ML20199F532
| ML20199F532 | |
| Person / Time | |
|---|---|
| Site: | Seabrook, 05000000 |
| Issue date: | 06/10/1985 |
| From: | Durr J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Baer R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| Shared Package | |
| ML20151L176 | List: |
| References | |
| FOIA-86-266 NUDOCS 8604030319 | |
| Download: ML20199F532 (2) | |
Text
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scarg UNITED STATES
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NUCLEAR REEULATCRY COMMISSION c
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.. sj JUN 101985 MEMORANDUM FOR:
Robert L. Baer, Chief, Engineering and Generic Communication Branch, IE FROM:
Jacque P. Durr, Chief, Engineering Branch, Division of Reactor Safety, RI
SUBJECT:
POTENTIAL GENERIC ISSUE CONCERNING RING SETTINGS OF CROSBY SAFETY VALVES Our memorandun to you dated March 6,1985, Enclosure 1, advised you of a poten-tial generic issue concerning ring settings which produce only 50". lift on Crosby Steam Generator Safety Valves (SG-SVs).
A 10 CFR 50.55(e) Construction Deficiency report was issued by Public Service of New Hampshire (PSNH) advising cf the ring setting problem at Seabrook Station.
Our major concern was not with the installation at Seabrook, a plant under construction, but that opera-ting plants ceuld have SG-SVs that may not achieve full lift; whereby, actual flow capacities and/or accumulation pressure are not representative of analyses.
We have held several phone conversations with both PSNH and Crosby to determine if the observed lift deficiences are common to this model of valve and possible effects on other plants.
To date, we have been unsuccessful in establishing this fact.
PSNH sent us a copy of the Wyle test data, Enclosure 2, which we form eded to MEB (NRR).
PSNH is planning additional tests to resolve their specific installation problem.
In our conversations with Crosby, we were' ol they were awaiting a purchase order to perform testing.
Regarding Crosby' response, we believe there should be a stronger commitment and confirmation that their factory guide ring setting of +150 is correct if such is the case.
Based on our review of the PSNH/Wyle test data of the Crosby 6R10 SV (this valve is in the' "R" orifice size category), it is apparent that design lif t is not achieved with the factory ring settings. We call to your attention that the "R" orifice is at the upper end of commercial valve sizes and has a nozzle bore dia. of 4.513 in. (16.00 sq. in. area).
Generally, the capacity certift-5 cation and functional tests required by ASME Section III are performed by the 9
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valve coefficient of discharge method where the manufacturers testing is done on much smaller valves and the design is then extrapolated to the larger size valves.
This raises the concern that functional demonstration of these large SVs may never have been performed.
Additionally, the problem may extend to other SG-SV suppliers.
We are recommending that NRC Headquarters advise licensee's of the problem through an IE Information Notice, Bulletin or Generic Letter and request licensee's to show that their SG-SVs have full lift.
H. Gregg of my staff is available to provide assistance or additional infor-mation at FTS 488-1295.
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Jacque P. Ourr, Chief V
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Enclosures:
As Stated
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S. Ebneter, Director, DRS S. Collins, DRP R. Gallo, DRP A. Cerne, SRI F. Cherny, NRR H. Gregg, EB 9
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e UNITED STATES
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E WASHINGTON. D. C. 20555 S, -
SEP 0 91985 6
i MEMORANDUM FOR:
Edward L. Jordan, Director Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement FROM:
Hugh L. Thompson, Jr., Director Division of Licensing Office of Nuclear Reactor Regulation
SUBJECT:
IE NOTICE REGARDING MAIN STEAM SAFETY VALVE TEST FAILURES AT WYLE LABORATORIES The purpose of this memorandum is to transmit a proposed IE Information Notice.
which describes failures of Crosby PWR Main Steam Safety Valves (MSSV) to attain full disc lift during full flow tests performed at Wyle Laboratories for Public Sarvice Company of New Hampshire's (PSNH) Seabrook plant. The tests showed that the disc only traveled 50% of the required distance. A lift of 50% in these valves translates to a flow capacity which is approximately 50% of rated capacity at rated pressure.
The test failures were orig (inally reported by PSNH as a potential design deficiency per 10 CFR 50.55 e). The test reports have been under review by NRR Division of Engineering per the February 5,1985 request of your Engineering and Generic Comunications Branch Chief. Our review, which included discussions with both Crosby and Wyle Laboratories, has been completed. Based on this review, the Mechanical Engineering Branch agrees with the PSNH evaluation which concluded that the vendor specified, factory set guide ring position was too high, resulting in an inadequate lift force on the valve disc. They believe the problem may exist in operating plant MSSVs due to similarities among vendors in valve design and methods for detemining ring settings. A preliminary evaluation of the consequences of a 50% degradation in MSSV flow capacity indicates that such a degradation may likely result in overpressurization of the main steam system in some. plants should a full load rejection event occur with the steam dump and bypass system and anticipatory reactor trip unavailable.
The pressure transient could be more severe in B&W designed units due to the relatively small liquid inventory in the once-through steam generator design.
t CONTACT:
Y M. Caruso, NRR P
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SEP 0 91955 Edward L. Jordan.
The enclosed draft Information Notice describes the Crosby valve t'est failures and the findings from the Division of Engineering's review of those tests.
In light of the significance of this potential equipment deficiency, we recomend that you issue the proposed IE Notice.
NRR will continue to investigate the generic implications and safety significance of this equipment deficiency to determine if additional staff action is required.
-We have discussed this proposed IEN with R. Oller and Bob Baer of your office.
cfrd.
/Hugh L@The p n, Jr., Director P Division of Licensing Office of Nuclear Reactor Regulation,
Enclosure:
Draft IE Notice cc:
M. Wegner IE L. Marsh J. Durr, RI R. Baer, IE J. Knight F. Cherny G. Hamer V. Nerses H. Nicolaras R. Oller J. Stolz G. Knighton O
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o UNITED STATES NUCLEAR REGULATORY COMISSION 0FFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 SEPTEMBER
, 1985 IE Information Notice No.: 85-xx MAIN STEAM SAFETY VALVE TEST FAILURES AND RING SETTING ADJUSTMENTS Addressees:
All PWR nuclear power reactor facilities holding an operating license (0L) or a construction pemit (CP)
Purpose:
~
This Information Notice is being provided as_a notification of a potentially significant problem pertaining to spring-actuated main steam safety valves (See Figure 1), that may possess less than the full rated flow capacity required for overpressure protection of the secondary cooling system in 1
PWRs. 'It is expected that recipients will review the information for applicability to their facilities and consider actions, if appropriate, to preclude a similar problem at their facilities. However, suggestions contained in this Information Notice do not constitute NRC requirements; therefore, no specific action or written response"is required.
Description of Circumstances:
Between October 16, 1984, and December 1, 1984 Wyle Laboratories conducted several full flow steam tests on two separate main steam safety valves (MSSVs) j manufactured by Crosby Valve and Gage Company. These Crosby 6R10 MSSVs are to, be installed by Public Service of New Hampshire on the Seabrook main steam system. The tests were conducted in order to determine the adequacy of 1
various MSSV discharge piping arrangements. During the tests the valves were instrumented to measure valve disk lift. The valves were installed on the test facility with the settings of the valve adjusting rings (see Figure 1) as received from the valve vendor. With these ring settings the valve achieved about 50% of the full disk lift required to develop full steam flow capacity within the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code required 3% accumulated overpressure limit. Adequate lift was not achieved for either valve with these factory) adjusted ring settings, even for the largest diameter (least flow resistance vent pipe tested. The guide ring of both valves was subsequently adjusted to a lower position by a significant amount (150 notches) during the course of testing 'and full disk lift was subsequently achieved.
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, 1985 Page 2 of 6
These types of full flow tests are not normally performed by either reactor owners or the valve vendor on MSSVs, nor are such tests required for capacity certification according to the ASME Code Section III.
In general, these valves are capacity certified by tests on much smaller size valves, and the capacities then extrapolated to larger size valves. The MSSVs on most PWRs, while not necessarily the same model or supplied by the same vendor, are like those at Seabrook in that they are generally at the upper end of the valve size range. This raises the concern that full flow functional demonstration of some valve types' may never have been performed, and that due to incorrect ring settings, the valve may not be capable of providing relief capacity in accordance with facility design requirements.
Based on the full flow tests performed at Wyle Laboratories, Public Service Company of New Hampshire (PSCNH) has concluded that the guide ring setting for the Seabrook MSSVs should be adjusted downward 150 notches to ensure full flow. capacity. The MSSVs will be installed at Seabrook with the guide ring adjusted downward 150 notches from the as delivered, factory adjusted setting.
No specific action or written response is required by this Information Notice.
If you have any questions about this matter, please contact the Regional Administrator of the appropriate regional office or this office.
Edward L. Jordan, Director Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement Technical
Contact:
G. Hammer, NRR 301-492-8963 O
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ggpy g MEMORANDUM FOR: Gary Holahan, Chief Operating Reactors Assessment Branch Division of Licensing FROM:
Frank C. Cherny, Acting Chief Mechanical Engineering Branch Division of Engineering
SUBJECT:
MEB C0pmENT ON DRAFT NUREG-0844 "NRC INTEGRATED PROGRAM FOR THE RESOLUTION OF USI A-3, A-4.iND A-5 REGARDING STEAM GENERATOR TUBE INTEGRITY 1
In the subject draft NUREG, there is information regarding the probability of a main steam safety valve (MSSV) sticking open after the valve has been challenged with water resulting from an overfilled stea generator. The estimatedprobabilitygivenforthiseventis3.0X10~g/ demand.
It is further stated that this probability is largely based on the results of liquid testing of pressurizer safety valves performed by EPRI in response to NUREG-0737 Item II.D.1.
However, based on statements in the draft NUREG, we feel obliged to point out that the EPRI data has been misinterpreted in the NUREG. Our review of this data indicates that the probability of such valves sticking open or leaking severely after relieving is probably much higher. Of the approximately 26 safety valve water tests performed by EPRI, about 1/3 resulted in disk to seat chatter which either caused severe internal damage to the valves or was terminated within a very short time (few seconds) by manually opening the valves in order to prevent severe valve damage. Therefore, this data alone, we believe, would seem to conservatively indicate a failure probability of about 0.3.
However, the actual probability for a specific event would be dependent on the liquid temperature, the flow rate, and the duration of flow.-
Some of the EPRI water tests were conducted with saturated liquid and much better performance was observed (mild fluttering instead of chatter) for these than for more subcooled liquid tests. 'Ine information we have indicates that safety valves, originally designed for steam service, perform better for saturated liquid because the water flashes to steam directly under the valve disk thus assisting the disk to lift. Conversely, such safety valves are likely to chatter at more subcooled liquid conditions because less flashing takes place under the disk.
The flow rates during the EPRI water tests were considerably less than the full flow the valves would be capable of. This is clearly sh'own in the data since i
the measured disk lifts were usually significantly less than maximum.
However, the disk to seat chattering was still severe in some tests and we suspect that as full liquid flow conditions are approached the chattering could become yet more severe since greater disk travel would occur. Also, the EPRI tests were conducted for short time periods (a few seconds) especially I
for those tests which were terminated because of chattering.
Longer periods of liquid discharge would result in more disk and seat damage.
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l Gary Holahan
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In sumary, it is difficult to precisely characterize the expected damage to a MSSV for a given water flow event and to acchrately quantify the probability the probability is much higher than the 3 X 10'gver, it is our opinion that of a stuck open or severely leaking valve. How value given in the draft NUREG. At the very least, we recasnend that the final NUREG not convey the impression that the EPRI test results show that spring safety valves designed for steam work well on water. The exact opposite is the case.
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J. Knight R. Bosnak D. Crutchfield T. Marsh E g urphy LAi. Hamer T. Speis B. Sheron i
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SSINS No.: 6835 IN 86-05
(( Ikb UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF~ INSPECTION AND ENFORCEMENT r
WASHINGTON, D.C.
120555
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January 31, 1986 1
IE INFORMATION NOTICE NO. 86-05:
MAIN STEAM SAFETY VALVE TEST FAILURES AND RING SETTING ADJUSTMENTS i
Addressees:
All pressurized water-reactor (PWR) facilities holding an operating license (OL) or a construction permit (CP).
Purpose:
This notice is being provided to alert recipients of a potentially significant problem pertaining to spring-actuated main steam safety valves.that may possess less than the full rated flow capacity.
It is expected that recipients will review the information for applicability to their facilities and consider actions, if appropriate, to preclude a similar problem at their facilities.
However, suggestions contained in this information notice do not constitute NRC requirements; therefore, no specific action or written response is required.
NRC is continuing to obtain and evaluate pertinent information.
If specific actions are determined to be required by NRC, an additional notification will be made.
Description of Circumstances:
In the fall of 1984, Public Service of New Hampshire sent the main steam safety valves (MSSVs) for its Seabrook plant to W' le Laboratories for full-flow y
testing to determine the proper vent stack size.
To determine full flow, Wyle measured disc travel of the model -number 6R10 valves manufactured by the Crosby Valve and Gage Company.
The results of the tests indicated that the valves could not achieve the required disc travel with the factory-set ring setting I
(+155 notches).
The disc travel achieved was 50% of the full lift necessary to i
develop required steam flow capacity.
Adequate lift was not attainable even with the largest diameter tailpipe.
Additional tests were done in July 1985 to determine the appro'priateness of the ring settings.
Specifically, the tests were to determine if the "as-shipped" ring settings of the valves would allow the required disc-travel with minimum tailpipe backpressure and to determine the effects on valve disc travel.for a j.
range of backpressures between 180 and 390 psig.
During these tests, the upper (guide) ring setting was adjusted from +155 to 0 and then to +25 to achieve the required disc travel.
This is a substantial adjustment.
Subsequently, the i
8601290054
o IN 86-05
_ January 31, 1986 j
Page 2 of 2 licensee consulted with the valve manufacturer and agreed on ring settings of
+25 for the guide ring and -25 (the original setting) for the lower (nozzle) ring (see figure 1).
Full flow, full size tests of.the sort described in this notice are not normal-ly performed by the licensee or valve vendor for large secondary safety valves, nor are they required by the ASME Code,Section III.
Instead the valves are certified by extrapolations on data from tests of smaller valves.
The MSSVs on most PWRs, while not necessarily the same model or manufacturer as those at Seabrook, are generally at the upper end of the valve size range.
This raises the concern that full-sized flow demonstration may never have been performed for many MSSVs and these may have incorrect ring settings.
In addition, similar problems with ring settings have been found when full-size tests were performed for PWR primary safety valves.
Thus, these MSSVs may not be capable of providing full-relief capacity in accordance with facility design requirements.
NRC is continuing to obtain and evaluate pertinent information.
If specific actions are determined to be required by NRC, an additional notification will be made.
No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the Regional Administrator of the appropriate regional office or this office.
dwar
. Jordan, Director Divis n of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement Technical
Contact:
Mary S. Wegner
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(301) 492-4511 Attachments:
1.
Figure 1, Typical Main Steam Safety Valve 2.
List of Recently Issued IE Information Notices S
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a WASHINGTON, D. C. 20655 SEP 2 7 25 MEMORANDUM FOR: Dennis Crutchfield, Assistant Director for Safety Ass 2ssment Division of Licensing FROM:
Robert J. Bosnak, Acting Assistant Director for Components and Structures Engineering Division of Engineering
SUBJECT:
JULY 23, 1985 OPERATING REACTOR EVENTS BRIEFING OPEN ITEM - INADEQUATE MAIN STEAM SAFETY VALVE CAPACITY AT SEABROOK At the Operating Reactor Events briefing on the subject issue on July 23, 1985 DE was asked to investigate the adequacy of testing of PWR main steam safety-valves (MSSVs) and the validity of extrapolating test data from small valves to-larger full size MSSVs. The Mechanical Engineering Branch (MEB) has planned the following tasks which we think adequately respond to this request:
(1) MEB has prepared a proposed IE Information Notice to advise the industry of the MSSV capacity problem as it relates to proper ring settings. The proposed Notice was transmitted to DL with our August 9, 1985 memorandum from R. Bosnak to D. Crutchfield.
(2) MEB will fomally request DST to prioritize a potential generic issue dealing with MSSV operability problems including that of inadequate flow capacity. The procedure outlined in Office Letter No. 40 will be followed.
l (3) MEB will discuss with the ASME Section III Subgroup on Pressure Relief possible changes to the ASME Section III Code Class 2 safety valve certification requirements. Currently CL. 2 safety valves can be capacity certified based on tests perfonned on prototypical valves much smaller in size and at much lower pressures than are applicable for PWR,, Main Steam Safety Valves. The ring adjustment problem encountered with the Seabrook MSSVs raises one of the same questions that arose during the recent EPRI testing of ASME Section III CL. I pressurizer safety valves. That is, do
.the valve manufacturers have an adequate understanding of how to extrapolate ring adjustments, that affect lift and blowdown, from very small test valves to the very large safety valves used on PWR plants?
Recently changes to the Code safety valve certification procedure, proposed by MEB, to address this concern for CL. I safety valves were accepted.by ASME for incorporation into the Code. The change will require that new CL. 1 safety valve designs be prototypically tested in sizes and-at pressures, temperatures, and flow rates that envelope those that the valve design will be used for in service. __
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Dennis Crutchfield MEB will explore with the Code Committee the feasibility / desirability of making similar changes to the Code CL. 2 safety valve certification requirements.
We believe these actions should adequately resolve the problem of inadequate
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MSSV capacity.
Robert J. B snak, Acting Assistant Director for Components and Structures Engineering Division of Engineering cc:
J. Knight F. Cherny B. Sheron G. Holahan D. Tarnoff R. Baer H. Gregg
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Please see previous concurrence.
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Clearance **
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(PSIG) 3992
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-2 very light / good OK 19t 995 987 3993
+22t
+12 very light / good OK 17t 995 1011 3994 ll3t
-15 very light / good 8t 18t 1005 997 3995 6t
-2 very light /some wear OK 15t 1005 1031 3996
-11t
-3 very light / good I 5t 15t 1025 1020 j
3997 12t
-2 very light / good St 13t 1025 1016 3998 13t heavy / good 10t 14t 1045 1040 3999
-7t
-1 heavy / good 24t 13t 1045 1053 4000 4t
-9 very light / good 21t 14t 995 1033 4001 12t
-1 heavy / good 14t 14t 995 1036 4002 95t
-3 very light / good 10t 14t 1005 1055 4003 12t
-7 heavy / good 13t 13t 1005 1043 j
4004 25t
-3 very light / good 29t 1It 1025 1065 4005 52t
-3 very light / good 20t 15t 1025 1050 1
4006 27t
-1 very light / good 13t 13t 1045 1100 4007 22t
-1 very light / good 18t 12t 1045 1102 1
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- Min. old disk to guide clearance 10t coMin. new disk to guide clearance I5t
- 1. No effect below 0.0625"
- 2. As-Found ring positions affect setpoint by less than 1% and yield 15% or less blowdown.
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Docket No. 50-346 TOLEDO License No. NPF-3 EDISDN Serial No. 1193 JOE WmAMs. JR.
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Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz Operating Reactor Branch No. 4 i
Division of Licensing United States Nuclear Regulatory Commission Washington, D.C.
20555 l
Dear Mr. Stolz:
This letter is in response for information requested during a September 30, 1985 telephone conversation between Mr. G. Hammer (NRC MEB) and members of the Toledo Edison Nuclear Staff. The requested information concerns details regarding Acti~on Plan No. 16 contained in Toledo Edison's Course of Action (Section IV.C.I.1) submitted September 10, 1985 (Serial No. 1182).
Request No. I b
Provide information regarding the required size of the inlet piping to i
the Main Steam Safety Valves (MSSV's) versus the installed inlet piping size.
Response
The Dresser Instruction Manual for Type 3700 Consolidated Safety Val,ves, May 1978 edition, requires the inlet pipe for the MSSV's to have at least I
a six inch bore. This requirement does not appear in the original instruction manual nor on the drawings which were supplied with the valves. Piping for the inlet pipe was specified in the original instal-lation to be six inch schedule 160 which has an inside dimension of 5.189 inches. This dimension has been verified by field measurements. The concern is that the "as built" t.onfiguration does not conform to the National Board configuration which was used for certification pdrposes.
Further evaluation indicates this condition will not impair MSSV perform-ance. The maximum flow restriction occurs in the upper portion of the nozzle. In this area, the flow diameters for Dresser valves with Q and R orifices are respectively 3.750 and 4.515 inches. Furthermore, there has never been any concern raised regarding the valves capability to relieve an overpressure condition. Per Dresser recommendation, flow testing will be performed to ensure that all applicable ASME Code and National' Board requirements are met. This concern is being tracked by Nonconformance Report 85-0117.
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THE TOLEDO EDISDN COMPANY EDISON PLAZA 300 MADtSON AVENUE TOLEDO OHIO 43652
0 Dockat.No. S'0-346 License No. NPF-3 Serial No. 1193 October 18, 1985 Page 2 Request No. 2 Provide information regarding any planned piping modifications resulting from analysis conducted under Action Plan No. 16.
Response
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To ensure that the main stesalines properly support the MSSV's, analyses are underway to examine the flexibility of the Main Steam System piping.
Fluid transient loads and structural system response will be included in the analysis. A modification is being considered, as an interia measure, to add a restraint to the stenaline for Steam Generator No.1-2.
The purpose of this restraint is to increase the stiffness of the main steam line.
Request No. 3 Provide the testing history for the Main Steam Safety Valves (MSSV's),
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Atmospheric Vent Valves (AVV's), and Main Steam Isolation Valves (MSIV's)..
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Response
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There have been several occurrences of MSSV blowdown in excess of rated 3%. Prior to the 1984 refueling outage, resent pressure experienced during recent plant trips ranged from approximately 980 to 900 psig.
In March 1984, the A4 MSSV stuck open after a reactor trip resulting in boiling dry Steam Generator 1-2.
The root cause of this occurrence was failure of a cotter pin permitting the release nut to travel unrestricted down the spindle threads. Maintenance Procedure MP 1401.28, Main Steam l
Safety Valve Disassembly Inspection / Repair and Reassembly, has been revised to require installation of new stainless steel cotter pins when maintenance is performed on a MSSV.
Previous maintenance experience on MSSV's has shown:
a.
Excessive wear of guides and holders b.
Bending of spindles c.
Damage to the disc seats requiring replacement d.
Greater maintenance requirements for the low set pressure MSSV's All MSSV's on the No. 2 ("A") header were rebuilt during the 1984 refueling outage. Valve B2 was rebuilt in. March 1985. Valves B1 and B7 were rebuilt in 1983, and the other valves on the B header were last rebuilt in 1982.
Four of the eighteen installed MSSV's have a smaller capacity ("Q" orifice).
These valves have required considerably less maintenance than the large
("R") orifice MSSV's.
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8 D$ck2tNs.50-346 License No. NPF-3 Serial No. 1193 October 18, 1985 Page 3 Both AVV's were rebuilt during the 1984/1985 refueling outage.
Recent maintenance and testing histories of the MSSV's, AVV's and Integrated Control System were reviewed. Nothing of significance was noted.
Additional Inservice Test (IST) histories for the MSSV's, AVV's and MSIV's were previously been made available for NRC Nechanical Engineering Branch review.
Request No. 4 Provide the results of the "as received" testing of the first eight NSSV's which were tested at Wyle Laboratory.
Response
The Wyle Laboratory test results for the first eight NSSV's tested in the "as received" condition is provided as an Attachment to this letter.
s The first part of this testing was witnessed by NRC personnel from the Vendor Programs Branch.
Very truly yours, W N' mrt; JW:LLH:DJS:lah attachments DB-1 NRC Resident Inspector cc:
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"g*g' Wyle Test Data Valve Ge4 Position Pa rameter Run 1 Run 2 Run 3 Run 4 Run 5 Run 6 Leakene Disposition Disc lift 1.16 0.20 1.16 1.18 Al Setpoint 1024 995 993 1009 No Leakage Rebuild j
Blowdown 3.8 NM 1.1 2.8 Disc lift 1.06 1.06 1.04 A2 Setpoint 1020 1025 1015 Yes Rebuild Blowdown 4.3 19.0 2.7 j
Disc lift 1.13 1.14
( 0.25 1.14 i
A3 Setpoint 1025 1020 1031 1025 i
No Leakage Rebuild Blowdown 1.3 2.5 NM 1.8 Disc lift 0.62-1.12 1.12 1.12 A4 Setpoint 1014 1015 1006 1005 No Leakage Rebuild i
Blowdown NN 5.3 5.7 5.2 Disc lift 1.15 1.14 1.16 NH 1.14 B1 Setpoint 1042 1041 1044 1044 1040 Yes Rebuild
, Blowdown 4.6 4.5 5.2 5.2 19.7 Disc lift 1.12 1.09 1.10 1.09*
1.10 1.10 32 Setpoint 1051 1037 1037 1042 1043 1049 No Leakage Passed Blowdown 4.8
-3.2 2.7 2.6 2.8 3.1 Disc lift 1.16 1.20 1.20 1.2C 33 Setpoint 1074 1070 1077 1070 Yes Rebuild Blowdown 1.8 2.8 2.4 3.4 Disc lift 1.20 1.20 1.18 l
B4 Setpoint 1080 1071 1069 Yes Rebuild Blowdop 3.3 2.5 2.3
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All disc lifts in inches All setpoints in psi l
All blowdown in percent 196 Indicates the parameter was not measured j
- Compression screw was adjusted one flat clockwise prior to this run.
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NUCLEAR REGULATORY COMMISSION WASHING TON. D. C. 20555
.',Y..*'
AUG 2 1985 MEMORANDUM FOR:
Hugh L. Thompson, Jr., Director Division of Licensing FROM:
Dennis M. Crutchfield, Assistant Director for Safety Assessment, DL
SUBJECT:
SUMMARY
OF THE OPERATING REACTORS EVENTS MEETING ON JULY 23, 1985 - MEETING 85-12 On July 23, 1985, an Operating Reactor Events meeting (85-11) was held to brief the Office Director., the Division Directors and their representatives on events which occurred since our last meeting on July 1,1985. The list of attendees is included as Enclosure 1.
The events discussed and the significant elements of these events are presented in Enclosure 2.
In addition, the assignment of follow-up review responsibility was discussed.
The assignments made during this meeting and the status of previous assignments are presented in Enclosure 3.
Completion dates have been assigned for items in Enclosure 3.
Note that we have revised Enclosure 3 to facilitate computerized tracking and provide additional details regarding responsibilities and status. Each assignee should review with regard to their respective responsibilities and advise OBAB if the target completion date cannot be met.
If an assignee has any questions, please contact D. Tarnoff, x29526.
ADennisM.$.
Cru ie id ss tant Director for Safety Assessme t, DL
Enclosures:
As stated cc w/ enc 1:
See next page e,}
a h
d st
AUG 2 1985 Hugh L. Thompson, Jr. cc:
H. Denton E. Rossi, IE R. Bernero R. Hernan J. Knight F. Schroeder T. Speis G. Knighton C. Heltemes D. Silver T. Novak J. Lyons W. Russell D. Brinkman J. Taylor E. Weiss E. Jordan R. Baer F. Rowsome J. Stolz W. Minners E. Butcher L. Shao R. Bosnak T. Ippolito P. Morriette S. Varga W. Jones J. Zwolinski G. Haumsc E. Sullivan D. Osborne D. Beckham R. Caruso G. Edison D. Lynch K. Seyfrit D. Mcdonald' T. Murley, R-I D. Neighbors J. Nelson Grace, R-II V. Nerses J. Kepper, R-III H. Nicolaras R. D. Martin, R-IV J. Wilson J. B. Martin, R-V P. O'Connor R. Starostecki, R-I T. Alexion R. Walker, R-II K. Jabbour C. Norelius, R-III L. 01shan R. Denise, R-IV H. Booher 5
D. Kirsch, R-V B. Sheron G. Lainas F. Cherney Baranowski, RES
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- e.
ENCLOSURE 1 LIST OF ATTENDEES OPERATING REACTORS EVENTS BRIEFING (85-11)
JULY 23, 1985 i
H. Denton, NRR J. Stolz, NRR/DL/0RAB#4 M.-Caruso,NiiR D. Neighbors, NRR/DL/0RB#1 J. Jackson, MRR/DE/EQB G. Knighton, NRR/DL/LB#3 l
i B. Jones, IE/DEPCR/EAB H. kood, NRC/DL/LBF3 F. Cherny, MEB/DE/KRR J. Lyons, NRR/DL W. Swenson, ORAB/DLINRR R. Bernero, NRR/DSI H. Nicolaras, NRR/DL/0RB#4 T. M. Novak, NRR/DL K. Mitchell, NRR/DL/0RAB/ ORB #5 D. Beckham, NRR/DHFS J. Stone, IE/VPB -
R. Wessman, NRR/DL G. Bagchi, NRR/DE/EQB T. Speis. NRR/ DST J. Wilson, LB#3/DL/NRR F. Senroeder, NRR/ DST T. Alexiorf, LB#1/DL/NRR W. Minners, NRR/ DST L. N. 01shan, LB#1/DL/NRR S. Varga, DL P. O'Connor, LB#1/DL/NRR D. Crutchfield, DL W. J. Collins, IE/DEPER J. P. Knight, NRR/DE N. P. Kadambi, NRR/DL/LB#3 G. Lanik, IE/EAB D. Humenansky, OCM/COMM ZECH S. Schwartz, IE/DEPER J
T. Rotella, NRR/DL/0RBf5 H. Thompson, NRR/DL B. Sheron, NRR/DSI/RSB P. Morriette, NRR/DL D. Tarnoff, NRR/DL B. Bosnak, NRR/DE A. W. Dromerick, IE/DEPER/EGCB K. Seyfrit AEOD/ROAB E. Weiss, IE/DEPER/EAB D. Zukor AE00/R0AB N. Lauben, NRR/DSI/RSB E. J. Brown, AEOD/ROAB i
SEAPROOK - CROSBY MAIN STEAM SAFETY VALVE FLOW DEFICIENCY - DECEMBER 1984 (G
HAMMER, NRR)
PROBLEM - FULL FLOW TEST RESULTS INDICATE SPRING-ACTUATED MAIN STEAM SAFETY VALVES MAY NOT ACHIEVE RATED FLOW CAPACITY.
SAFETY SIGNIFICANCE - POSSIBLE INADEQUATE OVERPRESSURE PROTECTION OF SECONDARY SYSTEM IN PWRs USING THESE VALVES WYLE LAB TEST RESULTS: INADEQUATE LIFT OF VALVE DISK (ABOUT 50%) WITH THE VENDOR (CROSBY) RECOMMENDED RING
~
SETTING ADJUSTMENTS.
TESTS WERE CONDUCTED TO DETERMINE ADEQUACY OF DISCHARGE PIPING.
CORRECTIVE ACTION - RINGS READJUSTED.OBTAINED FULL l
LIFT ON SEABROOK VALVES GENERIC IMPLICATION - SEABROOK VALVES AND DISCHARGE PIPING SIMILAR TO OTHER PWRs.
FULL FLOW TESTS NOT NORMALLY RUN TO ADJUST RINGS.
NRC FOLLOWUP ACTION:
(1) DEVELOPING IE INFORMATION NOTICE (2) STAFF MAY PURSUE AS A GENERIC ISSUE (3) DISCUSSIONS WITH CROSBY BY REGI0N 1 AND NRR REGARDING ADEQUACY OF VENDOR GUIDANCE AND SRV
~
RING SETTINGS.
l i!
I
Page No. 4 Report No. 47447-0 VENT STACK (P
T )
- DISCHARGE S
S TEMPERATURE & PRESSURE l
TAILPIPE INLET TEMPERATURE & PRESSURE (P, T )
I 2
2
~ 7 VENT STACK ELB0W s
j TEMPERATURE & PRESSURE IP'T) 4 4
<5 TAILPIPE DISCHARGE s
TEMPERATURE & PRESSURE I'3, T )
3
/
Y INLET STEAM TEMPERATURE & PRESSURE (P, T )
j j
U FIGURE 1.
INSTRUMENTATION LOCATIONS (16,18, AND 20-INCH VENT STACKS)
I' i
OCONEE 2 - EXTENDED BLOWDOWN FROM MAIN STEAM SAFETY VALVES
\\
JULY 11, 1985 (H, NICOLARAS, NRR)
OCONEE UNIT 2 REACTOR TRIP FROM 94% POWER CAUSED BY PERSONNEL ERROR TWO MAIN STEAM SAFETY VALVES DID NOT RESEAT AT SETPOINT -
EXTENDED BLOWDOWN - TO ABOUT 990 PSI TO RESEAT VALVES, OPERATORS REDUCED STEAM PRESSURE THROUGH TURBINE BYPASS VALVES.
FAILURE OF CROSBY MAIN STEAM SAFETY VALVES TO PROPERLY RESEAT HAS ALSO OCCURRED REPEATEDLY AT OCONEE UNIT 1 IMPROPER RING SETTING IS A LIKELY CAUSE OF EXCESS BLOWDOWN, BUT NOT CONFIRMED, DUKE POWER COMMITTED ~ CORRECTIVE ACTIONS TO REGION II
SUMMARY
OF PLANTS REPORTING SIMILAR BLOWDOWN PROBLEM PLANT KNOWN # OF EVENTS OCONEE 1, 2, 3 31 TROJAN 1
i SALEM 1
i 1
l
e d
CAUSE: lNEN POSTULATED PROBLEM WITH RING SETTINGS CORRECTIVE ACTIONS:
READJUST, FESET SETPO!fC, VALVE DISASS9BLY Li (0UT OF 16) VALVES REWORKED DURING EAG REFUELING LICENSEE PURSUING ETHODS FOR GECKING BLOWDOWN St ilNCG T
IPPLICATIONS: RCS OVERC00 LING CHALLENGE TO ADDITIONAL SYSTEMS POTENTIAL PRECURSOR TO STUCK OPEN VALVE OR FULL LIFT CAPACITY e
M e
B e
e e
O 4
i e
gaN;.
3.
J/2T/85 OPDATING REA: TORS' EVENTS REETINS FOLLOMUP ITEMS AS OF REETINS B5-12 DN JULY 23,1985 (ORDERD BY AS ENDING MEETING EATES, NSSS VEN00RS, FACILITY)
- ETINE FA !LITY RESPONS!!LE TASK DESCRIPTION SCHEDULE CLDSD DATE COMMEk'S m!ER' NSS$ VENDOR /
DIVIS!CN!
COMPLET. IT DOCUMENT /
- E!!'J EVD1 DESCRIP.
INDIVIDUAL DATEtS)
IEff!NS,ETC.
LIE l-12 SEA 3RDOK DS! IRARSH ANALY!E SAFETY IMPLICATIONS OF 09/30/85 OPEN
//
'/23/25 N / CROSBY RAIN
/
VALVE FLOM DEFICID:Y
/ /
STEAMSAFETT
//
VALVE FLOM DEFICIENCY 12/14 5-12 SEAIR00K CE / CHERNEY
! WEST!6 ATE ADEGUA Y OF TESTING 09/30/85 DPEN
//
r/23/85 N / CROSIT MAIN
/
AND VALIDITT OF EITRAPOLATING
/ /
STEAMSAFETT DATA FROM SMAR TO LARSE VALVES //
VALVE FLOW DEFICIENCY 12/84
~
l-12 CTCH I ORAL /CARUSD,R.
N!u DEVELOP TIA TO C00Rt!NATE 07/30/85 CLOSD 07/23185
(/30/85 SE / STUCK QPEN
/
IE NOTICE AND FURTHER
/ /
TIA IN SAFETT RELIEF INVESTIEATIVE EFFORTS.
//
CONOURRENE VALVE D
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d Seebre:k Stcticn Engineering Office
(_
e,
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E. Brown Projects - WJD A. Cerne Projects - Chrono R. Cummings Ropes 6 Gray (3)
R. DeLoach F. Sabadini OOdM W. Derrickson A. Shepard J. DeVincentis R. Sweeney N;w Hompshire Yankee Division August 27, 1985T Feigenbaum T.F. Q2.2.2 G. Gram G. Thomas SBN-863 W. Hall H. Tracy T.F. Q2.2.2 R. Harrison J. Tribbic D. Hunter UE6C 6 W (SB-19770l W. Johnson M. Wiley G. Kingston ASLB j
United States Nuclear Regulatory Commission G. F. Mcdonald 10CFRSO.SS(e) File Region I M. McKenna J. Allen j
631 Park Avenue B. Middleton INPO i
King of Prussia, PA 19406 D. Moody NRC Subject File Attention:
Mr. Richard W. Starostecki, Director Division of Project and Resident Programa
References:
(a) Construction Permits CPPR-135 and CPPR-136 Docket Nos. 50-443 and 50-444 (b) Telecon of December 21,1984.. A. L. Legendre, Jr.
(YAEC) to J. Grant (Region I)
(c) NHY Letter SBN-751 dated January 17,1985, John Devincentis to R. W. Starosteeki, NRC Region I (d) NHY Letter SBN-788 dated April 8,1985, John DeVincentia to R. W. Starostecki
Subject:
Final 10CFR50.55(e) Rep. ort.." Main Steam Safety Valve Ring Setting Deficiency," (CDR'84-00-19f,
Dear Sir:
In Ref erences (c) and (d), we filed interim 10CFR50.55(e) reports re-garding a ring setting deficiency for the main steam safety valves. The valves were sent to Wyle Laboratories for testing for determination of the proper ring settings. The tests were completed and the results are contained in Wyle Laboratories Report No. 47787-01 dated July 12, 1985.
The objectives of the tests were to:
1.
Determine if the "as-shipped" ring settings of the valves would allow the required disc travel with minimum tailpipe backpressure.
2.
Determine the ef fects on the valve disc travel for a range of backpressures between 180 and 390 psig.
The results of the "as-shipped" ring setting tests indicated that the valves could not achieve the required disc travel with 3% steam accumulation at minimus :ailpipe pressures of 15-20 psig.
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United States Nuclear Regulatory Commission Attention:
Mr. Richard W. Starostecki Page 2 During the test, the upper (guide) ring setting was adjusted f rom +155 notches to 0 and +25 notches, and full required disc travel was achieved at 3% steam accumulation under the full range of tailpipe backpressure tested.
As a result of these test, we have agreed with Crosby, the valve manu-f acturer, that the optimus ring settings for the Seabrook main steam safety valves shall be -25 notches for the lower (nozale) ring (original setting) and +25 notches for the upper (guide) ring. The corrections were completed by a Crosby service representative at the Wyle f acility prior to returning the velves to the Seabrook gito.
This is our final report on this subject.
Very truly your,
n DeVin'centis, Director ngineering and Licensing ec: Atomic Safety and Licensing Board Service List Director, Of fice of Inspection and Enf orcement U. S. Nuclear Regulatory Commission Washington, DC 20555 E
4
4 j
a Donald E. Chick William S. Jordan III Town Manager Dicne Curran Town of Exeter Harmon, Wedse & Jordan 10 Front Street 20001 S. Street, N.W.
Exeter, NH 03833 Suite 430 W:shington, D.C.
20009 Brentwood Board of Selectmen RED Dalton Road j
Robert G. Parlis Office of the Executive Legal Director Brentwood, NH 03833 f
U.S. Nuclear Regulatory Commission Richard E. Sullivan, hayor i
W;ahington, DC 20555 City Ball Newburyport, MA 01950 R2bert A. Backus, Esquire 126 Lowell Street Calvin A. Canney P.O. Box 516 Manchester, NH 03105 City Manager City Hall 126 Daniel Street Philip Ahrena Esquire Assistant Attorney General Portsmouth, NH 03801 Augusta, ME 04333 Dana Bisbee, Esquire Assistant Attorney General Mr. John B. Tanzer Office of the Attorney General D2signated Representative of 208 State house Annex the Town of Hampton Concord, NH 03301 5 Horningside Drive Hampton, NR 03842 Anne Verge, Chairperson Board of Selectmen Roberta C. Pevear Town Hall Designated Representative of the Town of Hampton Falla South Hampton, NH 03827 Drinkwater Road H:apton Falls, NH 03844 Patrick J. McKeon Selectaen's Office 10 Central Road Mrs. Sandra Cavutis Designated Representative of Rye, NH 03870 the Town of Kensington Carole F. Kagan, Esquire RFD 1 Atomic Safety and Licensing Board Panel East Kingston, NH 03827 U.S. Nuclear Regulatory Commission Jo Ann Shotwell, Esquire Washington, DC 20555 Assistant Attorney General Mr. Angi Machiros Environmental Protection Bureau Chairman of the Board of Selectmen i
Department of the Attorney General l
Town of Newbury One Ashburton Place, 19th Floor Boston, MA 02108 Newbury, MA 01950 I
Town Manager's Office Senator Gordon J. Humphrey Town Hall - Friend Street U.S. Senate Washington, DC 20510 Amesbury, MA 01913 (AITN: Tom Burack)
Senator Cordon J. Husphrey 1 Pillsbury Street Diana P. Randall Concord, NH 03301 70 Collins Street Seabrook, NH 03874 (ATIN: Herb Boynton)
d
- 4j y*
s SEABROOK STATION Engin93 ring Office f
y Pubic Service of New HampeNro N;w Hampshire Yankee Division January 17, 1985
~
SBN-751 T.F. Q2.2.2 o
United States Nuclear Regulatory Commission Rogion I 631 Park Avenue King of Prussia, PA 19406 Attention:
Mr. Richard W. Starostecki, Director Division of Project and Resident Programs s
References:
(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) Telecon of December 21, 1984, A. L. Legendre, Jr. (YAEC) to J. Grant (Region I)
M
Subject:
Interra 10CFR50.55(e) Report, " Main Steam Safety Valve Ring Setting Deficiency" Dscr Sir:
We previously reported (Reference (b)] a potential 10CFR50.55(e) item to Region I regarding the apparent deficiency in Crosby Main Steam Safety Valve ring settings.
The deficiency was discovered at Wyle Labs during full flow tasting of the Crosby valves to determine the proper vent stack size.
In crder to determine full flow, the valve disc travel was measured and a disc travel of 1.12 inches was confirmed to us as being required by Crosby Valve Corpany via telecon on December 3, 1984 The tests determined that the disc enly traveled approximately 50% of the required distance with 3% accumulation.
Tho same limited disc travel occurred on later tests with larger diameter vent cercks. Af ter discussions with Crosby, the-ring setting was adjusted." The valve was ratested and the required disc travel was achieved.
The problem and its reportability occurs because uncorrected reduced valve disc travel n'ay invalidate certain assumptions on the Main Steam System in the FSAR. Accident Analysis.
The Seabrook FSAR Accident Analysis performed by Westinghouse Corporation assumes that the Main Steam Safety Valves are fully open with 3% accumulation. The ring settings are adjusted at Crosby Velve Company and the valve disc travel is not normally measured to confira tha required lift in field test.
I n0 P.O. Box 300 Secticci.NH03874
- Telephone (603)474 9521
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United States Nuclear Regulatory Commission Attention:
Mr'. Richard Starostecki, Director Page 2 In conclusion, we are performing additional testing on the Crosby Main Steam Safety Valve to verify the ring setting problem without any stack in-fluence whatsoever, and to determine the new ring settings. The results of these tests will be reported in a future interim 10CFR50.55(e) report by March 31, 19E.
p[
4 Very truly yours, V
W
}
.J" i
J. DeVincentis Director Engineering and Licensing cc: Atomic Safety and Licensing Board Service List
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Director, Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, DC 20555 1
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Pubic Service of New luTMe M w Hampshire Yankee Division January 17, 1985
~
SBN-751 T.F. Q2.2.2 United States Nuclear Regulatory Commission Ragion I 631 Park Avenue King of Prussia, PA 19406 Ac tsn tion:
Mr. Richard W. Starostecki, Director Division of Project and Resident Programs Roforences:
(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) Telecon of December 21, 1984, A. L. Legendre, Jr. (YAEC) to J. Crant (Region I)
Subject:
Intertu 10CFR50.55(e) Report, " Main Steam Safety Valve Ring Setting Deficiency" Dazr Sir:
We previously reported [ Reference (b)] a potential 10.CFR50.55(e) item to Ragion I regarding the apparent deficiency in Crosby Main Steam Safety Valve ring settiegs.
The deficiency was discovered at Wyle Labs during full flow testing of the Crosby valves to determine the proper vent stack size.
In ordar to determine full flow, the valve disc travel was measured and a dise l
travel of 1.12 inches was confirmed to us as being required by Crosby Valve Ccupany via telecon on December 3, 1984 The tests determined that the disc 1
caly traveled approximately 50% of the required distance with 3% accumulation.
Tho same limited disc travel occurred on later tests with larger diameter vent ctccks.
After discussions with Crosby, the ring setting was adjusted.
The volve was retested and the required disc travel was achieved.
The problem and its reportability occurs because uncorrected reduced
,,volve disc travel may invalidate certain assumptions on the Main Steam System in the FSAR Accident Analysis.
The Seabrook FSAR Accident Analysis performed by Westinghouse Corporation assumes that the Main Steam Safety Valves are fully open with 3% accumulation.
The ring settings are adjusted at Crosby Velve Company and the valve disc travel is not normally measured to confirm tha required life in field test.
J
, c. p e"3 Cef
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s
- )L2k P.O. Box 300 Secticok.NH03874 Telephone (603)474 9521 e
h United States Nuclear Regulatory Commission Attention:
Mr. Richard Starostecki, Director Page 2 In conclusion, we are performing additional testing on the Crosby Main Steam Safety Valve to verify the ring setting problem without any stack in-fluence whatsoever, and to determine the new ring settings.
The results of these tests will be reported in a future interim 10CFR50.55(e) report by March 31, 1985.
Very truly yours.
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,4-se "
J. DeVincentis, Director Engineering and Licensing cc:
Atomic Safety and Licensing Board Service List Director, Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, DC 20555 y.,y
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E WASHINGTON, D. C. 20555 x.....j m
MEMORANDUM FOR:
Robert Bosnak, Chief Engineering Issues Branch Division of Safety Review and Oversight FROM:
Frank C. Cherny, Section Leader Section B Engineering Issues Branch
SUBJECT:
TRIP REPORT - MEETING 0F ASME SECTION III SUBGROUP ON PRESSURE RELIEF, MARCH 12 & 13, 1985 The referenced meeting was held at the United Engineering Center in New York City. A synopsis of major items of interest to NRC is contained in the attachment to this memorandum.
Information on all the agenda items is available in my file.
phy b.
Y-lY Frank C. Cherny, Section L der Section B Engineering Issues Branch
'e'
Attachment:
As stated cc:
T. Speis B. Sheron J. Richardson W. Campbell W. Norris G. Millman H. Gregg, RI K. Kniel G. Hanner E. Brown M. Wegner s M. Caruso EIB Members
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2 ASME CODE COMMITTEE: ASME Section III Subgroup on Pressure Relief (SGPR)
DATE OF MEETING: March 12 and 13, 1985 NRC COMMITTEE MEMBER:
F. Cherny MEETING ATTENDED AND REPORTED BY:
F. Cherny I.
ADMINISTRATIVE ITEM As previously reported, F. Catudal, the SGPR Chairman is retiring by mid,
Jr -
1986 and resigning from the Subgroup. This was his last SGPR meeting.
His last Code Committee meeting as SGPR Chairman will be the May S/C III meeting in Baltimore. The writer has received both NRC and ASME approvals.to succeed as SGPR Chairman and will fully assume that position after the Baltimore meeting.
II. TECHNICAL ITEMS A.
Definition of System Operating Condition - NP-3-82 The proposal, prepared by the writer, to delete the " Upset, Emergency, etc. System Condition " terminology from NB-7000 was passed by the MC at the February meeting. The SGPR agreed that the same changes should be incorporated in NC/ND-7000 as part of the changes made pursuant to the
~
ongoing re-review effort (NP-1-85).
B.
Ring Adjustment Problems on ASME CL.2 Main Steam Safety Valves (MSSVS)
NP-1-86 In accordance with agreements reached at the last SGPR meeting, the i
writer submitted a formal proposal at this meeting that would revise NC-7000 to require that each CL.2 MSSV production valve be full flow, full pressure tested by the manufacturer prior to shipment. These tests would confirm that the value adjusting rings were in the correct position l
to assure full stamped relieving capacity and that blowdown was in accordance with the valve specified in the valve Design Specification.
With minor editorial changes the proposal was unanimously passed.
It will be incorporated as part of the revisions made pursuant to the NC-7000 re-review (NP-1-85). As previously reported, Crosby Valve currently has the capability to perform such tests. Target Rock will have the capability in the near future. Dresser has stated that valves.
will have to be shipped to Wyle Laboratory in Huntsville, Alabama. As '
part of the discussion on this item, the writer specifically asked if there was anything less expensive"than testing each valve that could be done to assure that the rings were reliably adjusted properly. None of a
9
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the SGPR members had any alternative suggestions.
Representatives from l
l Crosby, Dresser, and Target Rock were in attendance. Th'e Chairman asked I
what the bases for ring setting has been up to the present time for I
MSSVs. Only the Crosby representative responded, saying they were l
extrapolated based on data from much smaller valves and " field l
experience".
C.
Re-Review of NB/NC/ND-7000 NP-1-85 Again for this meeting the review focused on NB-7000. Significant actions include:
1.
Safety Valves - reaffirming restriction for use to only stem, air, or gas service. Even with the addition of the full size Demonstration of Function Text requirement, SGPR concluded that the Code still should not permit the use of this type of valve for liquid service.
2.
Safety Valves With Aux 11ary Actuating Devices - Revised to permit use of this type of valve for air, gas and liquid service in addition to steam, This is consistent with EPRI and other test data obtained over the last several years.
e a
UNITED STATES NUCLEAR REGULATORY COMMISS10N OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 j
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It is expected that recipients will review the information for applicability to their facilities and consider actions, if appropriate, to preclude a similar problem :::r'.; at their facilities. However, suggestions contained in this information notice do not constitute NRC requirements; therefore, no specific action or written response is required.
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}'s No specific action or written response is required by this information notice.
If you have any questions about this matter, please contact the Regional Administrator of the appropriate regional office or this office.
Edward L. Jordan, Director Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement p
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UNITED STATES
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WASHINGTON, D. C. 20555
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MEMORANDUM FOR:
Hugh L. Thompson, Jr., Director Division of Licensing FROM:
Dennis M. Crutchfield, Assistant Director for Safety Assessment DL
SUBJECT:
SUMMARY
OF THE OPERATING REACTORS EVENTS
~~
~
MEETING ON JULY 23, 1985 - MEETING 85-12 On July 23, 1985, an Operating Reactor Events meeting (85-11) was held to brief the Office Director, the Division Directors and their representatives on events which occurred since our last meeting on July 1,1985. The list of attendees is included as Enclosure 1.
The events discussed and the significant elements of these events are presented in Enclosure 2.
In addition, the assignment of follow-up review responsibility was discussed. The assignments made during this meeting and the status of previous assignments are presented in Enclosure 3.
Completion dates have been assigned for items in Enclosure 3.
Note that we have revised Enclosure 3 to facilitate computerized tracking and p,rovide additional details regarding responsibilities and status. Each assignee should review. with regard to their respective responsibilities and advise ORAB if the target completion date cannot be met.
If an assignee has any questions, please contact D. Tarnoff, x29526.
3 DennsM.jp.yted s
Cru ss tant Director for Safety Assessme t. DL
Enclosures:
As stated I
cc w/ enc 1:
See next page '
l 0
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~
AUG 2 1985 Hugh L. Thompson, Jr. -
cc:
H. Denton E. Rossi, IE R. Bernero R. Hernan J. Knight F. Schroeder T. Spets G. Knighton C. Heltemes D. Silver T. Novak J. Lyons W. Russell D. Brinkman J. Taylor E. Weiss l
E. Jordan R. Baer F. Rowsome J. Stolz W. Minners E. Butcher L. Shao R. Bosnak
.T. Ippolito P. Morriette S. Varga W. Jones J. Zwolinski G. Hammer E. Sullivan D. Osborne D. Beckham R. Caruso G. Edison D. Lynch i
K. Seyfrit D. Mcdonald i
T. Murley, R-I D. Neighbors J. Nelson Grace, R-II V. Nerses J. Kepper, R-III H. Nicolaras R. D. Martin, R-IV J. Wilson J. B. Martin, R-V P. O'Connor R. Starostecki, R-I T. Alexion R. Walker, R-II K. Jabbour C. Norelius, R-III L. 01shan R. Denise, R-IV H. Booher D. Kirsch, R-V B. Sheron G. Lainas F. Cherney Baranowski, RES i
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ENCLOSURE 1 LIST OF ATTENDEES OPERATING REACTORS EVENTS BRIEFING (85-11) s JULY 23, 1985 H. Denton, NRR J. Stolz, NRR/DL/0RABf4 M. Caruso. NRR D. Neighbors, NRR/DL/0RBf1
{
J. Jackson, NRR/DE/EQB G. Knighton, NRR/DL/LBf3 B. Jones, IE/DEPER/EAB H. Rood, NRC/DL/LBf3 F. Cherny, MEB/DE/NRR J. Lyons, NRR/DL i
W. Swenson, ORAB/DL/NRR R. Bernero NRR/DSI 4
i H. Nicolaras, NRR/DL/0RBf4 T. M. Novak, NRR/DL i
K. Mitchell, NRR/DL/0RAB/0RBf5 D. Beckham, NRR/DHFS j
J. Stone, IE/VPB R. Wessman, NRR/DL (fs G. Bagchi, NRR/DE/EQB T. Speis NRR/ DST
.u j
J. Wilson, LBf3/DL/NRR F. Schroeder, NRR/ DST l
T. Alexion LBil/DL/NRR W. Minners, NRR/ DST j
L. N. 01shan, LBil/DL/NRR
- 5. Varga, DL l
P. O'Connor, LBil/DL/NRR D. Crutchfield, DL i
{
W. J. Collins, IE/DEPER J. P. Knight, NRR/DE N. P. Kadambi, NRR/DL/LBf3 G. Lanik, IE/EAB D. Humenansky, OCM/COMM ZECH S. Schwartz IE/DEPER T. Rotella, NRR/DL/0RBf5 H. Thompson, NRR/DL B. Sheron, NRR/DSI/RSB P. Morriette, NRR/DL B. Bosnak, NRR/DE D. Tarnoff NRR/DL A. W. Dromerick, IE/DEPER/EGCB X. Seyfrit. AE0D/ROAB t
j E. Weiss, IE/DEPER/EAB D. Zukor AEOD/ROAB
(*,
N. Lauben, NRR/DSI/RSB E. J. Brown, AEOD/ROAB
= _, _ -. - _. -
j OPERATING REACTORS EVENTS BRIEFING (85-12)
JULY 23, 1985 INDIAN POINT UNIT 3 -
b STEAM GENERATOR WELD INDICATIONS, SEABROOK MAIN STEAM SAFETY VALVE TEST FAILURE OCONEE UNIT 2 EXTENDED BLOW DOWN FROM MAIN STEAM SAFETY VALVES WATERFORD UNIT 3 PLANT TRIPS JULY 4-7, 1985 WATERFORD/WOLFCREEK -
STARTUP EXPERIENCE COMPARISON, CALLAWAY/ CATAWBA / BYRON
<, COMBUSTION ENGINEERING LOCA ANALYSIS ERROR g:f M0JAVE GENERATING REHEAT LINE FAILURE I
STATION PALUEL UNITS 1, 2 t
IN-CORE INSTRUMENTATION TUBE VIBRATION PROBLEMS
{
OTHER EVENTS OF INTEREST MILLSTONE UNIT 2 -
PRESSURIZER SPRAY VALVE FAILURES LASALLE UNIT 1 RHR FLOW SWITCHES IMPROPERLY INSTALLED e
FERMI UNIT 2 INADVERTENT CRITICALITY
("URKEYPOINTUNIT3-REACTOR TRIP AND AFW VALVE FAILURE
INDIAN POINT 3 - SG WELD INDICATIONS JUNE 27, 1985 (D, NEIGHBORS, NRR)
.r' PLANT IN REFUELING STATUS
~
T.S. REQUIRES INSPECTIONS OF SG TRANSITION ZONE UPPER GIRTH WELDS INDICATIONS FOUND BY UT:
SG 31 I
SG 32 - 2
QV ? -
SG-34 HAD WELD REPAIR OF LEAK FOUND IN 1983 3
MT ON SG-34 SHOWED CLEAN ON 16 0F 23 INDICATIONS REMAINING 7 WELD INDICATIONS ON 34, AND 3 ON 31 AND 32 WERE FOUND TO BE CODE ACCEPTABLE BY VIRTUE OF SIZE OR.BY FRA MECHANICS ANALYSIS LICENSEE STILL EVALUATING NRR HAS LEAD (SINCE 7/15/85) 1 l
NRR a IE DEVELOPING GENERIC CORRESPONDENCE
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SERIES 51 STEAM GD ERATOR
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SEABROOK - CROSBY MAIN STEAM SAFETY VALVE FLOW DEFICIENCY - DECEMBER 1984 (G. HAMMER, NRR)
PROBLEM - FULL FLOW TEST RESULTS INDICATE SPRING-ACTUA MAIN STEAM SAFETY VALVES MAY NOT ACHIEVE RATED FLOW
- CAPACITY, SAFETY SIGNIFICANCE - POSSIBLE INADEQUATE OVERPRESSURE PROTECTION OF SECONDARY SYSTEM IN PWRs USING THESE V 4
WYLE LAB TEST RESULTS: INADEQUATE LIFT OF VALVE DISK
\\
(ABOUT 50%) WITH THE VENDOR (CROSBY) RECOMMENDED RING SETTING ADJUSTMENTS. TESTS WERE CONDUCTED TO DETERMINE ADEQUACY OF DISCHARGE PIPING, (u:
CORRECTIVE ACTION - RINGS READJUSTED.
OBTAINED FULL LIFT ON SEABROOK VALVES i
GENERIC IMPLICATION - SEABROOK VALVES AND DISCHARGE i
PIPING SIMILAR TO OTHER PWRs.
FULL FLOW TESTS NOT NORMALLY RUN TO ADJUST RINGS, NRC FOLLOWUP ACTION:-
(1) DEVELOPING IE INFORMATION NOTICE (2) STAFF MAY PURSUE AS A GENERIC ISSUE (3) DISCUSSIONS WITH CROSBY BY REGION 1 AND NRR a
Q^, -
REGARDING ADEQUACY OF VENDOR GUIDANCE AND SRV RING SETTINGS,
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Page No. 4 Report No. 47447-0 s
VENT STACK
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DISCHARGE S
S TEMPERATURE & PRESSURE
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TAILPIPE INLET TEMPERATURE & PRESSURE
(,
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2 VENT STACK EL80W sp j
TEMPERATURE & PRESSURE (P ' T )
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TAILPIPE DISCHARGE TEMPERATURE & PRESSURE IE,T)
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TEMPERATURE & PRESSURE (P, T )
3 3
FIGURE 1.
INSTRUMENTATION LOCATIONS (16,18, AND 20-!NCH YENT STA I
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o OCONEE 2 - EXTENDED BLOWDOWN FROM MAIN STEAM SAFE JULY 11, 1985 (H. NICOLARAS, NRR) j OCONEE UNIT 2 REACTOR TRIP FROM 94% POWER CAUSED BY t
i PERSONNEL ERROR 1
TWO MAIN STEAM SAFETY VALVES DID NOT RESEAT AT SE
. EXTENDED BLOWDOWN - TO ABOUT 990 PSI
~
TO RESEAT VALVES, OPERATORS REDUCED STEAM PRESSURE THROUG i
FAILURE OF CROSBY MAIN STEAM SAFETY VALVES TO P l,
m.
RESEAT HAS ALSO OCCURRED REPEATEDLY AT OCONEE U i!.
IMPROPER RING SETTING IS A LIKELY CAUSE OF EXCESS BLO l
BUT NOT CONFIRMED.
i I
DUKE POWER COMMITTED CORRECTIVE ACTIONS TO RE l
l j
SUMMARY
OF PLANTS REPORTING SIMILAR BLOWDOWN PR i
t PLANT KNOWN # OF EVENTS '
OCONEE 1,'2, 3 31 i
TROJAN 1
I SALEM
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GAljE: LN00N POSTILATED PROBLEM WITH RING SETTIfr6 CORRECTIVE ACTIONS: READJUST, RESET SETP0llR, VALVE DISASS& BLY l
4 (OUT OF 16) VALVES REW) RED DURING EACH IEFUELING LI NSEE PURSUING ETHODS FOR CHECKING BLOWDOWN SETTIhr6 l
IPPLICATIONS: RCSOVERC00 LING CHALLENGE TO ADDITIONAL SYSTUS POTENTIAL PRECURSOR TO STUCK OPEN VALVE OR FULL LIFT CAPACITY O
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l WATERFORD 3 - PLANT TRIPS JULY 4-7, 1985 (J. WILSON, NRR)
WATERFORD 3 EXPERIENCED FOUR REACTOR' TRIPS IN LESS T THREE DAYS b
DURING A PORTION OF THIS TIME, THE EFW TURBINE-DRIVEN PUMP WAS UNAVAILABLE DUE TO BEING INADVERTENTLY TRIPPED JULY 4 AT 0950 HOURS - 100% PWR - LOW SG LEVEL-HIGH VIBRATION ON "A" MAIN FEEDWATER PUMP JULY 4 AT 2217 HOURS - 6% PWR - CPC AUXILIARY TRIP ON AXIAL SHAPE INDEX - XE OSCILLATIONS JULY 5 AT 2219 HOURS - 60% PWR - HIGH SG LEVEL DUE T0 (g,.
OVERFEEDJNGSGWHILEINMANUALCONTROLWITHONEMAIN FEEDWATER PUMP RUNNING JULY 6 AT 0915 HOURS - 70% PWR - EFW PUMP TERRY TUf OVERSPEED LATCH WAS FOUND TO BE TRIPPED JULY 7 AT 0121 HOURS - 90% PWR - LOW SG LEVEL - LO FEEDWATER PUMPS ON LOW SUCTION WHILE AN OPERAT ING TO BACKWASH A CONDENSATE POLISHING SYSTEM FIL LPat CORRECTIVE ACTIONS:~
REMOVING TRIP ON MAIN FEEDWATER PUMP VIBRATIO ONLY (i '
REVISING OPERATING PROCEDURES TRAINING, NIGHT ORDERS REGION IV AND IE MONITORING STARTUP ACTIVITIES
- i
WATERFORD - 3, WOLF CREEK, CALLAWAY, CATAWBA AND BYRON -
STARTUP EXPERIENCE COMPARISON (W. JONES, IE)
- t..
FACILITY FULL POWER LICENSE DATE WATERFORD - 3 (03-16-85)
WOLF CREEK (06-04-85)
CALLAWAY.
(10-18-84)
CATAWBA (01-17-85)
BYRON (02-14-85)
BASED ONLY ON 50.72 NOTIFICATIONS-- SIGNIFICANT EVENTS ONLY (I.E. NOT CONTROL ROOM ISOLATIONS ETC) 6 FOR 4 MONTHS FOLLOWING FULL POWER LICENSE (ADJUSTED FOR SIGNIFICANT SHUTDOWNS)
Te
r c.#
6 REPORTED EVENT COMPARISON
SUMMARY
MONTH IsT 2ND 3RD llTH WATERFORD 14 2
2 9
1 WOLF CPEEK 9-I CALLAWAY 11 2
3 1
P-f?S BYRON 7
3 2
3 CATAWBA 5
0 1
5' MONTHLY AVG W/0 WATERFORD &
WOLF CREEK 7.6 1,6 2
3
~
- 1 MONTH SHUTDOWN NOT INCLUDED IN PERIOD fa, My
PEACTOR TRIP AND FEEDWATER C0WARIS0N DETAll i.
SIGNIFICANT REACTOR +
INEOLVING P0mH EVENT REPORTS TRIPS LOFW WATERFORD 3 1
4 4
3 2
2 2
0 3
2 2
i 4
1 1
.a SUBTOTAL 17 16 7
i WOLF CREEK 1
9 5
8 CALLAWAY 1
11 6
8 2
2 2
0 3
3 3
2 4
l l
1 17 12 10 -
t CATAWBA 1
5 3
1 2
-0 0
0 3
1 1
1
~
4 1*
1 1
11 6
3
)
BYRON 1
7 6
2 2
3 3
0 3
2 1
1 4
1 1
1 5
B 3
- MAY REFLECT RETURN TO POWER AFTER SRJTDOWN
+ BASED ON AEDD a NRR S"UDY FOR Ig@s, AVERAGE IUBER OF TRIPS FO 410 NTH PERICO WAS M, FOR ALL 9
m o'~
PERSONNEL ERROR & E0lliftENT PROBLEMS C0ffARIS0N D7t All SIGNIFICANT PERS EQP t(
M) NTH EVENTS ERROR M
' OTHER WAlttuRD 3 1
4 3
1 0
2 2
1 1
0 3
2 1
1 0
4 9
_2 3
1
\\
17 7
9 1
WOLF CREEX 1
2 l
l 1
1 9
3 5
1 CALLAWAY I
11 2
7 2
2 3
2 0
1 1
3 0
3 0
4 1
3
_1 9
P-2 U
3 CATAWBA 1
5 3
1 1
2 0
0 0
0 C9ll E'
_2
_1 2
..c 3
1 1
0 0
4 11 6
2 3
BYRON 1
7 2
4 1
2 3
2 0
1 3
2 0
1 1
4 I
J l
9 15 4
' 8 3
e
%.S l
- - -.-. = - -
CE LOCA ANALYSIS ERROR JULY 2, 1985 (H. ROOD, NRR)
NON-CONSERVATIVE ERROR FOUND IN CE LARGE-BREAK LOCA MODEL CENTER PEAK AXIAL POWER SHAPE YIELDS 34*F HIGHER PEAK CLAD TEMPERATURE (PCT) THAN "REVIOUSLY ASSUMED TOP-PEAKED SHAPE.
FOR THREE CE PLANTS THAT ARE ON IST CYCLE THIS WOULD YIELD A PCT IN EXCESS OF THE 220,0*F LIMIT OF 10 CFR 50.46.
PLdNTSARE:
PALO VERDE 1 SAN ONOFRE 3 3
Gd WATERFORD 3 BASED ON CE REANALYSIS, OTHER FACTORS IN LARGE-BREAK LOCA MODEL WILL REDUCE PCT TO BELOW 2200*F.
LETTERS FROM THESE 3 LICENSEES BEING SUBMITTED GIVING BASIS FOR 00NTINUED OPERATION, OTHER CE LICENSEES BEYOND CYCLE 1 AND (EVEN WHEN OTHER FACTORS NOT INCLUDED) HIGHER PCT DOES NOT REACH 2200*F LIMIT, NJ-
M0HAVE GENERATING STATION - REHEAT LINE FAILURE JUNE 9, 1985 (R. B0SNAK, NRR) b FAILURE OCCURRED JUNE 9, 1985 WHEN A 30" REHEAT LINE SUDDENLY SPLIT LONGITUDINALLY
-FRACTURE WAS FISH MOUTH RUPTURE APPROXIMATELY 20' x 6' FIG 1A 8B SAFETY SIGNIFICANCE FOSSIL PLANTS OF SIMILAR VINTAGE (h
NUCLEAR PLANTS 4
O e
G S
Y
REHEAT LINE - VITAL STATISTICS DESIGNED TO B31.1 CODE FOR STEAM CONDITIONS OF 1000*F AND 600 PSIG COMMENCED OPERATION 1971 CONSTRUCTION LATE 1960's FAILURE IN A HORIZONTAL SPOOL 30"-DIAMETER ROLLED AND W. ELDED OF A-378C PLATE (1 1/4 CR-1/2 N0) TO MEET A-155 s.
WELDED PIPE COMPARISON WITH LWR PIPING
~
MATERIAL NOT USUALLY USED IN LWR
' ('ll. -
UPPER TEMPERATURE NOT IN CREEP RUPTURE AND CREEP
'~
FATIGUE RANGE IN LWR FABRICATION CONTROLS INCLUDING NDE SUPERIOR IN LWR LEAK DETECTION REQUIREMENTS IN LWR INSERVICE INSPECTION IN LWR FAllURE ANALYSIS RESULTS EXPECTED FROM SCE BY EARLY AUGUST iLis:
PALUEL 1 & 2. IN-CORE INSTRUMENTATION TUBE VIBRATION PROBLEMS MARCH 29, 1985 (P. MORIETTE, NRR) b INITIAL EVENT: MARCH 29,1985, PALUEL 1 IN COLD SHU7DOWN, LEAK DETECTED ON ONE THIMBLE TUBE, WHILE LEAK TESTING IN-CORE INSTRUMENTATION SYSTEM, SUBSEQUENT FINDINGS:
APRIL 5:
MECHANICAL WEAR (WITHOUT LEAK) ON 4 OTHER
- THIMBLES, APRIL 16: A PROBE CANNOT BE COMPLETELY INSERTED IN ONE THIMBLE (PALUEL 1).
MAY-JUNEr-2-LEAKS ON PALUEL 2, ANOTHER LEAK ON PALUEL 1 SAFETY SIGNIFICANCE: REACTOR COOLANT LEAKS, OR: NO FLUX MAPS.
POSSIBILITY OF MIGRANT OBJECTS.
MAJOR POINTS:
DEFECTS (OR L$AKS) LOCATED AT
(
DISCONTINUITY IN GUIDING STRUCTURE CAUSE THOUGHT TO BE HYDRAULIC EXCITA ON DUE TO TURBULENCE IN THE 3
CORE SUPPORT PLATE - BOTTOM OF FUEL ASSEMBLY REGION.
DIFFERENCES (FROM 900MWE SERIES) IN LOWER INTERNALS DESIGN AND MEASURED FLOW PARAMETERS SUPPORT THIS HYPOTHESIS.
LOWER INTERNALS M DESIGN.
CORE INSTRU-MENTATION SYSTEM (0UTSIDE VESSEL)
FRAMATOME DESIGN,
~
GENERIC IMPLICATIONS:
ALL 1300MWE SERIES REACTORS AFFECTED 7
IN FRANCE v.
~ -
.i..
s?
k
\\
LICENSEE CORRECTIVE ACTIONS:
SHORT TERM:
JUSTIFY OPERATION WITHOUT IN-CORE INSTRUMENTATION FOR I 1/2 MONTH.
LONG TERM:
MODIFY THIMBLE GUIDING PIECES ON TOP
~
OF CORE SUPPORT PLATE FOR BETTER PROTECTION, REDUCE TURBULENT FLOW AROUND THIMBLES, ONLY AFFECTED US FACILITY: SOUTH TEXAS PROJECT 1 a 2 C.-*
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MILLSTONE UNIT 2 - FAILURE OF PRESSURIZER SPRAY VALVES TO SHUT JULY 15, 1985 (D. OSBORNE, NRR) s RETURNED TO 100% POWER ON 7/11/85 AFTER 135 DAY REFUELING OUTAGE REACTOR TRIPPED ON JULY 15 PRESSURESETPOINT(2130 PSI)1985ONTHERMALMARGIN/ LOW PRESSURE DROP HALTED AT 1725 PSI TRIP ATTRIBUTED TO FAILURE OF BOTH PRESSURIZER SPRAY VALVES TO SHUT FISHER CONTROL AIR OPERATED 3 INCH ANGLE VALVES ONE MISADJUSTED - (WOULD NOT FULLY CLOSE)
OTHER HAS MECHANICAL BINDING (PRELIMINARY)
DURING INVESTIGATION DISCOVERED SPRAY VALVES WIRED TO WRO CONTROLLER, AS SPECIFIED IN THE OUTAGE DESIGN CHANGE (NOT CAUSE OF FAILURE)
LICENSEE HAS REVIEWED OTHER REFUELING OUTAGE DESIGN CHAN AND TESTING
(-
CORRECTIVE ACTIONS: ONE VALVE WAS READJUSTED TO FULLY C OTHER VALVE WAS REPACKED AND REBUILT m
REGION REVIEWED PRIOR TO STARTUP PLANT RESTARTED ON JULY 20, 1985 4
4 N
+
~.--
~
FERMI 2 - INADVERTENT CRITICALITY JULY 2, 1985 (D. LYNCH, NRR)
(
b PROBLEM INADVERTENT CRITICALITY JULY 2, 1985, DLE TO OPERATOR ERROR i
SAFETY SIGNIFICANCE -
1 PLANT WAS IN ANALYZED CONDITION AND FEVER EXCEEDED SOURCE RANGE, REACTOR OPERATOR PULLED ROD BAM( (ABOUT 8 To 10 GS) FULL DUT (NOTCH 48) RATHER THAN NOTCH 4, AS SPECIFIED, JULY 2 INADVERIENT CRITICALITY DECLARED BY OPERATOR AND E S REINSERTED SHIFT SUPERVISOR REVIEWED OPERATOR Ahl0NS AND DETERMIfED THAT INADVERTENT CRITICALITY HAD NOT OCCURRED, NORPAL STARTlP COPPLETED, POSSIBLE LACK OF EFFECTIVE ASSISTANCE OR (ESERVATION BY SRO SHIFT TECWICAL ADVISOR WAS IN TRAINING 9RWFNT SEQUENCE OF EVENTS:
r
'?
JULY 7 LICENSEE TECHNICAL STAFF COWIIPED
. INADVERTENT CRITICALITY JULY 8 INADVERTENT CRITICALITY RE00 WIRED INDEPENDENTLY BY LICENSEE JLLY 10 COPMISSION BRIEFING ON FULL POWER LICENSE JLLY 15 REGION III NOTIFIED BY LIGNSEE (1:30 PM CDT) i JULY 15 FULL POWER LIGNSE ISSUED BY NRC (3:15 PM EDT)
REGION III ISSLED CONFIRMATORY ACTION LETTER LIMITING FERMI 2 TO 5 PERCENT POWER ENFORGPENT CONFERENCE SCHEDULED FOR JULY 23 INVESTIGATION CONTINUING 4
D i
l
LA SALLE UNIT 1 - RHP FLOW SWITCHES IIPROPERLY INSTAllID JULY 17, 1985 (R. CARUS 0, NRR)
PROBLEM FOUR RHR FLOW ISOLATION SWITCHES FOLND TO IE '
PIPED BACIMARDS SINCE MARCH 1985 EQ UPGRADE, AS A RESULT OF BREAKDOWN IN MANAGEPENT CONTROL AND QA EVENT IS A REPEAT OF PROBLEM WITH UNIT 2 ECCS AND RHR SWITCHES IDENTIFIED ON 6/10/85
' SAFETY SIGNIFICANT - RHR SWITCES SERVE DIVERSE AfD REDUNDANT FUNCTION TO ISOLATE RHR IN CASE OF BREAK - MIPOR -
SIGNIFICANCE, f
N.
LICENSEE WAITED FROM JtME 10 TO JULY 17 TO TEST INSTALLATION, SINCE PLANT SHUTDOWN FOR UNRELATED ACTIVITY.
CORRECTIVE ACTION - REGION III REVIEWING LIENSEE PLANS TD REVIEW P0DIFICATION CONTROL SYSTEM, DRAWING CHANGES TO BE RE-EVIEWED, BOTH UNITS TO REMAIN SHUTDOWN UNTIL REVIEW COPPLETE.
ESCALATED ENFORCEFENT ACTION BEING CONSIDERED, e
{J C..
UNPLANNED REACTOR TRIPS
- s AVERAGE WEEKLY TRIP FREQUENCY FOR PAST 74 WEEKS IS APPROXIMATELY 10 TRIPS / WEEK, WHICH IS NEAR AVERAGE BREAKDOWN OF REPORTED CAUSES AUTOMATIC
~
EQUIPMENT FAILURES 52%
PERSONNEL ACTIVITIES 40%
MANUAL 8%
v_.
- BASED ON 10 CFR 50 72 REPORTS FOR PLANTS WITH LICENSES FOR FULL POWER OPERATION 6
i e
f
\\
%+
4 m
I I
'FieNo.*
I -
t 07/29/85 OPERATIE REACTCRs' EVENTS REETINS FOLLONUP.ITERS At M E ETING B5-12 ON JULY 23,1985 GDERD BY ASCEEING EETING DATES, NSSS VEWOR$, FACILITY) i -
.6 FACILITY RESPONSIBLE-TASK ESCRIPTION SCHDULE CLOSEB M TE,
CORRDTS NUR8ER/
E555 VE CORI DIVISION /
COMPLET. BY SOC 1NIENT/
EETING EVENT ESCRIP.
ISIVIDUAL MTEIS)
EETIE,ETCs MR RANCMD SECO I E IVIS$!NS SUMARIZE MN LIC. RESPONSES TO 88/30/85 SPEN
//
STAFF REVID OF
. C /10/84 W / RAIN IEN
/
IUESTIONS SUBSEQUENT TO RANCHO
$7/30/I5 M N ONNERS ROUP
)
NYORDEEE SECO LOSS OF MI EVENT AND
//
SUBRITTE F EIPLOSION -
PRESENT AT FOLLON-IP R EVENTS 1/11/85 !$ IN PARTIE LOSS OF MIEFIM PROSESS ITH15 UNI 3/19/94 REACTIVATD-i CRYS.RIV. 3)
SALD 2 E /FISCER, 3.
ETEMIE IF VELM (PORV) KOCK 08/30/85 OPS
//
AE00 REPORT IN CC/07/84 5 / STUCK OPD
/
VEVE RIALIFID TO CLOSE
$7/30/85 P E PARATION NILL RELIEF V EVEl AEAINST 7/25/84 STEM KONDOW
//
ADDRESSTHAT ECCS ETUATIM TRANSIDT AT SALD 2. CECK ISSE.
JEY 25,1984 EPRI TEST PROGRM RESETS.
CRYSTE RIVER 3 ICSI/ ROSA,F.
CONSIDER EED FOR AD81T10NE 09/30/85 IPEN
//
ICSB !$
01/03/85 N / TEN. LOSS
/
REDUIREENTS ON ALARMS /
07/30/85 CONSIDERINE OCONEE
. OF M I 12/28/M ANNUNCIATRS
//
I LOSS OF ANNUNCIA.
14/25/M ) IN ANAL.
('.
0F RE9UIE RENTS.
I
'io CATANIA 2 K /JABSOUR, K.
FOLLON IF. MIEFIN FTER DE.
09/30/85 IPEN
//
PRELIR. REPORT 05/07/B5 5 / BOTH bit
/
OF OVERPRESS. EFFECTS M 08/01/95 REC'8. ADB'L 1 5 0.
TRA15 VARIOUS SYSTB S A B COR E CTIVE 08/07/2 REOUESTD FROM
~
DVERPRESSURIZD ETIONS.
LICDSE.
4/19/E RESEUTIONPEnlNE REVID BY IItR 6 RII.
R!LLSTOE 2 ORAS/MINtPHY,E.
REVIEN SEDE LETTB W DIY 10/01/85 IPO
//
ORNL CONTRACTD TO C2/07/85 CE / D9Y E /CO RAD, M.
CURRENT TESTI N ISSUE, IN VIEN 08/01/95 EVE UATE EFFECT F CURRENT TEST F OF MILLSTOE 2 FIGIN88.
08/07/85 COPPER R UDY STEM EEN.
CURENT TEST!N TUES,4/10/E i
1 85-10 MVIS DESSE I DST /SPEIS, T.
DET BRINE STATUS F 08/30/I5 OPEN
//
REMIN!E 2 ISSUES M/12/95 N / LOSS OF EL E /DEAGA210, A.
IMPLEENTATION OF TRI ITERS, 06/19/B5 STATUS OF (EEMERIC & IFAs)
MIN An AUI.
ENERIC ISSUES, RPAs AT MVIS
//
IfrLEIENT. OF' IN PROGRESS. IIT FEDNATER 6/9/85 BESSE TMT MY BE RELAfD TO TRI 195tES REPORT UE D 6/9/85 EVENT.
CLOS D - E.
REVIEN THOlFSON TO N.
DENTE EIIO F 6/20/I5.
35 '"
lAVISDESSEI SSI/PARR,O.
RE-EIARIE STAFF REVIEN OF.
01/30/85 OPEN
//
IIT REPORT UNDER (6
M / LOSS OF EL E /E AGA!!O, A.
ACCEPTABILITY / DIVERSITY OF 06/19/05 REVIEN 1
C MIN AS AUI.
M VII BESSE E N SYSTS.
//
I FEDNATER 6/9/05 l
PigeNo.*
2-07/29/85 OPBATINS REACTORS' EVENTS EETINS FOLLONUP ITEMS AS OF MEETINS 85-12 DN JULY 23, 1985 (DRDERED BY ASCECING EETING DATES, NSSS VECORS, FACILITY) iS FACILITY RESPONSIBLE TASK M SCRIPTION SCHEDULE CLOSD DATE COMMENTS BudERI WSSS VEEOR/
DIVISION /
COMPLET. 8Y 80CUMENT/'
REETING EVD T MSCRIP.
ICIVIDUAL MTE(S)
EETIE,ETC.,
MTE 85-10 MVIS ESSE 1 R /DEABA210, A.
PROVIDE H. DENTON WITH 08/30/85 OPEN
//
ITENS 2 & 3 IN 06/12/85 BN / LDSS OF ALL
/
1)COMPARISDN ETNED MVIS 06/19/85 ITER l-6/20/85 PRO 6RESS.
MIN AC AUI.
BESSE 1 AND THI-I DF TMI ACT!DN
//
ERO H.
FEDNATER 611/85 PLAN IRPLIENTATION 5TATUS THORPSON TO H.
IPDATE TO NURES TMI-I 1066
- 3) STATUS OF STARTUP FN PURP UP6RADE.
35-!!
LASALLE2 IE / M ER, R.
CONSIDER ISSUANCE OF IE NOTICE. 08/30/85 OPD
//
07/01/85 BE / D
/
//
RODIFICATION
//
PROBLERSAND LDSS OF E L ECCS JUNE 510,1985 85-!!
OYST S CREEK 1 DST / MINERS LEAD PRESSUREISOLATIONVEVE 08/30/85 OPEN
//
07/01/85 SE /
BE /CHERMY ASST. TESTING REDU!REENT TO E
//
INICONTROLLD ADDRESSD IN CRER BRIEFI E.
//
{-~
- g. -
LEAKAGE OF
~,
REACTIR C00U WT
.+
DUTSIM CONTAINMENT 6/12/85 85-!!
DYSTR CREEK 1 DHFS/800HER,H.
DETERMINE EFFICACT OF D6/
08/30/85 OPEN
//
07/01/95 GE /
/
OPERATOR ACTION AND EICES$1VE
//
18CDITROLLD RESPONSE TIE.
//-
LEAKAGE W
. REACTOR C00LANT DUTSIM CONTAINMENT 6/12/85
'ANCHO SECD 1 DL /MIER, S.
ETERMINE STATUS OF ID 79-14.
08/30/85 OPEN
//
THE LICENSEE IS 85-!!
R 07/01/85 N / RCS HIGH
/
//
COMPLETING THE POINTVSTLEAK,
//
INSPECTION 6/23/85 85-11 RAN HD SECO 1 R / MINER, S.
SCHEDULE CONFERD CE CALL NITH 08/15/85 IFEN
//
EVALUATION IN 07/01/85 3N / STARTUP
/
REGION AO LICENSE PRIOR TO 07/12/85 PR06RESS PUBLBS-JUNE RESTART TO DETERMIE PLANT
//
1985 READINESS FOR OPERATIONS.
h DCONE 2 DL /NICOLARIS, H SCHD ULE CONFERENCE CALL WITH 09/30/85 OPEN
//
07s. 5 BN / EITES O
/
LICENSEE TO DISCUSS CROS8Y
//
KONDOINIFROR VEVE PERFORMANCE
//
MIN STEAR SAFETT VEVES
Page D.
3 07/29/B5 OPERATIN6 REACTORS' EVENTS REETIN6 FOLL0rJP ITEMS AS OF EITING 85-12 DN JULY 23,1985 (DRDERED BY ASCENDING E ETING DATES, NSSS VENDORS, FACILITY) t U.'.d6 FACILITY RESPONSIBLE TASK MSCRIPTION SCHEDULE CLOSED MTE t-
.COMENTS NUMBER /
NSSS VE CORI DIVISION /
CORPLET. BY SOCUMENT/
EET!3 EVEN1 M SCRIP.
INDIVIDUAL MTE(S)
EETING,' ETC.-
MTE 85-12 SEABROOK 951 /MARSM ANALYIE SAFETY IMPLICATIONS OF 09/30/85 OPEN
/ l' 07/23/B5 W / CROSBY MIN
/
VEVE FLON KFICIENCY
//
STEM SAFETY
//
VEVEFLON BEFICIENCY 12/04 35-12 SEABROOK M / CHERNEY INVEST 16 ATE AE BUACY OF TESTING 09/30/85 OPEN
//
07/23/85 N / CROS3Y MIN
/
AND VE!DITY OF E1 TRAP 0LATINE
//
STEAR SAFETY M TA FROM M ALL TO LARSE Y E VES //
VEE RN DEFICIENCY 12/94
~
85-12 MTCN 1 ORAS/CARUSD,R.
N!U. MVEL W TIA TO COORDINATE 97/30/85 CLOSE3 07/23/85 05/30/85 SE / STUCK DPEN
/
IE NOTICE AND FETER
//
TIA IN SAFETYRELIEF INVEST!SAilVE EFFORTS.
//
CONtsNEEE VEVE e
W t
TURKEY POINT 3 - REACTOR _ TRIP AND AFW VALVE _FAltliRE JULY 21, 1985 AND JULY 22, 1985 (T. ROTELLA, NRR)
L s.
LICENSEE: FLORIDA POWER & LIGHT FACILITY: TURKEY POINT UNIT - 3 VENDOR:
WESTINGHOUSE INITIAL CONDITIONS: 100% POWER; STEADY STATE CURRENT CONDITIONS: PLANT IN COLD SHUTDOWN EVENT-SEQUENCE:
EVENT b/85i bl LIGHTNING STRIKE IN OR NEAR PLANT REACTOR TRIP FROM
.00% POWER FIRST-0UT ANNUNCIA;..OR TURBINE TRIP (CAUSE UNVER INV8STIGATION)
+1 7/22/85 00:40 LO-L6 LVL S/G #3B DUE MFW BYPASS VALVE FAILURE TO OPEN ti UPON OPERATOR INITIATION (CAUSE UNDER INVESTIGATION)
AFW PUMPS AUTO STARTED PROPERLY HOWEVER THE A&C PUMPS TRIPPED ON MECHANICAL OVERSPEED (CAUSE 3
UNDER INVESTIGATION)
+3.5 7/22/85 ' 04:05 MFW PUMP TRIP DUE TO HI LVL S/G #3C (CAUSE WAS DUE TO LEAKING MFW BYPASS bA P
UCCESS-FULLY STARTED AND RAN
+5.5 7/22/85 06:04 gN]EgNgLgNFROM AFW FLOW CONTROL VALVE FAILED TO CLOSE (AFW V-2833TOS/G#$C)CAUSE UNDER INVESTIGATION) 7 L:
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l UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
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IE Information Notice No. 85~-v: Main Steam Safety Valve Test Failures and /g Ring Setting Adjustments
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op Addressees:
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All PWR nuclear power reactor facilities holding an cperating license (OL) or a construction permit (CP)
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Purpose:
This Information Notice is being provided as a notification of a potentially significant problem pertaining to spring-actuated main steam gafety valves i
(See Figure 1), tAat may possess less than the full rated flow capacity required for overjressure protection of the secondary cooling system in c-PWRs.
It is expected that recipients will review the information for applicability to their facilities and consider actions, if appropriate, to i
preclude a similar problem at their facilities. However, suggestions contained in this Infomation Notice do not constitute NRC requirements; therefore, no specific action or written response is required.
I Description of Circumstances:
Between October 16, 1984, and December 1,1984, Wyle Laboratories conducted p
,i several full flow steam tests on two separate main steam safety valves (MSSVk) e-i manufactured by Crosby Valve and Gage Company. These Crosby 6R10 MSSVs are to be installed by Public Service of New Hampshire on the Seabrook main steam system. The tests were conducted in order to determine the adequacy of various MSSV yischarge piping arrangements. During the tests the valves were l
instrumentet*to measure valve disk lift. The valves were installed on the test facility with the Crosby recomended settings of the valve adjusting rings. With these factory ring settings the valve achieved about 50% of the full disk lift required to develop full steam flow capacity within the ASME-Code required 3% accumulated overpressure limit. Adequate lift was not achieved for either valve with the factory. ring settings, even for the largest i
diameter (least flow resistance) vent pipe tested. The guide ring of both j
valves was subsequently adjusted a significant amount (150 notches) during the course of testing and full disk lift was subsequently achieved.
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T'n' y ; *,r These types of full flow tests are normally not performed by either%)or the valve vendor on MSSVs, nor are such tests required for capacity certification according to the ^::rie:n S0:icty of "^:hnic:1 Engin n 4ASME) m,_
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hee =1 Code,Section III.
In general, these valves are capacity certified by tests on much smaller size valves, and the capacities
- then extrapolated to larger size valves. The MSSVs on most PWRs, while not necessarily the same model, are like those at Seabrook in that these valves are generally at the upper end of the valve size range. This raises the concern that full flow functional demonstration of these type valves may never have been performed.
A related.MSSV problem which has occurred at several PWRs in the past few years pertains to excessive blowdown of main steam system pressure during transients which have actuated some MSSVs.
On separate occasions at the Oconee, Salem, Trojan, and Davis-Besse nuclear facilities, MSSVs have remained open below the correct reseat pressure and have blown down excessive amounts of steam. The design blowdown value is usually 5% of setpoint pressure and is also dependent on the specific valve ring setting adjustments. Some valves at these facilities have exhibited as much as 10% blowdown. This raises the concern that some MSSVs may remain open too long, relieving excessive quantities of steam possibly adversely affecting cooling of the primary system and causing excessive thermal stresses on primary system components.
No specific action or written response is required by this Information Notice.
If you have any questions about this matter, please contact the Regional Administrator of the appropriate regional office or this office.
Edward L. Jordan, Director Division of Emergency Preparedness and Engineering Response Office of Inspection and Enforcement
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Coordination Justih RDAARKS n
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r MORNINO REPORT - REOION II DATES Aueust 8.
1903 LICENSEE / FACILITY NOTIFICATION / SUBJECT
' DESCRIPTION OF ITEM OR EVENT Orend Outf HQ Duty Officer 8/7 On 9/7. at 3802 p.m..
Unit i tetered frem 941C power due to a main turbine DNs30-416 eenerator trie.
The turbine senerator triered on "Iow cooline water flow s
to the rotor" due to a flow trensmitter falline low. _ Safety relief valves' lifted and resented as they should have The licensee is makins necessary s
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'reesses and Plans to restarr sne7nst today.
For information only.
Unit 3 Shut Down en 8/8. for a scheduled fifty-seven day refueline outase.
I Resident Inspector 8
M hg g}\\ The maJer maintenance teoks planned are studse j
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lancine and eddy current e
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testino of steam renerators..I Arts inspection finservice) of reactor hot y
d, /1/11e i.es, r.6unidine on. c. actor c..iant pune- --. haus of < - code s.fety d'
Q valves imain steam). overhaul of two-low pressure turbines, overhaul of l
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both main feedwater turbines and testine and eluesine of moisture 4
separator reheater tubes.
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St. t_ucie 2 HQ Duty Officer. 9/8 At 2sO6 a.m.. 9/8. a fuse blew in the normal rowtr feeder to the "A"
c.
2
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DNI 50-309 Reactor Shutdown safeeuerds actuation cabinet.
Imme d i a t e l v'. q_._ f u s e blew in the backup e'owse feed.
These two power feeds are rectified - then auctionered to provide
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g u3 Interrupted power to the cabinet. With both power feeds lost. the l
actuation cabinet properIV initiated "A" side safesuards featuress hish i
and low Pressure safety injection Punes, start of the "A"
emeroency diesel
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.]J g eenerator, and containment isolation. The injection Punes did not inject because plant pressure never reduced to the injection point.
The
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enereency diesel venerator did not load onto the bus because normal e
'A power was available. Containment isolation interrupted cam.anent cooline
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" d'" ICCW) to all four reacter coolant pumps (RCPs) at their containment Penetras...... -..... ;,..... n s s y Enese valves were reopeneds however individual e
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.d RCP CCW lines to 2A1 and 292 RCPs were isolated because these individual RCP CCW isolatten valves are Powered from a power sueely which is striered j
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" uren safesuerds initiation. These valves could not be roepened. The e
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j plant used natural circulation to remove decay heat while the pumps were Q8#6lf
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e stoceed. 2A1 and 292 RCPs remain stoamaff__rendine shaft seal inspection i
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at cold shutdown.
"B" safesuards initiated due to low pressurizer Pressure and was reset in about five minutes when Pressure was restored
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above the setreint. Auxiliary Feedwater operated as required. The unit l
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'is beine cooled to cold shutdown for RCP seal inspection. An Unsual Event e
4.
was declared at 3:10 a.m. and terminated at 6:31 a.m.
PN issued.,
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MORNINO REPORT - REGION II DATE: Acoust 8.
1995 1 r General i
l Rece-sentatives of the Mississire! Power and Lisht Comeany are in' the Recion II Of fice to attend a meetine reoardine the arrasent-
j 3 I na d.riue t e safety evaluation of the Accendix R revised Safe Shutdown Systems at the Grand Gulf facility.
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Engincaring Offica k
E. Brown Projects - WJD l
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A. Cerne Projects - Chrono R. Cummings Ropes 6 Gray (3)
Ohd W.
e ickson A.
h a J. DeVincentis R. Sweeney N w Hampshire Yankee Divialen enbaum T.F. Q2.2.2 G. Gram G. Then:as SBN-863 W. Hall H. Tracy T.F. Q2.2.2 R. Harrison J, Tribbic D. Hunter UECC 6 W (SB-19770:
W. Johnson M..Wiley_
G. Kingston ASLB United States Nuclear Regulatory Commission G. F. Mcdonald 10CFRSO.SS(e)l;ile Region I M. McKenna J. Allen 631 Park Avenue B. Middleton INPO King of Prussia, PA 19406 D. Moody NRC Subject File Attention:
Mr. Richard W. Starostecki, Director Division of Project and Resident Programs References (a) Construction Permits CPPR-135 and CPPR-136, Dceket Nos. 50-443 and 50-444 (b) Telecon of December 21, 1984, A. L. Legendre, Jr.
(YAEC) to J. Grant (Region I)
(c) NHY Letter SBN-751 dated January 17,1985, John DeVincentis to R. W. Starostecki, NRC Region I (d) NHY Letter SBN-788 dated April-8,1985. John DeVincentia to R. W. Starortecki
Subject:
Final 10CFR50.55(e) Report." Main Steam Safety Valve Ring Setting Caficiency," (C'R 84-00-19),
D
Dear Sir:
In Ref erences (c) and (d), we filed interim 10CFR50.55(e) reports re-garding a ring setting deficiency for the main steam safety valves. The valves were sent to Wyle Laboratories for, testing for deter aination of the proper ring settings. The tests were completed and the results are contained in Wyle Laboratories Report No. 47787-01 dated July 12, 1985.
The objectives of the tests were to:
1.
Determine if the "as-shipped" t hs, settings of the valves would allow the required disc tradi vi h ulnimum tailpipe backpressure.
disc travel for a range of 2.
Detensine the ef fects ot. f 4 r.
os backpressutes between 180 and 3% psig.
The results of the "as-shipped" rir.g setting tests indicated that the valves could not achieve the required disc travel with 3% steam accumulation at minimum tailpipe pressures of 15-20 psig.,
l S-(Nh co-~-> ci. en < ~..,n.re,nn s.oiet rw.c.. P O Her 7CC a Sect > rook, NH 03074
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Attention:
Mr. Richard W. Starostecki Page 2 During the test, the upper (guide) ring setting was adjusted f ros +155 notches to 0 and +25 notches, and full required disc travel was achieved at 3% steam accumulation under'the full range of tailpipe backpressure tested.
As a result of these test, we have agreed with Crosby, the valve manu-f acturer, that the optinua ring settings for the Seabrook main steam saf ety valves shall ba -25 notches for the lower (nozale) ring (original setting) and +25 notches for the upper (guide) ring. The corrections were completed by a Crosby service representative at the Wyle f acility prior to returning the valves to the Seabrook gita.
This is our final report on this subject.
Very truly your,
A n DeVincentis, Director ngineering and Licensing ec: Atomic Safety and Licensing Board Service List Director, Office of Inspection and Enforcement U. S. Nuclear Regulatory Commission Washington, DC 20555 l
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T William S. Jordan, 11I Donald E. Chick Diane Curran Town Manager Ha rmo-Weiss & Jordan Town of Exeter l
i 2000' ;. Street, N.W.
la Front Street Suite 430 Exeter, NH 03833 Washington, D.C.
20009 Brentwood Board of Selectmen Robert G. Perlis RED Dalton Road Office of the Executive Legal Director Brentwood, NH 03833 U.S. Nuclear Regulatory Commission Washington, DC 20555 Richard E. Sullivan, Mayor City Ball Robert A. Backus, Esquire Newburyport, MA 01950 116 Lowell Street P.O. Box 516 Calvin A. Canney Manc he ste r, NH 03105 City Manager City Hall Philip Ahrens Esquire 126 Daniel Street Assistant Attorney General Portsmouth, NH 03801 Augusta, ME 04333 Dana Bisbee, Esquire Mr. John B. Tanzer Assistant Attorney General Designated Representative of Office,of the Attorney General the Town of Hampton 208 State House Annex 5 Horningside Drive Concord, NH 03301 Hampton, NH 03842 Anne Verge, Chairperson Roberta C. Pevear Board of Selectmen Designated Representative of Town Hall the Town of Hampton Falls South Hampton, NH 03827 Drinkwater Road Hampton Falls, NH 03844 Patrick J. McKeon Selectmen's Office Mrs. Sandra Gavutis 10 Central Road Designated Representative of Rye, NH 03870 the Town of Kensington RFD 1 Carole F. Kagan, Esquire East Kingston, NH 03827 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Jo Ann Shotwell, Esquire Washington, DC 20555 Assistant Attorney General Environmental Protection Bureau Mr. Angi Machiros Department of the Attorney General Chairman of the Board of Selectmen One Ashburton Place,19th Floor Town of Newbury Boston, MA 02108 Newbury, MA 01950 Senator Gor' don J. Humphrey Town Manager's office U.S. Senate Town Hall - Friend Street Washington, DC 20510 Amesbury, MA 01913 (ATTN: Tcm Burack)
Senator Gordon J. Humphrey Diana P. Randall 1 Pillsbury Street 70 Collins Street Concord, NH 03301 Seabrook, NH 03874 (ATTN: Herb Boynton)
.