.,,~,,_n.,-.. -
M i
Using the properly determined loadings and loading combinations, Nutech used computer programs STARDYNE and PISTAR as major tools to perform the analyses.
STARDYNE and PISTAR are programs that have been used widely in the industry for similar purposes and have been approved by NRC staff. In addition, Nutech used computer program CMDOF. In response to a staff question on the verifica- l
.. tion of the CM00F program, the Mark I Owners Group submitted a comparison of l analytical results from the program with observed stress values on applicable piping systems from full-scale testing conducted at the Monticello Nuclear Gen-
} erating Plant (Musolf,1985). Because the analytic results were more conserva-l tive than the Monticello test results, the staff concluded that the use of the l CMDOF program for the Hope Creek piping is acceptable. However, because the staff has not performed a detailed evaluation of CM00F, the use of this program j to analyze piping systems that are different from those at Hope Creek will be ,
i i
evaluated on a case-by-case basis.
Results of the analyses were summarized to show that modifications are adequate under various loading combinations. The adequacy of the modified containment structures and torus-attached piping was audited by FRC. FRC developed audit procedures for all Mark I lo'ng-term program users, which are described in detail in FRC TER-C5506-308, " Audit Procedures for Mark I Containment Long-Term
! Program - Structural Analysis." The review performed by FRC has followed this
! document closely. Results and conclusions of this effort were reported in
-FRC TER-C5506-327, " Audit for Mark I Containment Long-Term Program Structural Analysis.for Operating Reactors - Public Service Electric and Gas Company."
The audit verified Nutech analyses by examining mathematical models and loading combination analyses used and summarized the results to see whether the modifi-1 cations met the required criteria. A check list was compiled to ensure the completeness of the auditing. The staff has reviewed the FRC report and con-curs with its conclusion that the modifications meet the Mark I Containment
- Long-Term Program objective. An augmented fatigue evaluation method for ASME i Code, Class 2/3 piping was developed by MPR Associates for GE (MPR-751). This l
-report was reviewed by the staff, and the conclusion that torus piping systems
! in all Mark I BWRs had a fatigue usage of less than 0.5 during the plant life is acceptable for Hope Creek.
i Except for the vacuum breakers, the modifications performed at Hope Creek fol-lowed the guidelines of NUREG-0661 and met the respective requirements of Sec-
- tions XI and III of the ASME Code. The applicant's analyses have been verified i
by the FRC audit and approved by the staff under the LOCA and SRV discharge loads. The staff considers SER Confirmatory Item 2 closed. The definition of l vacuum breaker loadings needs further clarification. A separate report address-
)
ing vacuum breakers will be prepared once the review has been completed.
3.9.6 Inservice Testing of Pumps and Valves j
The applicant had not submitted an inservice testing (IST) program for pumps
! and valves at the time the SER was issued. Thus, the SER stated that the IST l
program would be reviewed later and the results reported in a supplement to the SER. By a letter dated July 12,1985 (R. Mittl, . PSE&G, to W. Butler, NRC),
the applicant forwarded Revision 0 of the Hope Creek IST program.
f _
The staff has not completed a detailed review of the Hope Creek IST program. A preliminary review was completed, and it was found that it is impractical with-in the limitations of design, geometry, and accessibility for the applicant to l
l Hope Creek SSER 4 3-6
- _ _ ~ . _ . _ _ . _ _ _ ~ _ . _ _ . _ _ _ _ _ _ . . _ . _ _ . _ . _ _ _ _ _ . _
i meet certain of the ASME Code requirements. A delay in the imposition of those requirements will not endanger life or property or the commcn defense and secur-ity of the public. Such a delay is in the public interest giving due considera-tion to the burden on the applicant that could result if the requirements were imposed. On the basis of experience at similar plants where no significant ad- ,
verse health and safety effects were found, the staff concludes that the re- '
quirements of 10~CFR 50.55a(g)(6)1 are satisfied. If relief were not granted, !
the applicant might be forced to curtail operation of the plant, which consti-tutes a considerabl'e burden. Therefore, pursuant to 10 CFR 50.55a(g)(6)(i), the relief that the applicant has requested from certain of the pump and valve test-ing requirements of the 1980 Edition of ASME Code,Section XI, through Winter 1981 Addenda should be granted for a period of no longer than 2 years from the date of issuance of the operating license or until the detailed review has been completed, whichever comes first. If the review results in additional testing requirements, the applicant will be required to comply with them.
3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment 3.10.1 Seismic and Dynamic Qualification 3.10.1.1 Introduction Evaluation of the applicant's program for seismic and dynamic qualification of safety related electrical and mechanical equipment consists of (1) a determina-tion of the acceptability of the procedures used, standards followed, and the completeness of the program in general, and (2) an audit of selected equipment items to develop a basis for judging the completeness and adequacy of the imple-mentation of the entire seismic and dynamic qualification program.
Guidance for the evaluation is_provided by Section 3.10 of the Standard Review Plan (NUREG-0800) and its ancillary documents, RGs 1.100, 1.61, 1.89, and 1.92, NUREG-0484, and Institute of Electrical and Electronics Engineers (IEEE)
Stds 344-1975 and 323-1974. These documents define acceptable methodologies for the seismic qualification of equipment. Conformance with these criteria is required to satisfy the applicable portions of GDC 1, 2, 4,14, and 30 as well as Appendix B to 10 CFR 50 and Appendix A to 10 CFR 100. The program is evaluated by a Seismic Qualification Review Team (SQRT), which consists of staff engineers and engineers from the Idaho National Engineering Laboratory (INEL,EG&G).
3.10.1.2 Discussion The SQRT reviewed the equipment dynamic qualification information in FSAR Sec-tions 3.9.2 and 3.10 and made a plant site visit from May 7 through May 10, 1985.
The purpose was to determine the extent to which the qualification of equipment, as installed at Hope Creek, meets the criteria described above. A representa-tive sample of safety-related electrical and mechanical equipment, as well as instrumentation, included in both the nuclear steam supply system (NSSS) and balance of plant (80P) scopes, was selected for the audit. Table 3.3 identifies the equipment audited. The plant site visit consisted of field observations of Hope Creek SSER 4 3-7
i i
the actual, final equipment configuration and its installation. This was fol-lowed by a review of the corresponding design specifications, test, and/or anal-
! ysis documents which the applicant maintains in the central files. Observing i the field installation of the equipment is necessary to verify and validate i' equipment modeling employed in the qualification program. In addition to the document reviews and' equipment inspections, the applicant presented details of the maintenance, startup testing, and inservice inspection programs.
4 3.10.1.3 Generic Item-i Review of the qualification documentation for the reactor core spray pump and j
' motor (NSSS-1, MPL No. 1AP-206/E21-C001) revealed a questionable methodology used to demonstrate operability of the pump. .The demonstration of adequate clearance during the safe shutdown earthquake was obtained by subtracting un-signed peak (square-root-of-the-sum of-the-squares) displacements. The proper l
methodology is to subtract displacements at the modal level, then combine modal clearances according to any of the methodologies in RG 1.92. A spot check of two other pump analyses showed that questionable methodology was uniformly ap-
- plied. All equipment qualified by the response spectrum method, and which re-
, quires a clearance check, is affected.
By letter dated August 16, 1985 (R. Mitt 1, PSE&G, to W. Butler, NRC), the ap-plicant indicated that a recalculation of relative displacements with the cor-rect methodology was performed. All of the relative displacements were still l within the allowable limit of clearances. This item is, therefore, closed.
l 3.10.1.4 Equipment-Specific Item.
The required response spectra used in qualifying switches mounted in the standby
) liquid control (SLC) panel (NSSS-5, No. 145C3040P003) were obtained by calcula-i tions involving the floor response spectra and the dynamic characteristics of
, the panel. Because this methodology is rather new, a more detailed evaluation i is required. From the review of a short summary and discussion during the audit, it appears that (1) the methodology is reasonable and (2) only two items in the
] plant (both-on the SLC panel) were qualified using it.
i
~
By letter dated June 11, 1985 (R. Mitti, PSE&G, to W. Butler, NRC), the appli-cant provided a detailed evaluation of the methodology. A review of the method-
- ology confirms that, in this particular case the results are acceptable. How-ever, the suitability of this methodology in a generic sense is not implied.
3.10.1.5 Confirmatory Issues Confirmatory issues are:
) (1) Inspection of the piping attached to the hydraulic control units (HCUs) i (one of which was NSSS-2, MPL No. C11-0001) revealed several spans of l'
piping that appeared to be inadequately supported. -The applicant indi-cated that the lines had been shown to be adequate by dynamic analysis.
i However, an inhouse walkdown had previously initiated additional evalua-tion to confirm the acceptability of the subject piping.
By letter dated August 16, 1985 (R. Mitti, PSE&G, to W. Butler, NRC), the applicant indicated that the piping was ana.lyzed for the anticipated
)
Hope Creek SSER 4 3-8
~. - _ . . _ . _ _ _ _ . . . _ . _ _ . _ _ - - - _ . _ _
dynamic displacements and found to be within code-allowable stress limits.
From results of this analysis, runs of piping were stiffened by installing additional supports. Design changes documented in design change packages DCP-7427 and -7429 through -7432 are in progress and approximately 50% com-plete as of the submittal date. The completion of the changes still has to be confirmed by the applicant.
(2) During the inspection of the HCU piping discussed above, a gang support at vertical member F56722Q appeared to be complete, except that one line was not attached, leaving an unsupported span approximately 14 ft long. The applicant indicated that the design is adequate, but that a quality control verification has not been performed.
By letter dated August 16,1985 (R. Mitti, PSE&G, to W. Butler, NRC), the applicant confirmed that the support (SP15Q3 on isometric 1-P-BF-249) is now complete and the line in question has been supported. This item is, therefore, closed.
(3) The qualification report for the ITT actuator attached to the control dam-per (80P-2, Tag No. 1HD-9603B1) consisted of a summary of testing performed at Wyle Laboratory by MCC Powers. For the qualification to be substan-tiated, the original test report by the test laboratory is required.
By letter dated June 11, 1985 (R. Mittl, PSE&G, to W. Butler, NRC), the applicant provided the test report. Review of the test report indicates that the damper is adequately qualified. This item is, therefore, closed.
(4) The ITT actuator for the control damper (80P-2, Tag No. 1HD-960381) was found to be poorly supported in the field. The applicant indicated, on inquiry, that a support modification was required and produced an existing design change package (OCP-299) that specified the installation of an addi-tional support for the actuator. Confirmation of the support installation was provided to the NRC staff by letter dated August 16, 1985 (R. Mittl, PSE&G, to W. Butler, NRC). This item is, therefore, closed.
(5) A draft response to draft SER Open Item 103 (see letter from applicant dated August 20,1984) was reviewed and found acceptable during the audit.
The formal response has been included in A.mendment 11 to the FSAR.
(6) Inspection of-the reactor pressure vessel (RPV) level and pressure rack (NSSS-9, Tag No. 10C-026/H21-P026) revealed a poorly supported run of SST tubing attached to a panel-mounted device (1321-NOA50). The applicant indicated that the tubing had not been approved by either the engineering or the quality control department. Work was still in progress for this
~
tubing; additional support will be provided in accordance with established acceptance criteria (Bechtel Power Corporation Specification 10855-J-S-1303).
The NRC staff was notified of the support installation by letter dated August 16, 1985 (R. Mitt 1, PSE&G, to W. Butler, NRC).
3.10.1.6 Summary On the basis of the observation of the field installation, review of the quali-fication documents, and responses provided by the applicant to SQRT's questions during the audit, the applicant's seismic and dynamic qualification program has Hope Creek SSER 4 3-9
- - .. . . - . - - - ._= ..._.. - . _ _.
1 r i
i been found to be well defined and adequately implemented. Upon closure of the i
issues identified in Table 3.3, and provided that the conditions delineated in the above sections are met, the seismic and dynamic qualification of safety-related equipment at Hope Creek will meet the applicable portions of GDC 1, 2, j- 4, 14, and 30, Appendix B to 10 CFR 50, and Appendix A to 10 CFR 100 as they
', relate to qualification of equipment.
3.10.3 TMI Action Plan Item II.K.3.28 !
Safety analysis reports claim that air (or nitrogen) accumulators for the auto- !
I matic depressurization system (ADS) valves are provided with sufficient capacity to cycle the valves open five times at design pressures. GE has also stated
' that the emergency core cooling systems are designed to withstand a hostile en-vironment and still perform their function for 100 days following an accident.
Licensees and applicants must demonstrate that the ADS valves, accumulators, and associated equipment and instrumentation meet the requirements specified in the plant's FSAR and are capable of performing their functions during and fol-lowing exposure.to hostile environments, taking no credit for non-safety-related equipment or instrumentation. Additionally, air (or nitrogen) leakage through^
a valves must be accounted for to ensure that enough inventory of compressed air
- is available to cycle the ADS valves. If this cannot be demonstrated, it must j be shown that the accumulator design is still acceptable.
J The commitment to satisfy the requirement of TMI Action Plan Item II.K.3.28 for Hope Creek is discussed in the applicant's letter from R. Mitt 1, PSE&G, to
' A. Schwencer, NRC, dated March 27, 1985 (response to requests for additional information).
j The design of Hope Creek is such that the ADS will be available for 100 days following an accident. During the time immediately following an accident, air '
is available to actuate each ADS valve from its own local accumulator. The supply of air to these accumulators, which have a capacity of 10 gal is provided from '
the primary containment instrument gas system (PCIGS). This system is classified as safety related, seismic Category I, and environmentally qualified.
(
)
During normal operation, the two1PCIGS headers are interconnected to provide i compressed gas to all PCIGS users (this includes the ADS accumulators). After a design-basis event, the headers are isolated from each other, with one header l supplying three of the ADS valves and the other supplying the remaining two ADS i valves.
I :
j
.The individual ADS accumulators are reported capable of providing the following number of safety / relief valve (SRV) actuations before draining the pressure ;
- required to operate the valves
l (1) 64 cycles at 0 psig drywell pressure '
l
! (2) 17 cycles at 43.8 psig (70% of drywell design pressure)
Should the makeup air system (PCIGS) fail, the individual accumulators 'are capa-L ble of providing at least one SRV operation against a drywell pressure of 43 psig
- up to 1.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, with less than or equal to 1 standard cubic foot per hour (scfh) l pneumatic leakage. This period accounts for the 20 to 30 min required to manu-ally realign the PCIGS compressor on the emergency diesel following an accident.
Hope Creek SSER 4 3-10
l I;
The two PCIGS receivers have a volume of 225 ft3 and cycle between 90 and 105 psig. High-high and low-low receiver pressures are annunciated in the main control room to alert the. operator to problems at'the receiver.
The allowable leakage criteria of 1.scfh is based on GE experience from previ-ous environmental qualification and ongoing NUREG-0588 tests simulating harsh and seismic environment. The results of these tests shall verify that SRV pneu-matic leakage will not exceed 0.5 scfh. Additionally, no credit is taken for non-safety-related equipment in establishing the criteria or in the operation of the system.
The check valves in the ADS accumulator subsystem are tested in accordance with the requirements of ASME Code, Section 1, Subsection IWV, " Inservice Testing of Valves in Nuclear Power Plants." These tests are conducted to (1) demonstrate that the check valves open on reversal of pressure differential and (2) that the disc travels to the seat promptly on stoppage or reversal of flow. The fre-quency of check valve testing is in accordance with inservice test program re-quirements. The following surveillance tests are performed on the ADS:
(1) At least once per 31 days, a channel functional test of the primary PCIGS low-low pressur.e alarm is performed.
(2) At least once per 18 months, a channel calibration of the PCIGS low-low
~
pressure alarm is perfcrmed. -
The ADS accumulator system, backup system, and associated equipment and control circuitry located in a harsh environment are environmentally qualified for the conditions postulated to exist at their locations.
In reviewing the submittal, it is evident that the capability of the ADS depends on the operation of the PCIGS. Except for a period of approximately 20-30 min following an accident when the PCIGS is manually realigned on the emergency power supplied by the diesels, there is a continuous supply of' air to the accu-mulators. During this 20- to 30-min period, the accumulator inventory, good for approximately 1.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and one actuation, is the source of air for the ADS valves. The applicant has confirmed that the emergency operating procedure re-quires that the.PCIGS compressor be realigned onto the diesel and that the con-tainment isolation signal be overridden permitting air to ~be supplied to the accumulators when accumulator pressure reaches a low point of approximately 40 psig.
The applicant stated in the submittal: "Should the makeup air system fail, each accumulator is capable of providing at least one SRV operation up to 1.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following the loss of the makeup system, with less than or equal to 1 scfh pneu-matic leakage." Therefore, the applicant has demonstrated that Hope Creek has the capability for both short- and long-term cooling. The staff finds this acceptable.
The allowable leakage criteria.of 1 scfh includes an SRV leakage of 0.5 scfh.
This is based on leak rate results from tests of the ADS SRV pneumatic opera-tors, GE environmental qualification tests, and recent NUREG-0588 tests. These tests reportedly have shown that, for well beyond the required time period for ADS /SRV operation, SRV pneumatic system leakage will not exceed abcut 0.5 scfh.
The. staff finds this acceptable.
Hope Creek SSER 4 3-11
The applicant has provided information acceptable to the staff indicative of the cevelopment of surveillance, maintenance, and leak testing programs for the ADS accumulator system and associated alarms and instrumentation.
The applicant has provided information confirming that (1) The normal and makeup air supplies are seismically and environmentally qualified.
(2) The accumulators and associated equipment are capable of performing their functions during and following an accident, while taking no credit for non safety-related equipment and instrumentation.
On the basis of the information provided by the applicant and the evaluation performed, the staff concludes that the applicant has verified the qualification of the accumulator (s) on ADS valves for Hope Creek, thereby satisfying the re-quirements of TMI Action Plan Item II.K.3.28.
i Hope Creek SSER 4 3-12
Table 3.1 Water depth and duration of flooding at plant grade and windspeed and direction at Hope Creek site Water depth at Duration 1 Windspeed Direction power block (ft) (hr) (mph) ()
>12 0.2 110 180
>10 0.6 105 196
>8 0.9 97 208
>6 1. 3 89 218
>4 1.6 70 224
>2 1.5 58 230
>0 2.6 53 232 1 Duration considers only the time the wind direction is blowing from the river toward the plant.
Table 3.2 Potential number of floating missiles at Hope Creek site Severe Dam storm Riverborne missiles failurel runaway 2 Large self propelled vessels 12 1 Naval vessels 50 0 Rail cars 50 0 Self propelled vessels 70 1 Tugs 100 0 Non-self propelled vessels (barges) 150 2 Total Uf 4 1The staff did not consider vessels with displacements less than 52,000 lb.
2 Conservative estimates based on discussions with the U.S.
' Coast Guard, the Pilots Association, and the Philadelphia Maritime Exchange. Most of the dam failure missiles are substantially north of the Hope Creek site. In any postu-lated high-wind /high-water scenario, any runaway vessel would be pushed north. As such, only vessels to the south of Hope Creek, with displacements greater than 25,000 lb, are vessels of possible concern.
Hope Creek SSER 4 3-13
?E Table 3.3 Equipment audited 3
Q SQRT Applicant Equipment name g ID no. ID no. and description Safety function Findings Resolution Status w
us BOP-1 1-EE-HV-4680 6-in. gate valve Isolates suppression Qualified X; pool / torus water cleanup 30 system from torus given
- a containment isolation signal BOP-2 1HD-9603B1 Control damper, Controls flow of out- (1) Qualifica- (1) Test report Qualified duct mounted side air to diesel- tion report has been submitted generator building was a summary and reviewed.
description. (2) Applicant has Original test confirmed support report is installation.
required.
u, (2) Damper's j, actuator was 4-poorly sup-
. ported because a support had not been in-stalled (DCP-299).
BOP-3 1EA-TPBC516-2 Termination Supports. transmission Qualified panel of electrical signals for station service water systes BOP-4 IKJ-P175380 Pressure Provides 1ccal indica- Qualified gage tion of pr4ssure in standby diesel generator starting a1d control air tank 1
'l
-- - ~ -
x 4j Table 3.3 (Continued) m Q SQRT Applicant Equipment name
$ ID no. ID no. and description Safety function Findings Resolution Status w
8;
- BOP-5 .1GQ-FSL9771D Flow switch Shuts down service Qualified gg water intake structure
,, supply fan DV-503 when system low flow is sensed B0P-6 108242 480-V motor Provides power to Qualified control center Class IE equipment 80P-7 18-21F037E Safety / relief Relieves vacuum in an Qualified valve (SRV) SRV line resulting from (vacuum breaker) steam condensation u, BOP-8 IEC-TE46758 Resistance Is part of fuel- pool Qualified
,L temperature pressure boundary on detector
~
BOP-9 10C-399 Remote shutdown Allows safe shutdown Qualified panel of reactor remote from control room BOP-10(a) HSS-4416B Rotary switch on Transfers control of Qualified 4'
10C-399 residual heat removal (RHR) pump BP202 from control room to remote shutdown panel 10C-399 BOP-10(b) TR-A201 Relay on 10C-399 Controls cooling suction Qualified inboard isolation valve
- HVF009 80P-10(c) SRU-C399Al Signal resistor Converts 4-20 mA to 1-5 V Qualified unit on 10C-399 de signal for suppres-sion pool level and reac-tor pressure RHR interlock 1
- r y Table 3.3 (Continued) e a
SQRT 10 no.
App 1fcant ID no.
Equipment name and description Safety function Findings Resolution Status M 80P-10(d) HS-F045 Switch on Controls reactor core Qualified Q 100-399 isolation cooling tur-
, bine shutoff valve 80P-11 1AT84207 Lube oil switch Monitors diesel genera- Qualified
- panel tor lube oil pressure; shuts down the generator and provides alarm sig-nal on loss of pressure MSSS-1 1AP-206/E21- Reactor core Supplies makeup water / A questionable The relative dis- Qualified
, C001 spray pump cooling to reactor methodology was placements were re-i and motor during loss-of- used to ensure calculated with the coolant accident clearance in correct methodology.
w the pump during The results were j
4 m
the safe shut- still within the 1
down earthquake. allowable limits.
NSSS-2 C11-D001 Hydraulic Operates scram valves (1) Associated (1) Very flexible Confirm-control unit on reception of a piping appeared piping runs are atory i scram signal in order flexibly sup- being stiffened.
! to force control rod ported. (2) The piping sup-insertion into the core (2) A support port was not com-for the asso- plete during the ciated piping walkdown. This j (vertical mem- support is now com-ber F56722Q) plete and the line appeared to be in question has complete except been supported.
that one line
- was not at-tached leaving approximately
- 14 ft of piping i unsupported.
1 i
i
L j
[* Table 3.3 (Continued) 1
! ? SQRT Applicant Equipment name
- - 8 ID no. ID no. and description Safety function Findings Resolution Status
- w l g NSSS-3 18FSVF18201/ Control rod Shuts off air to and 1 m C11F182 drive solenoid vents air from scram Qualified
- valve discharge volume isola-tion valve air header NSSS-4 163C1303 Limit switch Provides a signal to Qualified
} reactor protection sys-tem upon turbine valve ,
3 closure NSSS-5 10C-011/H21- Stan6y liquid Monitors SLC system In-cabinet Detailed descrip- Qualified i
P011 control (SLC) panel spectra were tion of the meth-generated. odology was accept-
, ising a new able in this par-4 y
methodology. ticular case but not on a generic ,
j basis.
I
! NSSS-6 1CP-216/E41- High pressure Is attached to a 1
C002 coolant injec- Class IE bus;. failure Qualified
' tion (HPCI) could short the. bus j gland seal pump j and motor 1 -
j NSSS-7 188V004/ Recirculation Is part of reactor Qualified l 831F0238 suction valve coolant boundary NSSS-8 163C1563 Pressure Is a passive element in Qualified a
transmitter a Class IE circuit and j is part of a pressure boundary I
i z
Ei Table 3.3 (Continued)
- o n
- 2 SQRT Applicant Equipment name 4
f ID no. ID no, and description Safety function Findings Resolution Status 1
l IS NSSS-9 10C-026/ Reactor pressure Monitors reactor vessel A poorly sup- Construction in- Qualified j E H21-P026 vessel. level and level and pressure ported run of sta11ation of the
- . pressure rack tubing was SST tubing supports 4
found attached has been completed.
! to a panel-i mounted device.
This deficiency 4
was listed on '
- the Construc-a tion Turnover l Exception list.
4 NSSS-10 10C-6081/ Power range Monitors reactor power Qualified 2
w H11-P608 neutron monitor level
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7 INSTRUMENTATION AND CONTROLS 7.2 ' Reactor Protection (Trip) System '
7.2.2 Specific Findings 7.2.2.3 Testability of Plant Protection Systems at Power As stated in the SER, the applicant was in the process of completing a review of the o:eline testability of the balance-of plant (80P) instrumentation. As a confirriatory item, the staff was to review the results of this review and con-firm that the applicant's analysis would demonstrate the existence of the capa-bility for at power testability for 80P systems or that acceptable justifica-tions would be provided.
By letter dated August 13, 1984 (R. Mitti, PSE&G, to A. Schwencer, NRC), the applicant provided the completed results of the review of the BOP instrumenta-tion. During the review, the at power testability of an item was established if an affirmative response could be verified for the following three questions:
(1) Is the item sufficiently accessible to conduct the test during normal
, operation?
l (2) Is the item sufficiently isolatable to permit its safety-related . function to be verified, or is a safety-related system or subsystem encompassing the item isolatable and testable?
(3) Does any bypassing method that must be used to accomplish the test conform to Position C.6 of RG 1.118?
Using the above criteria the applicant has determined that the following safety systems items not related to the nuclear steam supply system were untestable at power.
Primary Containment Isolation System (PCIS)
(1) 1he loss-of-coolant-accident (LOCA) signals of reactor low level (level 1),
drywell high pressure, or manual initiation originating from core spray system relay K18A-D cannot be completely isolated for testing at power.
This affects actuation signals to close their associated containment iso-lation valves, to trip 16 motor control center breakers, and to initiate control room isolation. However, all other methods for actuation cf this equipment can be verified at power. The LOCA signals and manual initiation i
are tested at least once every 18 months (550 days).
(2) The coincidence circuitry for the reactor building area and refueling floor area high-high radiation signals cannot be completely isolated for testing at power. The individual high-high radiation signals can be veri-fied up to the input buffers of the logic modules but must be tested one i
l at a time because each signal is transmitted through isolation devices to l
Hope Creek SSER 4 7-1
all four channels of the PCIS simultaneously. This only affects the logic circuitry of the PCIS itself and does not inhibit the testing of the actual actuation signals from the PCIS to the individual actuated compo-nents. The coincidence circuitry for these two signals is tested at least once every 18 months (550 days).
On the basis of the above information, the staff has concluded that the BOP por-tion of the Hope Creek design has provided adequate on-line testing capability for the actuation instrumentation channels, logic, and actuation devices of safety systems and that adequate justification has been provided for the two exceptions identified. Therefore, the staff has found the 80P on-line testing capability at Hope Creek acceptable and has concluded that Confirmatory Item (1) as listed in SER Section 7.1.4.2 is resolved.
The staff will verify that the Technical Specifications for Hope Creek include appropriate surveillance requirements to require periodic (on-line) demonstra-tion of the operability of the reactor protection system (RDS) and engineered safety feature (ESF) instrument channels logic and actuation devices.
7.2.2.5 Instrumentation Setpoints In the SER, the staff identified a concern with respect to instrument setpoints for the reactor protection system. The staff determined that additional infor-mation was required to confirm the applicant's conformance with the Commission's regulations relevant to the issue of protection system setpoints.
GDC 20, " Protection System Functions," states:
The protection system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.
10 CFR 50.36(c)(1)(ii)(A) states:
Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having signif-icant safety functions. Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting shall be so chosen that automatic protective action will cor-rect the abnormal situation before a safety limit is exceeded.
10 CFR 50.46 specifies the performance criteria for the emergency core cooling systems.' These criteria include a maximum peak cladding temperature, maximum cladding oxidation, maximum total amount of hydrogen generated, and require-ments that core geometry remain amenable to cooling for long-term decay heat removal. Guidance on acceptable methods for complying with these regulations is contained in RG 1.105, " Instrument Setpoints."
To conserve resources and to take advantage of an ongoing review effort, the applicant joined with several other BWR owners that had formed a Licensing Review Group (LRG) - Instrumentation Setpoint Methodology Group (ISMG) - so Hope Creek SSER 4 7-2
that the requested information could be provided to the NRC. The applicant's commitment to follow ISMG developments was provided in a letter dated February 15, 1985 (R. Mitt 1, PSE&G, to A. Schwencer, NRC).
On July 14, 1983, the NRC staff met with the ISMG at their request. At this meeting the ISMG presented an outline of a setpoint methodology. In response to additional questions from the NRC staff, another meeting was held on January 31, 1984. By letter dated May 15, 1984 (T. M. Novak, NRC, to J. F. Carolan, Chairman, ISMG), the NRC staff provided its assessment of the ISMG methodology. The staff's evaluation identified several deficiencies in the methodology and requested that the ISMG provide additional information in response to 10 specific concerns 'In response to the staff's evaluation, by letter dated June 29, 1984 (J. Carolan to T. M. Novak), the ISMG provided an action plan for resolving the outstanding issues. By letter dated July 23, 1984 (B. J. Youngblood, NRC, to J. F. Carolan),' the NRC staff accepted the proposed action plan.
In addition, a letter dated May 24, 1985 (J. F. Carolan to J. J. Stefano, NRC) forwarded a draft copy of the proposed instrument setpoint methodology. The draft document stated: "It is incumbent upon the utility to establish calibra-tion and survei.llance testing procedures which support design bases utilized in the evaluations regarding calibration and drift." The draft document fur-ther mentioned that the utilities are responsible for (1) surveillance testing procedures '
(2) calibration procedures (3) environmental qualification program In addition to following ISMG developments and the action plan, the applicant in a letter dated February 15, 1985 (R. Mitti, PSE&G, to A. Schwencer, NRC) has committed to provide the resulting setpoint methodology (ISMG). This methodol-ogy will address the staff concerns listed below and document the plant unique setpcint evaluations of each RPS and ESF trip function assumed to operate in the analyses described in FSAR Chapters 6 and 15 for Hope Creek.
[1) Identification of the values assigned to each component of the combined channel error allowance (e.g., modeling uncertainties, analytical uncer-tainties, transient overshoot, response times, trip unit setting accuracy, sensor accuracy, test equipment accuracy, sensor drift, nominal and harsh environmental allowances, and trip unit drift), the basis for these values, and the methods used to sum the individual errors. Where zero is assumed for an error, a justification that the error is negligible shall be pro-vided.
(2) Confirmation that-the setpoints selected for the initiation of protective actions ensure that the reactor core and reactor coolant system are pre-vented from exceeding the licensing safety limits for the transients and accidents analyzed.
On the basis of staff participation in meetings with the ISMG, the work per- 3 formed by the ISMG, the applicant's commitment noted above, and the conserva- i tisms inherent in the safety analyses and the protection system design, the staff concludes that there is reasonable assurance that the results of the ISMG j effort will verify the acceptability of the proposed setpoints and, therefore, Hope Creek SSER 4 7-3 !
that plant operation in the interim until the ISMG effort and plant-specific technical assessment are complete is acceptable. The staff has learned that instrument uncertainties are treated properly. That is, the independent uncer-tainties are combined by the root mean square method, whereas the dependent uncertainties are combined algebraically. Moreover, the setpoints to be used at Hope Creek are consistent with those used at other operating BWR plants.
Therefore, the staff concludes that Confirmatory Item (2) as listed in SER Section 7.1.4.2 is resolved.
7.5 Safety-Related Display Instrumentation 7.5.2 Specific Findings 7.5.2.4 Bypassed and Inoperable Status Indication The applicant indicated that bypassed and inoperable status indication has been provided for the reactor protection system, engineered safety features, and other systems required for safety, including auxiliary and support systems.
Automatic indication is provided in the control room to inform the operator that a system is out of service. Indication lights show which part of a system is out of service. Examples of out-of-service conditions that automatically energize the annunciators are pump motor breaker in pull-to-lock position, by-pass or test switches actuated, loss of motor-operated valve control power or overload conditions, remote shutdown panel takeover, loss of pump motor control power, and diesel generator out of service.
Automatic bypass of certain infrequently used pieces of equipment is not pro-vided. RG 1.47 indicates that it may not be necessary to provide automatic indication of a bypassed or inoperable status condition if the condition is not expected to occur more frequently than once a year. Hcwever, manual activation of a system-level bypass is recommended and has been implemented at Hope Creek for those systems that have these infrequently used bypasses. Once a system has been manually bypassed, operability is verified by system testing before the system is placed back in service and the system bypass indicating light is deactivated. Inoperability of a support system actuates an out-of service alarm for the protection systems that are supported by that system.
The circuits of the bypassed or inoperable status-indication system are physi-cally~and electrically isolated from safety circuits they monitor so that no credible failure of the annunciator circuits will degrade the safety circuit below an acceptable level. The status-indication lights can be tested by depressing integrated test pushbuttons.
On the basis of its review of the bypassed and inoperable status indication, the staff finds that the applicant had provided an acceptable method of indi-cating the bypass or inoperable status of portions of the protection system, )
systems actuated or controlled by the protection system, and auxiliary support i systems. However, the information provided in FSAR Section 7.5.1.3.2 appeared i to conflict somewhat with FSAR Table 7.1-2 and Section 7.3.2.2 of the SER re- '
garding the applicability of the guidance of RG 1.47 to the ESF equipment area cooling system (as indicated in FSAR Section 7.5.1.3.2).
Hope Creek SSER 4 7-4 1
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As a confirmatory item, the staff was to v'erify that the descriptions provided in the FSAR regarding bypassed and inoperable status indication (RG 1.47) for the ESF equipment area cooling system are consistent and acceptable to the staff. This verification has been performed, and the staff has concluded that the bypassed and inoperable status indications for the ESF equipment area cool-ing system are acceptable.
Therefore, the staff concludes that Confirmatory Item (9) as listed in SER Sec-tion 7.1.4.2 is resolved.
7.6 Interlock Systems Important to Safety 7'.6.2 Specific Findings 7.6.2.1 Isolation of Low-Pressure System From the High-Pressure Reactor Coolant System During normal and emergency conditions, it is necessary to keep low pressure systems that are connecced to the high pressure reactor coolant system prop-erly isolated to avoid damage by overpressurization or the potential for loss of integrity of the low pressure system and possible radioactive releases. To accomplish this, at least two valves in series should be provided to isolate
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the low pressure system from the reactor coolant system. It is the staff's
, position (Branch Technical Position (BTP) ICSB 3 (NUREG-0800)) that where motor-l operated isolation valves are provided, the motor-operated valves should have independent and diverse interlocks to prevent the valves from opening (automat-ically or by remote-manual action) whenever the primary system pressure is above the subsystem design pressure and to close automatically whenever the primary system pressure exceeds the subsystem design pressure.
Low pressure coolant injection (LPCI) injection-line check valves F041 A, B, C, and D and relief valves F025 A, B, C, and D prevent overpressurization of the low pressure piping and components of this residual heat removal (RHR) discharge line when LPCI injection valves F017 A, B, C, and D are open. Additionally, to ensure that the lcw pressure piping and components are not overpressurized during routine surveillance and operability testing of injection valves F017 A, B, C, and D, the low leak rates (i.e., less than the capacity of relief valves F025 A, B, C, and D) of check valves F041 A, B, C, and D are verified before opening the injection valves by opening drain line valves V530 and V537 for a brief period and verifying that the differential pressure across the F017 valves, as indicated by differential pressure switches N7658 A, B, C, and D, is lowered. This routine verification of low leak rates of the F041 check valves provides a high level of assurance that tha leak rates will not be 'ex-cessive even if the F017 injection valves are to open during an accident when reactor pressure is greater than the design pressure of low pressure piping of the RHR system.
This design permitted the injection valve to open when the differential pres-sure across the valve was equal to or less than 730 psi. Therefore, the injec-tion valve could open when the reactor pressure was equal to 1,080 psig (i.e.,
730 psi plus the LPCI pump discharge pressure of approximately 350 psi). The staff position is that this design is unacceptable because a single failure of Hope Creek SSER 4 7-5
the inboard check valves (F041 A, 8, C, and D) could result in overpressuriza-tion of the_LPCI low pressure piping upstream of the injection valve (causing a LOCA).
The applicant proposed a design that would eliminate the single-failure concern by preventing the injection valve from opening when the pressure downstream (sensed by pressure transmitter PT-N058) of the injection valve is greater ,
than the design pressure of the LPCI piping. The pressure-indicating switch (PIS-N658) has a nominal trip setpoint (NTSP) of 460 psi. Pressure downstream of the injection valve must be equal to or less.than this NTSP before the auto-matic or manual open signal will be transmitted to the injection valve. There-fore, the LPCI low pressure piping that has a design pressure of 500 psi cannot be overpressurized by injection valve openings.
The staff finds that the LPCI interlock design satisfies BTP ICSB 3 and is, therefore, acceptable. However, as a confirmatory item, the applicant was to provide the final LPCI interlock design drawings for staff review and final confirmation of acceptance. These drawings were provided by letter dated September 18, 1985 (R. Mitt 1/R. P. Douglas, PSE&G, to Walter Butler, NRC). The staff concluded that the drawings reflect the proposed designs and are therefore acceptable. Therefore, the staff concluded that the LPCI interlock design is acceptable and Confirmatory Item (10) as listed in SER Section 7.1.4.2 was re-solved. However, because of ongoing staff concerns with other 8WRs, the staff decided to reexamine the high pressure / low pressure boundary devices associated with the remote shutdown system.
The RHR valves controlled at the remote shutdown panel (RSP) are interlocked by pressure switches to prevent inadvertent breach of the high pressure / low-pressure boundary when reactor pressure is above RHR system design pressure.
These valves are E11-HV-F008 (outboard shutdown isolat an), E11-HV-F009 (in-board shutdown isolation), E11-HV-F0158 (shutdown cooling injection), Ell-HV-F022 (reactor head spray), and E11-HV-F023 (reactor head spray). The staff reviewed the design information associated with these valves and found the de-sign of the pressure permissive interlocks for the RHR valves located at the RSP acceptable.
The only high pressure / low pressure boundary device associated with the alter-nate remote shutdown system is RHR valve E11-HV-F015A (shutdown cooling-injec-tion). There is no pressure permissive interlock associated with this valve.
However, it is controlled by a keylocked switch located in a motor control center (MCC); cons'equently, the operator would have to obtain the key from the
, normally locked RSP room before the control switch could be actuated and RHR
- valve F015A opened. The staff found this design acceptable but was concerned that an inadvertent manual actuation could occur and cause a breach of the high pressure / low pressure boundary. Because of this, the staff's acceptance of this design was based on the applicant providing a confirmatory written re-sponse to the following items
(1) The labeling for alternate remote shutdown system RHR valve E11-HV-F015A will be appropriately changed to reflect a warning that inadvertent actua-tion could cause overpressurization of the RHR piping. This labeling should be designed, fabricated, and located in accordance with good human factor *.
engineering principles.
, . Hope Creek SSER 4 7-6
It (2) Before full power operation, the applicant shall install a low pressure permissive indication (e.g. , blue light) so that the operator at the MCC i
panel will know when RHR valve F015A can be opened without causing a breach of the reactor pressure boundary because of overpressurization of the RHR piping. l t
By letter dated November 8, 1985 (R. Mitt 1, PSE&G, to W. Butler, NRC), the ap- !
l plicant committed to implement the above items. On the basis of this commitment, the staff considers this issue closed.
The staff will verify that the Technical Specifications include appropriate limiting conditions for operation and surveillance requirements on the high-pressure / low pressure interlocks.
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11 RADI0 ACTIVE WASTE MANAGEMENT 11.4 Solid Waste Management System The SER listed as a license condition that the applicant should obtain NRC approval of the solid waste process control program (PCP) before processing solid waste. In a letter dated August 21, 1985 (R. Mitt 1, PSE&G, to W. Butler, NRC), the applicant submitted the PCP to solidify Class A waste only. The Hope Creek solid waste processing design, a fully automated volume reduction system that removes all water and mixes the dried solids with asphalt, was designed and constructed without any provisions for sampling the product. Drums are filled, capped, measured for surface contamination, tagged, and stored automat-ically for future shipment by equipment inside a shielded space.
Because the product cannot readily be sampled and evaluated against prequalifi-cation standards, assurance that the product is acceptable lies mainly with the control of process parameters. To provide this control, the applicant's PCP requires (1) sampling and testing of the supplier's asphalt at receipt before use, (2) establishing process parameters during prequalification testing, (3) adhering to these process parameters during production, (4) appropriate operator training and qualification, (5) document control of the product, and (6) an administrative / management quality assurance program.
Even though the PCP does not provide for sampling and evaluation of the end product, the staff finds the PCP acceptable to process Class A waste. The conclusion is based on the following considerations:
(1) The PCP utilizes incoming sampling / evaluation to ensure compliance with material qualified during prequalification testing.
(2) Appropriate quality assurance steps (operator training, management involve-ment and audits, and process parameters established during prequalification testing on the various waste streams) are in force to control the operation of the system.
(3) To provide assurance the system will be operated properly, operating procedures are specified for system operation, waste feedstream sampling and evaluation, waste classification, waste product documentation, and signoff control before release for shipment.
Before Class 8 and C production level waste processing, the staff will review the prequalification test results and the PCP document prepared for the hand-ling of Class B and C waste processing.
The staff concludes that the PCP provides reasonable assurance that the waste product will meet the requirements for Class A wastes and, therefore, is acceptable. However, until the PCP is revised to include the solidification of Class 8 and C wastes, the proposed ifcense condition, as identified in the SER, will be changed to read:
Hope Creek SSER 4 11-1
Before processing Class B and C solid wastes, the applicant must obtain NRC approval of the Class B and C solid waste process con-trol program addressing the requirements of 10 CFR 61 and Branch Technical Position ETSB 11-3.
Hope Creek SSER 4 11-2
14 INITIAL TEST PROGRAM 14.2 Initial Plant Test Program - Final Safety Analysis Report By Generic Letter 83-24, "TMI Task Action Plan Item I.G.1, 'Special Low Power Testing and Training,' Recommendations for BWRs," applicants for a license were requested to commit to the recommendations of the BWR Owners Group with respect to this additional testing and to respond to the generic letter by analyzing or demonstrating the effects a proposed station blackout (580) test would have on plant equipment.
In a letter from N. W. Curtis, Pennsylvania Power and Light Company (PP&L), to A. Schwencer, NRC, dated June 15, 1982, PP&L presented the results of the 580 event for the Susquehanna Steam Electric Station. These results indicated the potential for damage to plant equipment if the 580 test were performed at Sus-quehanna. In Supplement No.3 to the Susquehanna SER (NUREG-0776), the staff accepted this evaluation and did not require the 580 test to be performed at Susquehanna. By letter dated September 17, 1985 (R. Mitt 1, PSE&G, to W. Butler, NRC), the applicant responded to Generic Letter 83-24. The appilcant utilized the Susquehanna evaluation and noted that the Hope Creek drywell design is 70%
the size of the Susquehanna drywell with approximately the same reactor rated thermal power level. Therefore, Hope Creek would have at least the same poten-tial risk of equipment damage as Susquehanna, and the temperature in the drywell would have a more rapid response to an SB0 test than at Susquehanna. Because of the potential to damage non-safety related equipment in the drywell, the ap-plicant proposes, consistent with the results at the Susquehanna Steam Electric Station, that the SB0 test not be performed at Hope Creek.
The staff agrees with the applicant's evaluation and finds that the 580 test need not be performed at Hope Creek because of the potential risk of damage to non-safety-related equipment in the drywell. Also, because the applicant has committed to the BWR Owners Group recommendations for augmented testing, the staff concludes that the Hope Creek test program described in the FSAR, without the 580 test, meets the requirements of Item I.G.I.
Hope Creek SSER 4 14-1
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a APPENDIX A CONTINUATION OF CHRONOLOGY September 10, 1985 Letter from applicant informing that GA Co contracted to provide radiation monitoring system (RMS) for facility be-cause Technology for Energy Corp filed for reorganization under Chapter 11 of Bankruptcy Act on March 29, 1985.
Deferral of certain RMS equipment requested.
September 12, 1985 Letter to applicant advising that methods of procuring re-placement parts for ASME Code,Section III, Class 3 pumps where current manufacturer has no ASME Code authorization l and N-type stamp are acceptable and that applicant must perform appropriate vendor survey.
September 16, 1983 Letter from applicant requesting that independent design and verification program (IDVP) protocol be formally dis- ,
solved as soon as practicable because final report has been issued and all IDVP matters are part of public record.
September 16, 1985 Letter from applicant providing response to SER Outstanding Issue 1, "Riverborne Missiles." Integrity of service water pump shaft seal and flange seal will not be compromised by floating missile.
i September 17, 1985 Letter from applicant responding to TMI Action Plan '
Item I.G.1, "Special Low Power Testing and Training," and Generic Letter 83-24 regarding station blackout (S80) test.
580 test requirement should be deleted because of result-ing adverse impact on plant equipment.
September 18, 1985 Letter from applicant forwarding Revision 10 to Elementary Diagrams 10855-N1-E11-1040-383(11) and 10855-N1-E11-1040-1 383(15) for residual heat removal system for staff review regarding SER Confirmatory Issue 24.
l September 18, 1985 Letter from appifcant forwarding response to Confirmatory Issue 5 and draft SER Open Item 103/SER Open Item 2, " Equip-ment Qualification," per May 7-10, 1985, Pump and Valve d
Operability. Review Team / Seismic Qualification Review Team audit. Information provided will be included in Amend-ment 13 to Final Safety Analysis Report (FSAR).
September 18, 1985 Letter from applicant forwarding response to May 8, 1985, J
request for additional information on emergency action levels and NRC staff comments on station dose assessment methodology. Information on emergency action levels has been incorporated into Revision 3 to emergency plan.
Hope Creek SSER 4 1 Appendix A
September 20, 1985 Letter from applicant forwarding detailed justifications on power ascension test modifications and marked-up FSAR changes reflecting proposed modifications. Enclosed safety evaluations find that proposed modifications pose no risk to public. Expedited review requested. FSAR will be amended.
September 20, 1985 Letter from applicant forwarding September 3,1985, marked-up draft Technical Specifications, indicating changes to second draft submitted on September 4, 1985. Changes con-cern FSAR and as-built plant and support' issuance of proof-and-review copy.
September 20, 1985 Letter from applicant forwarding FSAR Amendment 12, which includes text changes resulting from resolution of SER open items and updated functional control diagrams and instru-ment electrical diagrams in Section 7.
September 20, 1985 Letter to applicant forwarding SER supporting June 11, July 3, and August 9, 1985, requests for authorization to e.iminate arbitrary intermediate pipe breaks. Use of ASME Code Case N-411 and portions of ASME Code,Section III, are acceptable.
September 20, 1985 Letter to applicant advising that Revision 2 to visual weld acceptance criteria for structural welding at nuclear power plants are not applicable to inservice inspections required by ASME Code,Section XI. FSAR amendment docu-menting proposed criteria will be reviewed.
September 23, 1985 Letter from applicant forwarding corrected FSAR m1rkup to replace Attachment 2 submitted with September 20, 1985, power ascension letter.
September 25, 1985 Letter to applicant agreeing that applicant's IOVP protocol can be terminated because final IDVP has been issued. IOVP documentation subject to protocol should continue to be maintained in location accessible for NRC staff examination.
September 26, 1985 Letter to applicant requesting formal confirmation that offsite emergency plans include list of local or regional medical facilities and commitment to full compliance with s NRC response to GUARD remand in order for NRC to issue license.
Sentember 27, 1985 Generic Letter 85-18, " Operator Licensing Exams."
September 30, 1985 Letter from applicant forwarding " Simplification of Power Ascension Test 20, Pressure Regulator" and " Simplification of Power Ascension Feedwater System Response Testing -
Test 21A," for NRC expedited review. Marked-up FSAR pages on test modifications enclosed.
Hope Creek SSER 4 2 Appendix A
September 30, 1985 Letter from applicant confirming that all electrical equip-ment important to safety will be environmentally qualified before fuel load, per staff's August 28, 1985, letter on compliance with 10 CFR 50.49 and Generic Letter 85-15. No extensions to deadline anticipated.
October 1, 1985 Letter from applicant responding to September 4, 1985, request for additional information on loss of control room ventilation. Analyses were performed on effect of loss of ventilation and cooling to control room and electrical equipment room.
October 1, 1985 Letter from applicant responding to SER Outstanding Issue 1, "Riverborne Missiles," per September 23, 1985, telcon. Review of service water pump (SWP) f'ange seal and bearing assembly design complete. SWPs seismically qualified to 0.2g acceleration.
October 1, 1985 Letter from applicant commenting on draft environmental protection plan submitted by NRC staff letter of July 26, 1985. Salt drift monitoring program will be implemented in plant vicinity.
October 2, 1985 Letter to applicant requesting additional information on nonmetallic materials used for Class E system snubbers and reactor core isolation cooling turbines, documents in public domain concerning nonmetallic materials capabilities, and confirmation of replacement intervals.
i October 2,1985 Letter to applicant advising that sufficient information i
was submitted to resolve SER Open Item 3 on preservice inspection program. Results of NRC review will be in Supplement No. 3. Relief requests from preservice inspec-tion requirements of ASME Code should be submitted by 4
l October 17, 1985.
l October 2, 1985 Letter to applicant forwarding " Audit of Pump and Valve l
Operability Assurance Program for Hope Creek Generating i, Station."
! October 4,1985 Letter from applicant forwarding Tests 288, " Recirculation i
Pump Trip Test," and 280, " Recirculation Pump Runback Test,"
per attached power ascension test modifications, for expe-dited review. Safety evaluation and marked-up FSAR pages also enclosed.
October 4, 1985 Letter to applicant forwarding proof-and-review Technical Specifications. Identification if sections that do not accurately reflect FSAR or as-built plant requested.
{ October 4, 1985 Letter to app 1tcant advising that revised process control i
i program is acceptable for use in processing Class A wastes, Proposed license will not be changed until process control l
i Hope Creek SSER 4 3 Appendix A
program is revised to include solidification of Class 8 and C wastes.
October 10, 1985 Letter from applicant requesting permission to procure re-placement service water spray water pump without N stamp from Hayward Tyler. Pump will meet all ASME Code, Sec-tion III requirements, including all quality assurance requirements, with exception of N stamp.
October 14, 1985 Letter from applicant forwarding information addressing request to consider additional safety parameter display system (SPDS) display parameter verification and valida-tion process, per August 27-28, 1985, audit exit meeting.
1 October 15, 1985 Letter to applicant forwarding SPDS audit report detailing results of August 27-28, 1985, audit for review.
October 16, 1985 Letter from applicant responding to October 2, 1985, letter regarding preservice inspection program relief requests.
October 16, 1985 Letter to applicant forwarding proof-and-review environ-mental protection plan revised to incorporate applicant's October 1, 1985, comments.
October 17, 1985 Letter from applicant forwarding " Hope Creek Generating Station Report of Evaluation of Effects of Radio Frequency Interference on Digital Solid State System" for review.
Report closes SER Outstanding Issue 5 regarding solid-state logic modules.
October 17, 1985 Letter from applicant forwarding detailed justifications and safety evaluations for power ascension test modifications.
October 18, 1985 Letter from appitcant forwarding elementary diagrams of automatic depressurization system showing completed instal-lation of manual switch in response to SER Confirmatory Item 37.
October 21, 1985 Letter to applicant advising that Revision 4 to security plan and Revision 1 to security contingency plan are con-sistent with regulatory requirements.
October 24, 1985 Letter to applicant forwarding comments on Offsite Dose Calculation Manual (ODCM) submitted on April 26, 1985.
l October 25, 1985 Letter from applicant forwarding Revision 9 to emergency plan.
l October 25, 1985 Letter from appItcant forwarding information regarding l NRC policy statement on emergency planning standard,
! 10 CFR 50.47(b)(12), and interim guidance concerning
) GUARD vs NRC court decision, per Septerr.ber 26, 1985, l request.
Hope Creek SSER 4 4 Appendix A
i October 28, 1985 Letter from applicant informing that proof-and-review envi-ronmental protection plan submitted by October 16, 1985, letter is acceptable with one comment. Term "& Audit" ,
should be removed from Section 5.1 title and table of con- -
tents because section does not discuss auditing.
October 28, 1985 Letter to applicant forwarding evaluation of applicant's July 12, 1985, submittal of pump and valve inservice test-i ing program.
! October 29, 1985 Letter to applicant acknowledging receipt of May 23, 1985, i
letter submitting additional information on quality assur-
) ance program to be implemented during operations phase.
October 29, 1985 Letter from applicant forwarding response to Generic Letter 85-18, " Operator Licensing Exams."
October 30, 1985 Letter from applicant forwarding draft Revision 8 to Oper-ating Procedure OP-AP.ZZ-101(Q), " Post-Reactor Scram /ECCS Actuation Review and Approval Requirements," in response i to staff's request for inforaation on Generic Letter 83-28.
October 30, 1985 Letter from applicant fo .t.rding additional information i on IOVP in response to Nk. staff's request of September 30, 1985. t October 30, 1985 Letter from applicant forwarding " Single Failure Analysis e
for Neutron Monitoring and Process Radiation Monitoring System" per Regulatory Guide 1.75, to closc SER Confirmatory Issue 18 and in response to action item to September 24-26, 1985, audit.
October 30, 1985 Letter to applicant forwarding draft SA.'C-85/1514-8, "Re-3 view of Licensee and Applicant Responses to NRC Generic 1
Letter 83-28, Item 1.2, ' Post-Trip Review:Otat and Infor-mation Capabilities'... " technical evaluation report.
October 31, 1985 Letter to applicant forwarding Supplement No. 3.
i November 5, 1985 Letter from applicant confirming information provided dur-ing October 30, 1985, telcon between NRC staff and appli-cant regarding existing design features that would provide ultimate protection capability.
November 6, 1985 Letter from applicaat forwarding PSE-SE-Z-017. "Elimina-tion of 00-11 Process Computer Test From Power Ascension Testing - Test 11," and PSE-SE-Z-024, " Deletion of Trans-versing In-Core Probe Uncertainty, Test 16."
i November 8, 1985 Letter from applicant formalizing results of November 6, '
1985, discussion regarding SER Section 7.6.2.1, " Isolation of Low Pressure System From High Pressure RCS."
j Hope Creek SSER 4 5 Appendix A t
November 12, 1985 Letter from applicant forwarding tabulation of FSAR commit-ments for August, September, and October 1985 and corre-sponding resolution for each commitment.
Hope Creek SSER 4 6 Appendix A
APPENDIX 8 BIBLIOGRAPHY Carolan, J. F., ISMG, letter to T. M. Novak, NRC, " Action Plan To Answer the NRC Staff Concerns on Setpoint Methodology for General Electric Supplied Protection System Instrumentation," June 29, 1984.
-- , letter to J. J. Stefano, NRC, " Instrument Setpoint Methodology Program,"
May 24, 1985.
Code of Federal Regulations, Title 10, " Energy," U.S. Government Printing Office, Washington, D.C. (includes general design criteria).
Curtis, N. W., Pennsylvania Power and Light Company, letter to A. Schwencer, NRC, " Safety Evaluation Report - Station Blackout Testing at SSES," June 15, 1982.
Franklin Research Center, TER-C5506-308, " Audit Procedures for Mark I Contain-ment Long-Term Program - Structural Analysis," June 25, 1982.
-- , TER-C5506-327, " Audit for Mark I Containment Long-Term Program - Structural Analysis for Operating Reactors - Public Service Electric and Gas Company,"
Nov. 1, 1985. '
General Electric Company, NEDO-24583-1, " Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide," Oct. 1979.
MPR Associates, MPR-751, " Augmented Class 2/3 Fatigue Evaluation Method and Results for Typical Torus Attached and SRV Piping Systems," Nov. 1982.
Musolf, D. . Northern States Power Company, letter to H. Denton, NRC, "Addi-tional Information Related to Computer Program CM00F," Feb. 25, 1985.
Novak, T. M. , NRC, letter to J. F. Carolan, ISMG, " Transmittal of NRC Staf f Report on Setpoint Methodology for General Electric Supplied Protection System Instrumentation," May 15, 1984.
Nutech Engineers, Inc.. BPC-01-300-1 through -6, " Hope Creek Generating Station -
Plant Unique Analysis Report," Jan. 1984.
U.S. Nuclear Regulatory Commission, Generic Letter 83-24 "TMI Task Action Plan Item I.G.1, 'Special Low Power Testing and Training,' Recommendations for BWRs,"
June 29, 1983.
-- , NUREG-0408, lark I Containment Short Term Program," Dec. 1977.
-- , NUREG-0484, " Methodology for Combining Dynamic Responses; NRC Staff 'Vorking Group Report," Sept. 1978.
Hope Creek SSER 4 1 Appendix B
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-- , NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment,".Nov. 1979.
-- , NUREG-0661, " Mark I Containment Long-Term Program," July 1980.
, -- , NUREG-0776, " Safety Evaluation Report Related to the Operation of Susque-p hanna Steam Electric Station, Units 1 and 2," Apr. 1981.
-- , NUREG-0800 (formerly NUREG-75/087), " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," July 1981 (includes branch technical positions).
- -- , NUREG-0803, " Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping," Aug. 1981.
3 -- , Office of Inspection and Enforcement, IE Bulletin 79-27, " Loss of
- Non-Class 1E Instrumentation and Control Power System Bus During Operation,"
, Nov. 30, 1979.
i Youngblood, B. J. , NRC, letter to J. F. Carolan, ISMG, " Acceptance of Action i
Plan To Answer NRC Staff Concerns on Setpoint Methodology for General Electric j Supplied Protection System Instrumentation," July 23, 1984.
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Hope Creek SSER 4 2 Appendix B
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APPENDIX D ACRONYMS AND INITIALISMS
, ADS automatic depressurization system ASME American Society of Mechanical Engineers I
BNL Brookhaven National Laboratory BOP balance of plant BTP branch technical position BWR boiling-water reactor CFR Code of Federal Regulations DCP design change package ESF engineered safety feature i
EWE extreme wind event j
4 FRC Franklin Research Center FSAR Final Safety Analysis Report GDC general design criterion (a) i GE General Electric
! HCU hydraulic control unit 4
HPCI high pressure coolant injection i IDVP independent design and verification program IE Office of Inspection and Enforcement
) IEEE Institute of Electrical and Electronics Engineers i
INEL Idaho National Engineering Laboratory
- ISMG Instrumentation Setpoint Methodology Group e
IST inservice testing LOCA loss-of-coolant accident i LPCI low pressure coolant injection i
LRG Licensing Review Group MCC motor control center i
l NRC U.S. Nuclear Regulatory Commission NSSS nuclear steam supply system 1
NTSP nominal trip setpoint PCIGS primary containment instrument gas system
- PCP process control program i PIS pressure-indicating switch Hope Creek SSER 4 1 Appendix 0
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, 'PMH probable maximum hurricane PP&L Pennsylvania Power and Light Company r
PSE&G Public Service Electric and Gas Company
, PT pressure transmitter l PUAR Plant Unique Analysis Report RG- regulatory guide RHR residual heat removal RMS radiation monitoring system RPS reactor protection system RPV reactor pressure vessel RSP remote shutdown panel i SB0 station blackout SER Safety Evaluation Report SLC standby liquid control i
2 SPDS safety parameter display system SQRT Seismic Qualification Review Team SRV safety / relief valve SWP service water pump TMI Three Mile Island J
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i Hope Creek SSER 4 2 Appendix 0
APPENDIX E PRINCIPAL STAFF CONTRIBUTORS AND CONSULTANTS This supplement to the Safety Evaluation Report is a product of the NRC staff and its consultants. The NRC staff members listed below were principal con-tributors to this report.
NRC STAFF Name Title Branch R. Becker Nuclear Systems Engineer Procedures and Systems Review R. Fell Nuclear Engineer Meteorology ar.d Effluents Treatment C. Ferrell Site Analyst Siting Analysis .
A. Lee Senior Mechanical Engineer Equipment Qualifications J. Lombardo Mechanical Engineer Equipment Qualifications J. Mauck Reactor Engineer Instrumentation and Control (Instrumentation) Systems
- 0. Rothberg Mechanical Engineer Mechanical Engineering H. Shaw Mechanical Engineer Mechanical Engineering J. Wilson Nuclear Engineer Auxiliary Systems i
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l l Hope Creek SSER 4 1 Appendix E