ML20055C053
| ML20055C053 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 08/06/1982 |
| From: | Bachmann R NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | WISCONSIN'S ENVIRONMENTAL DECADE |
| References | |
| NUDOCS 8208100076 | |
| Download: ML20055C053 (17) | |
Text
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D IGNATED ORIGIng Iff ed &
J-08/06/82 Mop
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMilSSION
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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of Docket Nos. 50-266 WISCONSIN ELECTRIC POWER COMPANY 50-301 (Point Beach Nuclear Plant, (Repair to Steam Generator Tubes)
Units 1 and 2)
NRC STAFF'S ANSWER TO DECADE'S INTERROGATORIES RELATIVE TO THE SAFETY EVALUATION REPORT ON FULL SCALE SLEEVING I.
INTRODUCTION On July 21, 1982, Wisconsin's Environmental Decade (Decade), in accordance with the telephone conference held on June 1,1982, submitted discovery requests I to the NRC Staff concerning the Staff's Safety Evalua-I tion Report (SER) issued on July 8, 1982. The Staff's answer to the discovery requests is to be filed within 15 days after receipt of the discovery requestse _/ In the interest of expediting this proceeding, the 2
Staff is providing answers to Decade's discovery requests without invoking the provisions of 10 C.F.R. SS 2.720(h)(2)(ii) and 2.744. The Staff does not waive the right to require the Intervenor to comply with 10 C.F.R.
SS 2.720(h)(2)(ii) and 2.744 for future discovery requests.
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Decade's First Interrogatories and Request for Production of Docu-ments on Staff Relative to the Safety Evaluation Report on Full Scale Sleeving, dated July 21, 1982.
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See letter from Bruce Churchill, Licensee's Counsel, to Peter Bloch, M ensing Board Chairman, dated June 7, 1982.
8a08100076 820806 PDR ADOCK 050CO266 G
2-Professional qualifications of Patrick G. Easley are attached.
Professional qualifications of Emmett L. Murphy were submitted pre-viously. ~
Respectfully submitted, Richard G. Bachmann Counsei for NRC Staff cc:
Service List Dated at Bethesda, Maryland this 6th day of August,1982 l
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O NRC STAFF'S ANSWERS TO INTERROGATORIES Interrogatory 1 With reference to pages 35 and 36 of the Safety Evaluation Report Relating to Full Scale Sleeving ("SER"):
a.
State any information related to operating experience with sleeved tubes that is not set forth on those pages and that has not previously been provided in response to prior discovery requests.
Response
Palisades Information pertinent to operating experience with sleeves at Pali-sades is attached and includes Consumers Power Company reports dated March 1, 1978, November 16, 1979, and October 16 1981.
Information from the March 1,1978 report was overlooked by the staff when preparing the background summary of operating experience for Pali-sades, and some correction and clarification of information presented in the SER for Palisades is necessary.
First, initial sleeve installation (ten sleeves) took place in February 1976, not 1978 as stated in the SER.
Secondly, the March 1, 1978 report notes that the 33 sleeves currently in service do not include six sleeves in four tubes which were plugged in February 1978 (note some tubes contain more than one sleeve). Two of these four tubes (each containing one sleeve) were plugged as a result of examinations of the sleeves performed immediately after they were installed in February 1978, and not as a result of service related problems e
with the sleeves. The other four sleeves had been installed in two tubes in February 1976. Eddy current test (ECT) inspection techniques used prior to i978 (i.e., circumferentially wound probe, single frequency) were not capable of inspecting the sleeve-to-tube joints.
In February 1978, a rotating pancake (RPC) probe was utilized which did have this capability.
However, two of those four sleeves could not be inspected with this probe due to an overly tight fit between the probe and the lower sleeve entry point. The other two sleeves exhibited RPC indications at the joints.
The two tubes containing those four sleeves were plugged. However, in
- the absence of previous baseline RPC data for these tubes, the February 1978 RPC inspection results do not establish whether the tight fit and indications were present at the time of installation, or whether they
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occurred as a result of subsequent service.
In the staff's judgment, the first explanation appears to be the most likely since inspections performed subsequent to February 1978 with the RPC probe have not revealed any new indications of degradation in the 33 sleeves which remain in service.
Oconee 1 Staff records regarding steam generator operating experience for Oconee 1 include no reports of sleeving related problems (i.e., sleeve wall degradation or sleeve related leaks). The absence of problems has been confirmed during phone conversations with the licensee for Oconee.
In addition, based upon discussions with licensee, we understand that a total of ten sleeves have been installed to date, rather than 26 as stated
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in the SER.
3 R. E. Ginna 1 No sleeve degradation or sleeve related leaks were reported during the April'1981 and February 1982 inservice inspections. Summary results of these inspections were submitted by Rochester Gas and Electric Company in reports dated July 8,1981 and April 26, 1982. The July 8, 1981 and all pertinent portions of the April 26, 1982 report are attached.
Point Beach 1 No sleeve wall degradation or sleeve related leaks were reported during the first inservice inspection of the sleeves in March 1982.
The licensee submitted the results of this inspection by letter dated April 16, 1982 which we believe is already in the possession of Decade.
San Onofre 1 San Onofre has operated approximately 4-1/2 effective full power months since the large scale sleeving repairs were performed. Results of inspections performed during the current outage were described during a meeting with the licensee on May 12, 1982. Hydrostatic tests performed during this outage revealed three sleeved tubes with very minor leaks as discussed in further detail in response to interrogatory 1.c.
Total primary to secondary leakage prior to shutdown was 1 gallon per day which is very small compared to the allowable leakage (215 gallons per day) as specified in the plant license.
A ten percent sample of the sleeves were inspected during this outage. Signals similar to those produced by magnetite were observed
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at the upper transition region of sleeve joints; however, no indications of sleeve wall degradation were found.
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_4 Westinghouse attributed the magnetite-like signals to residual magnitite remaining on the ID of the tubes following the decontamination work in the channel head area and subsequent rinsing of the tubes. This magnetite is believed to have accumulated at the upper joint transition after the sleeved tubes were filled with water. Meeting passouts sumariz-ing the inspection results are attached. A slide describing the results of the Westinghouse laboratory evaluation of the magnetite like signals and marked as " Westinghouse Proprietary" is not attached. Westinghouse should be contacted if copies of the May 12, 1982 meeting passouts marked " Westinghouse Proprietary" are desired by Decade. A formal report of the inspection results will be submitted by the licensee before San Onofre returns to power.
Interrogatory 1 b.
State the type of joint design in each of the five plants which have had tubes sleeved, the vendor for each of the five sleeving operations, and details of any tests performed on the sleeved tubes in the five plants.
Response
l Post-Installation Plant Type of Joint InspectionandTest(1) i Palisades (sleeve vendor:
Upper and Lower Joints: Baseline ECT Combustion Engineering)
Expansion Diameter Measurements Note (2)
Oconee 1 (sleeve vendor:
Upper and Lower Joints: Baseline ECT Babcock & Wilcox)
Expansion Note (3) i l
l Ginna 1 (sleeve vendor:
Upper Joint: Braze In-situ leak test Babcock & Wilcox)
Note (4)
Lower Joint: Explosive Visual l
Weld Note (4) l i
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San Onofre 1 (sleeve vendor: Upper Joint: Braze Note (4)and(5)
Expansion Diameter Measurements Notes (2) and (4)
Lower Joint: Expansion Diameter Measurements Notes (2)and(4)
Point Beach (sleeve vendor:
Upper Joint: Braze Notes (4) and (5)
Expansion Diameter Measurements Notes (2)and(4)
Lower Joint: Expansion Diameter Measurements Notes (2)and(4)
Notes:
(1) Listed items are in addition to reviewing process records to ensure that process input parameters were within acceptable ranges.
(2) Diameter measurements performed on a sampling basis at Point Beach (for full scale sleeving) and San Onofre. To our knowledge, we have no information regarding whether diameter measurements at Palisades were performed for all joints or just a sample.
(3) Records of NRC staff members do not contain details of other post process inspections which may have been performed.
(4) The installed sleeves also reviewed a baseline ECT inspection and system hydrostatic pressure test.
(5) Proprietary volumetric inspection of braze.
Information in our records regarding the results of the post installation tests are as follows:
i Palisades l
The only information we have in our records is contained in Consumer Power Company's report dated March 1,1978.
Post-installation inspections resulted in the plugging of two tubes in February 1978.
Oconee 1
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l We can find no information in our records regarding the
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results of post-installation tests and inspections.
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_0 R. E. Ginna 1 Results of the post-installation inspections and test results have not been reported to the Staff. Based on discussions with the licensee for Ginna, all sleeves installed to date were found acceptable for service.
Point Beach 1 Information relating to post-installation inspection and testing was submitted by the licensee by letter dated January 25, 1982 which is already in the possession of Decade.
San Onofre 1 A number of sleeves were plugged as a result of post process inspections and tests. The information below is based upon passouts marked Westinghouse proprietary which were provided to the Staff by Southern California Edison during meetings on May 19, 1981, with additional information provided on May 12, 1982. Westinghouse should be contacted if copies of the meeting passouts are desired.
At least 83 tubes (possibly more) were plugged as a result of excessive dissolution at the braze joints as determined by a review of the process records, eddy current testing, and hydrostatic testing.
Five additional tubes (with expansion and/or braze joints) were plugged as a result of leaking joints during hydrostatic testing.
An unreported number of tubes were plugged as a result l
of the expansion joints being formed outside the specified locations or as a result of the expansions E..
being formed at a location where IGA was present on the outer tube. These unacceptable conditions were identified as a result of the baseline ECT examination.
We do not have information regarding any additional l
sleeved tubes which may have been plugged as a result of unacceptable post-installation inspections and tests.
Interrogatory 1 c.
State the amount of the leakage at San Onfore 1 at initial testing following sleeve installation compared to the amount of leakage at the testing following 4-1/2 effect full power t
months of operation.
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_7 Responsq Information regarding these leakers was provided to the staff by Southern California Edison during a meeting in May 1981. Additional information was provided during a meeting on May 12, 1982.
Pertinent meeting passouts were marked " Westinghouse Proprietary" and thus Westinghouse should be contacted for copies of these passouts.
Seven of the twelve leakers observed following sleeve installation leaked during an unpressurized water fill of the secondary side. These leaks were the result of through wall penetrations of the sleeve and tube wall as a result of the dissolution phenomenon described in the SER.
The amount of leakage was not recorded. We understand from phone conver-sations with the Licensee that at least some of some of these seven tubes had " gushing" type leaks rather than " dripping" type leaks.
This occurrence led to an investigation by Westinghouse to establish the cause of the problem and to develop corrective measures (including revised eddy current test procedures) to ensure that all tubes with excessive dissolution of the base metal were removed from service.
An additional five tubes leaked during a 700 psid secondary side hydrostatic test. The leaks were attributed to leaking expansion and/or braze joints rather than to base metal degradation. We understand on the basis of recent phone conversations with the Licensee that the precise amount of leakage was not recorded at the time. However, the leaks were dripping type leaks and thus very small in magnitude.
Three tubes leaked during the 700 psid secondary side hydrostatic test which was performed after 41 effective full power months of plant operation with the sleeves installed. The leakage rates observed during the test
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ranged from 1 drop per five minutes to two drops per minute for these tubes.
Total primary to secondary leakage observed prior to shutdown was 1 gallon per day.
Interrogatory 1 d.
State the results of any quality assurance tests of the joints in the three leaking sleeved tubes at San Onofre.
Response
The Staff does not have any information regarding whether the expansion joints in these tubes were included in the sample of tubes which received the post-process diameter measurement checks. As indicated by the test verification program, however, some leakage is to
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be expected from the upper expansion joints during service. The braze joints which leaked (See page 36 of SER) did not receive the proprietary post process volumetric inspection of the braze. The likely explanation for the leak is that the braze was not a fully continuous seal.
All sleeved tubes at San Onofre received a baseline eddy current inspection and were subject to the system hydrostatic test subsequent to the sleeving operation. Results of the baseline eddy current examination of these tubes were not reported to the Staff. The three sleeved tubes which leaked following 41 effective full power months operation had not been reported as leakers during the post-installation system hydrostatic test.
As previously noted, the leakage at San Onofre was extremely small relative to acceptable limits for normal operation. We do not believe these leaks. to be of any potential safety concern, either for San Onofre or Point Beach.
_9 LIST OF DOCUMENTS 1.
Consumers Power Company report dated March 1, 1978,
" Palisades Plant - Steam Generator Inspection Report."
2.
Consumers Power Company report dated November 16, 1979,
" Palisades Plant - Steam Generator Inspection Report."
3.
Consumers Power Company reoort dated October 16, 1981,
" Palisades Plant - Steam Generator Inspection Report."
4.
Rochester Gas and Electric Company, letter dated July 8,1981, "LER 81-009/01X-1, Abormal Degration of Steam Generator Tubes, Update Report Previous Report dated May 29, 1981, R. E. Ginna Nuclear Power Plant, Unit No.1."
5.
Rochester Gas and Electric Company letter dated April 26, 1982,
" Steam Generator Evaluation, Steam Generator, Tube Rupture Incident, R. E. Ginna Nuclear Power Plant."
(Only those portions of this report relevant to sleeving are enclosed) 6.
Passouts from NRC meeting with Southern California Edison Company / Westinghouse on May 12, 1982.
Passouts marked
" Westinghouse Proprietary" are not enclosed.
In addition, we have deleted words in the enclosed passouts which describe the proprietary NDE method for brazes.
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Interrogatory 2.
With reference to pages 42 and 42 of the SER:
- a. State any other informatinn related to the " generic review" Iot set forth in those pages or in response to prior discovery
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requests.
Response
As a result of the Ginna steam generator tube rupture (SGTR) accident, the staff is conducting a generic review of the consequences of SGTR accidents.
In this generic evaluation, the staff will consider issues such as primary system pressure control following the loss of primary coolant out the broken tube, the potential for steam generator overfill and its associated problems, the duration of primhry-to-secondary leakage, the importance of operator actions in mitigating the effects of the accident, assumed single failures as they effect the overall outcome of the accident, l
and the adequacy of current technical specifications for radiciodine activity I
1 concentration in the primary coolant for limiting releases of radioactive iodine to the environment.
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- b. State all parameters in which two-loop Westinghouse plants may not have been analyzed in an appropr'iately conservative manner.
Response
Two-loop Westinghouse plants may not have been analyzed conservatively in past SER analyses with respect to the following parameters:
the duration of primary-to-secondary coolant leakage, primary-to-secondary decontamination factors for iodine following potential overfilling of the steam generators, and the potential for relief or safety valvas failing to re-seat properly following actuation. However, the possibility that some of the parameters that affect an offsite dose calculation were not analyzed conservatively does not necessarily mean that the calculation as a whole is not conservative.
Because of conservative assumptions.for several other parameters in the dose calculation, there is little doubt as to the over-all conservatism of the results of the analysis.
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PATRICK G. EASLEY OFFICE OF NUCLEAR REACTOR REGULATION U.S. NUCLEAR REGULATORY C0t1 MISSION PROFESSIONAL QUALIFICATIONS I am Patrick Easley, and I work in the Accident Evaluation Branch of the Division of Systems Integration. My duties include evaluation of engineered safety features and the consequences of design basis acci-dents, both for licensing actions and for operating reactors. This has included the evaluation of offsite dose consequences resulting from steam generator tube ruptures and main steam line breaks.
I have written work descriptions for contracts to model transients following a
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steam generator tube rupture.
I have participated in research meetings on the modelling of iodine transport during a steam generator tube rupture; associated with these was a tour of a steam generator fabrica-tion facility.
I calculated releases resulting from the Ginna steam i
generator tube rupture; and contributed to the NRC staff report on the l
Ginna accident, the staff Ginna restart Safety Evaluation Report, and staff generic recomendations from the Ginna study.
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1 received a B.S. degree in chemical engineering from Oregon State l
University in 1975, and an M.S. in chemical engineering from the University of Washington in 1980.
I have a total of five years of i
experience in the nuclear field; three years with NRC and two years with Atlantic Richfield Hanford, at the Hanford site in eastern Washington.
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J AFFIDAVIT OF PREPARATION
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f I, Emmett L. Murphy, being duly sworn, state that I was responsible for preparing the foregoing response to Interrogatory No.1.
That reponse is true and correct to the best of my knowledge.
4 LPk thmett L/ Murphy Sworn to and signed before me this day of August, 1982 8?fY:
?,{iD k l' Notary fullic My Comission expires: 0k')i$h
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v AFFIDAVIT OF PREPARATION I, Patrick G. Easley, being duly sworn, state that I was responsible for preparing the foregoing response to Interrogatory No. 2.
That response is true and correct to the best of my knowledge.
Patrick G. Easley Sworn to and signed before me this day of August, 1982.
hotary Public My Commission expires:
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UNITED STATES OF AMERICA NUCLEAR REGULATORY C0f t*'.ISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
Docket Nos. 50-266 WISCONSIN ELECTRIC POWER COMPANY
)
50-301
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(Point Beach Nuclear Plant,
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(Repair to Steam Generator Tubes)
Units 1 and 2)
)
CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF'S ANSWER TO DECADE'S INTERR0GATORIES RELATIVE TO THE SAFETY EVALUATION REPORT ON FULL SCALE SLEEVING," in the above captioned proceeding have been served on the following by deposit in the United States mail, first class or, as indicated by an asterisk, t.hrough deposit in the Nuclear Regulatory Commission's internal mail system or, as indicated oy a double asterisk by Express Mail, this 6th day of August,1982.
Peter B. Bloch, Chairman
- Bruce Churchill, Esq.
Administrative Judge Gerald Charnoff, Esq.
Atomic Safety and Licensing Board Shaw, Pittman, Potts & Trowbridge U.S. Nuclear Regulatory Commission 1800 M Street, N.W.
Washington, DC 20555 Washington, DC 20036 Dr. Hugh C. Paxton Atomic Safety and Licensing Board Administrative Judge Panel
- 1229 - 41st Street U.S. Nuclear Regulatory Commission Los Alamos, New Mexico 87544 Washington, DC 20555 Dr. Jerry R. Kline*
Atomic Safety and Licensing Appeal Administrative Judge Panel (5)*
Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 Washington, DC 20555 Docketing and Service Section (1)*
Kathleen M. Falk, Esq. **
Office of the Secretary Wisconsin's Environmental Decade U.S. Nuclear Regulatory Commission 114 North Carroll Street Washington, D.C.
20555 Madison, WI 53703
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Francis X. Davis, Esq'.
P.O. Box 355 Pittsburg, PA 15230 4.
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Barton Z. Cowan, Esq.
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John R. Kenrick, Esq.
Ricliard G. Bachmann Eckert, Seamans, Cherin & Mellott Counsel for NRC Staff 42nd Floor, 600 Grant Street Pittsburgh, PA 15219
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