ML20093B062
| ML20093B062 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 10/01/1984 |
| From: | Ellis J Citizens Association for Sound Energy |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| CON-#484-322 OL, NUDOCS 8410090440 | |
| Download: ML20093B062 (87) | |
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00l!KETEg0/1/84 usmc UNITED STATES OF AMERICA
'84 OgJ -9 N0 :46 NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of I
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TEXAS UTILITIES GENERATING i
Docket Nos. 50-445-D L COMPANY, et al.
I and 50-446-0 C -
1 (Comanche Peak Steam Electric Station i Station, Units 1 and'2)
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CASE'S ANSWER TO APPLICANTS' 9/13/84 SUPPLFMENT TO MOTION FOR AUTHORIZATION PURSUANT TO 10 C.F.R. 50.57(c)
Applicants' filed a Motion for Authorization to Issue a License to Load Fuel and Conduct Certain Precritical Testing on 8/7/84 (received by CASE on 8/8/84). CASE responded to Applicants' Motion on 8/18/84 (opposing it), and the NRC Staff responded on 8/22/84. On August 24, 1984, the Licensing Board issued its MEMORANDUM (Request for Evidence Relevant to Fuel Loading).
On September 13, 1984, Applicants filed their Supplement to Motion for Authorization Pursuant to 10 C.F.R. 50.57(c) (received by CASE on 9/14/84).
In an off-the-record discussion initiated by CASE with Judge Bloch and Applicants' Counsel Nicholas Reynolds d / on 9/24/84, CASE sought and was granted (with no opposition com Applicants) leave to file our response to Applicants' Supplement aftet we had completed and filed our Answer to Applicants' pleading on A500 Steel (which the Board is treating as a Motion for Summary Disposition); it was agreed that CASE could file its response today, 10/1/84. (The NRC Staff counsel, Mr. Treby, had been advised earlier d/ It should be noted that the Board Chairman offered Applicants' counsel the option of the Board Chairman's not engaging in this off-the-record conference calls; however, Applicants' counsel stated that he had no objection.
1 8410090440 841001 PDR ADOCK 05000445
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in the day that CASE intended to contact the Board regarding this matter, but Staff counsel was unavailable at the time of the conference call.)
On Thursday, 9/27/84, CASE was contacted by Staff's counsel Mr.'Scinto, with Mr. Mizuno also on the line, regarding the Staff's desire to postpone their response to Applicants' pleading until October 12, 1984. As CASE advised at that time, we have no objection to this postponement and will be back in touch should it appear that CASE also needs additional time to respond. CASE does, in fact, plan to supplement our. response as soon as we
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have in hand additional information which we are currently preparing, and will ask at that time that the Board also consider this additional information. However, we wanted to go ahead and get the information contained in this pleading into the hands of the Board for their consideration.
Applicants' Responne to the Board's Order Is Inadequate In the Board's 8/24/84 Order, it stated, in part:
the section [10 CFR 50.57(c), covering a license for low power testing] requires us to make the findings listed in 50.57(a) with respect to the contested activity sought to be authorized." (Emphasis in the original.)
"The contested activities involve at least the following plant systems:
(a) boron addition and monitoring equipment, (b) neutron monitoring equipment sufficient to detect significant increases in Keff above 0.95, (c) fuel handling equipment, and (d) reactor protection systems.
Each of the components of these systems is relevant, including mechanical, electrical and instrumentation systems."
(Emphasis added.)
"Because of the broad quality control contention pending in this proceeding, we must have evidence concerning the adequacy of quality control for the contested systems.
In particular, we require evidence concerning the current status of QA/QC oversight of these systems, i
including evidence that documentation is adequate to assure that unsatisfactory or non-conforming conditions have been corrected and evidence concerning whether or not there are allegations known to the applicants or Staff about the intimidation of QA/QC personnel who were working on these systems." (Emphases added.)
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- oc "We'also require evidence: '(1) that appropriate OA/QC procedures have been completed for all phases of the activities for which a license is sought, (2) concerning the maximum Keff to be permitted during pre-critical testing and the Keff that analysis suggests may be achieved during pre-critical testing if all control rods were inadvertently removed while the boron concentration was 2000 ppa, and (3) that non-borated water will never be injected into the core, substantially 1
diluting the boron below 2000 ppa."
(Emphasis added.)
. ORDERED:
"That Texas Utilities Electric Co., et al. shall supply the evidence' requested in this order to facilitate further consideration of its Motion for Authorization to Issue a License to Load Fuel and i
Conduct Certain Precritical Testing."
CASE submits that Applicants' response to the Board's Order is inadequate, as discussed below.
Applicants' Witness Antonio Vega states in his 9/I8 /84 Affidavit (Attachment I to Applicants' Supplement, Affidavit of Antonio Vega Concerning Board Questions Regarding OA/QC Oversight), at page 2:
"In response to the Board's request, an evaluation of all plant systems was conducted to determine the systems that fell into the category specified by the Board, as noted above. Ten systems / equipment groupings were identified. These systems are listed in Attachment B.
With regard to these systems, a thorough review was conducted to determine if all required inspections had been conducted and verified, as applicable. This review reflected that QC inspections have been performed and documented on the necessary mechanical, electrical and instrumentation components of these systems. These inspections include in-process inspections, final inspections. as-built verification inspections, and Authorized Nuclear Inspector ( ANI) inspections, as applicable. Continuing reinspections will be made as appropriate to preserve the integrity of completed inspections." (Emphases added.)
To begin with, it appears that rather than making a genuine attempt to ascertain which other systems might need to be looked at in addition to those specifically mentioned by the Board, Applicants have Instead used the Board's specified systems to limit their review.
Further, CASE challenges the statements made by Applicants' witness.
It is obvious that Mr. Vega could not have personally thoroughly reviewed 3
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1such documentation himself (and he does not claim that he did so),
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especially in light of the fact that he was testifying (and it is reasonable
.to assume, preparing to testify) in the. intimidation portion of these operating; license proceedings'during part of the just less than three weeks' time between the filing of the Board's Order (August 24, 1984) and the filing of Mr. Vega's Affidavit (September 13, 1984).
It is therefore reasonable to assume that Mr. Vega is relying heavily upon someone else's actual review of the systems in question.
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Further, based on the personal experiences of CASE's representatives and our witnesses, as well as on common sense and logic, regarding the amount of time necessary to thoroughly review such documentation, CASE submits that there is simply nct way that Mr. Vega could have had others make the necessary evaluation to identify plant systems, then thoroughly review the necessary documentation (such less for him to have done so himself. even to the extent that he would be able to state with certainty that his statements are accurate) in the less than three weeks' time. (Indeed, based upon recent past performance in regard to Applicants' ability to find, much.
less produce, documentation in the intimidation portion of the proceedings and in regard to Motions for Summary Disposition, one must seriously question Mr. Vega's statements.) Since no documentation was attached to support Applicants' assertions in this regard, CASE proposes a test of Applicants' assertions, as will be discussed later herein.
Mr. Vega further states (Affidavit at page 3):
"In addition, an extensive testing program on these systems has been implemented and will be completed prior to fuel load, including, as applicable, hydrostatic tests on pressure retaining systems, prerequisite testing on components to assure proper component 4
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Ig functional operability, and preoperational testing to assure proper operation.as a system. 'Preoperational testing provides assurance that the systems in question will operate as designed by requiring demonstration testing of the capability of the systems to meet safety-related performance requirements. A summary of the preoperational testing for each of the ten systems in question is set forth in Attachment C."
It'should be noted that Applicants' witness has not made even a minimal attempt to respond to the specific concerns raised by CASE regarding the neutron detectors and neutron detector wells at pages 19-23 in CASE's 8/18/84 Partial Answer (in Opposition to Applicants' Motion for Authorization to Issue a License to Load Fuel and Conduct Certain Precritical Testing and Motion for Additional Time to Respond). CASE submits that, absent documented proof to the contrary by Applicants, the Board should accept the documented information provided by CASE as being the current status of the items in question. Further, at least a portion of the matters discussed by CASE (item (1) on page 19) appears to have been I
confirmed by the NRC's Technical Review Team (TRT)'s request for additional
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information, attached to the 9/18/84 letter from Darrell G. Eisenhut, Director, Division of Licensing, NRR, to TUGCO's M. D. Spence, at page 7, item II.a.
(See Attachment A hereto.) f1/
The representations in Mr. Vega's Affidavit (page 3) regarding the
" extensive testing program" give the impression that (1) the testing will be completed prior to fuel load, and (2) all is well with the testing program.
However, it appears that the TRT Report calls into question certain statements made by Mr. Vega regarding the testing program. See Attachment A hereto, pages 11-14.
The information in the TRT Report takes on added
/1/ An inquiry from the Board to Mr. Ippolito's Technical Review Team regarding the amount of time necessary to thoroughly review documentation would in itself undoubtedly be very enlightening.
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N significance in light of the events leading up to and culminating in the Licensing Board's 10/1/84 Memorandum (Concerns About Start-Up Quality Assurance). /2/ We call the Board's attention especially.to item III.b.
Containment Integrated Leak Rate Testing (CILRT) at pages 13-14 of the TRT Report, which states, in part:
"Apparently after repairing leaks found during the first two attempts, the third attempt at a CILRT was successful. It was successfully completed after three electrical penetrations were isolated because the leakage through them could not be stopped. Though the leaks were subsequently repaired and individually tested with satisfactory results, NRC approval was not obtained to perform the-CILRT'with these penetrations isolated. In addition, leak rate calculations were.
performed using ANSI-ANS 56.8, which is neither_ endorsed by the NRC nor in accordance with FSAR commitments." - (Emphases sdded.)
"Accordingly, TUEC shall' identify to NRC any other differences in the conduct of the CILRT as a result of using ANSI /ANS 56.8 rather than ANSI N45.4-1972. Additionally, TUEC shall identify to NRC all other deviations from FSAR commitments."
' Although CASE applauds the fact that the TRT identified this problem, a reading of the transcript and attachments of the TRT's 9/18/84 meeting with Applicants is not quite so encouraging.- However, in any event, the Licensing Board must go beyond the additional information which the TRT is requiring Applicants to provide. The Board must also be concerned with how and why and at whose instigation this and other testing problems have been allowed to arise at this late date, as well as the extent of the OA/QC i
breakdown indicated by these events.
We also call the Board's attention to the attached 7/18/84 NRC Region IV Inspection and Enforcement (I&E) Report 84-21 advising of two Notices of Violation regarding preoperational testing and procedures, and the attached
/2/ We also invite the Board's attention once again to CASE's 10/13/83 (1)
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Motion to Add a New Contention, (2) Motion for Discovery, and (3) Offer of Proof, regarding hot functional and.other preoperational and acceptance tests.
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8/14/84 follow-up Report 84-21 (Attachments B and C hereto, respectively).
Although there are several portions of these reports which are pertinent to the issue at hand (and we urge that the Board read them in their entirety),
it should be espacially noted that on page 8 of the Appendix attached to the 8/14/84 letter (Attachment C hereto), it is stated, in part:
" Prior to issuance of this inspection report, the Notice of Violation was transmitted to the licensee as Severity Level IV Violation 445/8421-01. This is the second violation issued in recent weeks pertaining to lack of procedure compliance. The previous violation was identified as 445/8418-01 and contains three examples of failure to follow procedures. The licensee was made aware by the resident inspectors of the importance of decisive permanent corrective action by senior management to prevent future procedure violations as the pace of testing and operations increases at CPSES."
(Emphases added.)
Again, although CASE is gratified that the NRC inspectors identified these problems, the Licensing Board must go beyond the future corrective actions being requested by the inspectors. The Board must also be concerned with_how and why and at whose instigation this and other similar problems have been allowed to arise at this late date, as well as the extent of the OA/QC breakdown indicated by these events.
They also cast doubt on the statements made in Mr. Vega's Affidavit at page 3 regarding the " methods of documentation" which " assure positive control and tracking of such conditions to preclude inadvertent use of defective materials, components or systems," and about their timely resolution.
In addition, the Board should be aware that, contrary to the implications in Mr. Vega's affidavit regarding the status and adequacy of the testing program, the NRC Staff has stated that they will allow the Applicants to defer the following preoperational tests (see Attachment D hereto, 8/17/84 letter from B. J. Youngblood, Chief, Licensing Branch No. 1, 7
Division of Licensing, NRC, Washington, to M. D. Spence, TUGCO, under Subj ect: Acceptance of Preoperational Test Deferrals for Comanche Peak Steam Electric Station, Unit 1):
1.
Containment Cooling Systems 2.
Safety Injection System Check Valve Leakage 3.
Turbine Drive Auxiliary Feedwater Pump Steam Supply Line Check Valve and Drain Pot Level Control Valve 4.
Reactor Coolant Pump Seal Performance 5.
Thermal Expansion Testing 6.
Control Room Ventilation System We're sure the Board will be as relieved as CASE was to find out that, according to the Staff's proposed findings for inclusion in a future SER supplement (see Attachment D hereto, Enclosure page 3, second paragraph):
"The deferral of the thermal expansion retest is acceptable because it_
is consistent with approved industry practice on other plant test programs..." (Emphasis added.)
The Board should also be aware that the NRC Staff has recently approved an exemption from GDC 4 to the installation of jet impingement shields in Unit 1 (see Attachment E hereto, 8/28/84 letter from B. J. Youngblood, Chief, Licensing Branch No. 1, Division of Licensing, to M. D. Spence, TUGCO, under subject of: Applicants' Request for Exemption from a Portion of General Design Criterion 4 of Appendix A to 10 CFR Part 50 Regarding the Need to Analyze Large Primary Loop Pipe Ruptures as the Structure D gn Basis for Comanche Peak Steam Electric Station (Units 1 and 2), and attachment thereto; see also Attachuent F hereto, pages 35058-35061 of the 9/5/84 FEDERAL REGISTER, especially page 35060, right-hand column, second paragraph of item (6)).
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CASE also calls the Board's attention to the problems with the concrete which have been identified in the following:
1.
Technical Review Team (TRT) Report, Attachment A hereto, pages 7 and 8, item II,b. Falsification of Concrete Compression Strength Test Results.
2.
CASE's 9/10/84 Answer to Applicants' Motion for Summary Disposition Regarding Richmond Inserts, especially answer 8, pages 13-22, of Affidavit of CASE Witness Mark Walsh,' and Attachment D thereto -- regarding compressive strength of concrete at Comanche j
Peak.
3.
8/24/83 Deposition of Arvill "J. R." Dillingham, Jr., pages 65-70, sent to Board and parties attached to 3/7/84 01 Case No. 4-84-006, re: Alleged Intimidation of QC Personnel -- regarding Unit I stainless steel liner hollow places.
4.
NRC Staff's Request for Additional Information Concerning the Handling of Heavy Loads at Comanche Peak (Units 1 and 2) (see Attachment G hereto, letter from B. J. Youngblood, Chief, Licensing Branch No. 1, Division of Licensing, NRC, Washington, to M. D. Spence, TUGCO).
5.
There are also many other documents in the record regarding defective concrete, etc., which we have not yet thoroughly analyzed (as we did recently regarding the NCR's on the compressive strength of the concrete).
It should also be noted that neither Applicants nor NRC Staff have yet responded to the Board's directive to address the substantive portions of 9
i CASE's requests for 'information on drug-related terminations of OC inspectors, etc.- It is unknown, therefore, what the effect may be on-
. safety-related systems pertinent to Applicants' Motion to Load Fuel.
a.
In addition to the preceding and the items discussed in our 8/18/84 Partial Answer, CASE is currently working on additional pleadings which also should be considered in regard to Applicants' Motion to load fuel. We hope to have them in the hands of the Board and parties within the next week.
An addit'ional area of concern is the the adequacy and safety of the
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fuel pool, refueling cavity, transfer canal, stainless steel liner plates, etc., for both Units 1 and 2 and the transfer canal.
(See CASE's 9/27/84 Evidence of A Quality Control Breakdown, attachments thereto, and referenced documents.) In addition, the NRC Technical Review Team has expressed concern regarding alleged unauthorized cutting of rebar in the Fuel dandling Building (see Attachment A hereto, page 11, item II.e.).
As discussed herein, there are far too many unanswered (as well as unfavorably answered) questions for the Board to allow Applicants to load fuel at this point in time or even in the near future, since it will obviously take some time for these matters to be resol' red.
With regard to the fuel pool liners, transfer canal, etc., CASE moves that the Board order Applicants to provide CASE with complete documentation regarding the fuel pool, refueling cavity, transfer canal, stainless steel liner plates, Unit 1, etc. (similar to the documents provided recently in 10
the intimidation portion of these proceedings for only Unit 2).
As discussed on page 4 of this pleading, CASE proposes that this discovery serve the additional purpose of testing the representations made by Applicants' Witness Vega regarding the thorough review of documentation in h
response to the Board's 8/24/84 Order. This can be done quite simply by i
having CASE's witnesses and representatives keep up with the amount of time necessary to thoroughly review the documents and provide the Board and parties with an analysis of the results.
CASE further moves that the Board accept the following documents into the record (see detailed list at end of this pleading): Attachments A, B, C, and E hereto.
Respectfully submitted, Os/ m A [ O
( p.) Juanita Ellis, President GSE (Citizens Association for Sound Energy) 1426 S. Polk Dallas, Texas 75224 214/946-9446 Attachments:
Attachment A 9/18/84 letter from Darrell G. Eisenhut, Director, Division of Licensing, NRR, to TUGCO's M. D. spence, to which is attached the NRC's Technical Review Team (TRT) Report -- see page 5 of this pleading Attachment B NRC Inspection and Enforcement (I&E) Report 84-21 under cover letter of 7/18/84 -- see page 6 of this pleading Attachment C NRC Inspection and Enforcement (I&E) follow-up Report 84-21 under cover letter of 8/14/84 - see page 6 of this pleading 11 w
Attachment D 8/17/84 letter from B. J. Youngblood, Chief, Licensing Branch No. 1, Division of Licensing, NRC, Washington, to M. D.
Spence, TUGCO, under
Subject:
Acceptance of Preoperational Test Deferrals for Comanche Peak Steam Electric Station, Unit 1 -- see pages 7 and 8 of this pleading Attachment E 8/28/84 letter from B. J. Youngblood, Chief, Licensing Branch No. 1, Division of Licensing, to M. D. Spence, TUGCO, under subject of: Applicants' Request for Exemption from a Portion of General Design Criterion 4 of Appendix A to 10 CFR Part 50 Regarding the Need to Analyze Large Primary Loop Pipe Ruptures as the Structure Design Basis for Comanche Peak Steam Electric Station (Units 1 and 2), and attachment thereto -- see page 8 of this pleading Attachment F Pages 35058-35061 of 9/5/84 FEDERAL REGISTER, granting exemption discussed in Attachment E preceding -- see page 8 of this pleading Attachment G 9/21/84 letter from B. J. Youngblood, Chief, Licensing Branch No. 1, Division of Licensing, NRC, Washington, to M. D.
Spence, TUCCO -- see page 9 of this pleading 12
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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ATTACINENT A 4EP 13 004 Dockets: 50-445 50-446 Texas Utilities Electric Company Attn:
M. D. Spence, President TUGC0 Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201
Dear Mr. Spence:
SUBJECT:
COMANCHE PEAK REVIEW On July 9, 1984, the staff began dn intensive onsite effort designed to complete a portion of the reviews necessary for the staff to reach its decision regarding..the licensing of Comanche Peak Unit 1.
The onsite effcrt covered a number of areas, including allegations of improper construction practices at the facility.
The NRC assembled a Technical Review Team (TRT) responsible for evaluating most of the technical issues at Comanche Peak, including allegations. The TRT has recently identified a number of items that have potential safety implications for which we require additional information. These items are 3
listed in the enclosure to this letter. Further background information regarding these issues will be published in a Supplement to a Safety Evaluation Report (SSER), which will document the overall TRT's assessaient of the significance of the issues examined.
The items in the enclosure to this letter, which are in the general areas of electrical / instrumentation, civil / structural and test programs, cover only a portion of the TRT's effort. The TRT evaluation of items in the areas of mechanical, OA/QC, and coatings, and its consideration of the programatic implications of these findings, are still is progress. A summary of these issues will be provided to you at a later date.
You are requested to submit additional information to the NRC, in writing, including a program and schedule for completing a detailed and thorough assessment of the issues identified. This program plan and its implemen-tation will be evaluated by the staff before NRC considers the issuance of an operating license for Comanche Peak, Unit 1.
The program plan should address the root cause of each problem identified and its generic implic-ations on safety-related systems, programs, or areas. The collective significance of these deficiencies should also be addressed. Your program plan should also include the proposed TUGC0 action to assure that such problems will be precluded from occurring in the future.
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SEP 18 ;ga4 Mr. M. D. Spence This request is submitted to you in keeping with the NRC practice of promptly notifying applicants of outstanding information/ evaluation needs that could potentially affect the safe operation of their plant. Further requests for additional information of this nature will be made, if necessary, as the activities of the TRT progress.
' Sincerely.
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b.' I,(li G;. Eisenhut,'biYector
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': I 11-6 aire Division of Licensing, NRR
Enclosure:
As stated cc w/ enclosure See next page i
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COMANCHEPEAKj Mr. M. D. Spence President Texas Utilities Generating Company 400 N. Olive St., L.B. 81 Dallas, Texas 75201 cc: Nicholas S. Reynolds, Esq.
Mr. James E. Cunnins Bishop, Liberman, Cook, Resident Intpector/Conianche Peak i
Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W.
c/o U. S. Nuclear Regulatory Washington, D. C. 20036 Cosmission P. O. Box 38 Robert A'. Wooldridge, Esq.
Glen Rose, Texas 76043 Worsham, Forsythe, Sampels &
Wooldridge Mr. John T. Collins 2001 Bryan Tower, Suite 2500 U. S. NRC, Region IV Dallas, Texas 75201 611 Ryan Plaza Drive 1
Suite 1000 Mr. Homer C. Schmidt Arlington, Texas 76011 Manager - Nuclear Services Texas Utilities Generating Company Mr. Lanny Alan Sinkin 3kyway Tower 114 W. 7th, Suite 220 400 North Olive Street Austin, Texas 78701 L. B. 81 Dallas, Texas 75201 B. R. Clements Vice President Nuclear Mr. H. R. Rock Texas Utilities Generating Company Gibbs and Hill, Inc.
Skyway Tower 393 Seventh Avenue 400 North Olive Street New York, New York 10001 L. B. 81 Dallas, Texas 75201
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Mr. A. T. Parker Westingbcuse Electric Corporation William A. Burchette, Esq.
P. O. Ec> 355 1200 New Hampshire Avenue, N. W.
Pittsburgn, Pennsylvania 15230 Suite 420 Washington, D. C.
20036 Renea Hicks, Esq.
Assistant Attorney General Ms. Billie Pirner Garde Environmental Protection Division Citizens Clinic Director P. O. Box 12548, Capitol Station Government Accountability Project Austin, Texas 78711 1901 Que Street, N. W.
i Washington, D. C.
20009 Mrs. Juanita Ellis, President Citizens Association for Sound David R. Pigott, Esq.
Energy Orrick, Herrington & Sutcliffe 1426 South Polk 600 Montgomery Street Dallas, Texas 75224 San Francisco, California 94111 Ms. Nancy H. Williams Anthony Z. Roisman, Esq.
CYGNA Trial Lawyers for Public Justice 4
101 California Street 2000 P. Street, N. W.
San Francisco, California 94111 Suite 611 Washington, D. C. 20036
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ENCLOSURE'1 REQUEST FOR ADDITIONAL INFORMATION Electricel/Instrumertation Area a.
Electrical Cable Terminations The Technical Review Team (TRT) inspected random samples of sifety-related terminations, butt splices inside panels, and-n vendor-installed teminal lugs in General Electric (GE) motor control centers, and reviewed documentation relative to the installations, s
1.
The TRT found a lack of awareness on the part of quality control (QC) electrical inspectors to document in the inspection reports when the installation of the " nuclear heat-shrinkable cable insulation sleeves" was required to be witnessed, p ;/
Accordingly, TUEC shall clarify procedural requirements and W
provide additional inspector training with respect to the areas in which nuclear heat-shrinkable sleeves are required on splices and assure that such sleeves are installed where required.
2.
The TRT found inspection reports that did not indicate that the required witnessing of splice installation was done. Examples are as follows:
IR ET-1-0005393 IR ET-1-0005396 IR-ET-1-0005394 IR ET-1-0006776
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IR ET-1-0005395 IR ET-1-0014790 Accordingly, TUEC will assure that all QC inspections requiring witnessing for butt splices have been performed and properly documented; and verify that all butt splices are properly identified on the appropriate drawings and are physically identified within the appropriate panels.
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3.
The TRT found a lack of splice qualification requirements and provisions in the installation procedures to verify the 1
operability of those circuits for which splices were being used.
Accordingly, TUEC shall develop adequate installation / inspection procedures to assure that the wiring splicing materials are qualified for the appropriate service conditions, and that splices are not located adjacent to each other.
4.
Selected cable teminations were found that did not agree with I
their locations on drawings. Examples are as follows:
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, Panel CP1-ECPRCB-14, Cable E0139880 Panel CP1-ECPRTC-16, Cable E0110040 Panel CP1-ECPRTC-16, Cable E0118262 Panel CP1-ECPRTC-27, Cable EG104796 Panel CPX-ECPRCV-01, Cable EG021856 Panel CP1-ECPRCB-02, Cable NK139853 (nonsafety)
Accordingly, TUEC shall reinspect all safety-related and associated terminations in the control room panels and in the termination cabinets in the cable spreading room to verify that their locations are accuratel,y depicted on drawings. Should the results of this reinspection reveal an unacceptable level of nonconformance to drawings, the scope of this reinspection effort shall be expanded to include all safety-related and associated terminations at CPSES.
5.
The TRT found cases where nonconformance reports (NCRs) concerning vendor-installed terminal lugs in GE motor control centers had been improperly closed. Examples.are NCR Nos.
E-84-01066 through NCR E-84-01076, inclusive.
Accordingly, TUEC shall reevaluate and redisposition all NCRs related to vendor-installed terminal lugs in GE motor control centers.
b.
Electrical Equipment Separation The TRT reviewed the separation criteria between separate cables, trays <A conduits in the main control room and cable spreading room in Unis 1, and the compatibility of the electrical erection specifications with regulatory requirements. The TRT reviewed documentation and inspected random samples of separation between safety-related cables, trays and conduits and between them and nonsafety-related cables, trays and conduits.
1.
In numerous cases, safety-related cables within flexible l
conduits inside main control room panels did'not meet minimum separation requirements. Examples are as follows:
Panel CP1-EC-PRCB-02 Panel CP1-EC-PRCB-07 Panel CP1-EC-PRCP-06 Panel CPI-EC-PRCB-08 Panel CP1-EC-PRCB-09 Accordingly, TUEC shall reinspect all panels at CPSES, in addition to those in the main control room for Unit 1, that contain redundant safety-related cables within conduits, or l
safety and non-safety related cables within conduits, and either correct each violation of the separation criteria, or I
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. demonstrate by analysis the' acceptability of the conduit as a barrier for each case where the minimum separation is not met.
2.
In several cases, separate safety and nonsafety-related cables and safety and nonsafety-related cables within flexible conduits inside main control room panels did not meet minimum separation requirements (Table 1 identifies examples of these cases). No evidence was found that justified the lack of separation.
Accordingly, TUEC shall reinspect all panels at CPSES, in addition to those in the main control room of Unit 1, and either.
correct each violation of the separation criteria concerning separate cables and cables within flexible conduits, or demonstrate by analysis the adequacy of the flexible conduit as a barrier.
3.
The TRT found that the existing TUEC analysis substantiating the adequacy of the criteria for separation between conduits and cable trays had not been reviewed by the NRC staff.
Accordiagly, TUEC shall submit the analysis that substantiates the acceptability of the criteria stated in the electrical erection specifications governing the separation between independent conduits and cable trays.
4.
The TRT found two minor violations of the separation criteria inside panels CP1-EC-PRCB-09 and CPI-EC-PRCB-03 concerning a barrier that had been removed and redundant field wiring not meeting minimum separation. The devices involved with the barrier were FI-2456A, PI-2453A, PI-2475A, and IT2450, associated with Train A; ar' FI-2457A, PI-2454A,- PI-2476A, and IT-2451, associated with ir.sn B.
The field wiring was associated with devices HS-5423 of Train B and H3-5574, nonsafety-related.
Accordingly, TUEC shall correct two minor violations of the separation criteria inside panels CP1-EC-PRCB-09 and CP1-EC-PRCP-03 concerning a barrier that had been removed and redundant field wiring not meeting minimum separation.
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Table 1 Examples of Cases of Safety or Nonsafaty-Related Cables In Contact With Other Safety-Related Cables Within Conduits in Control Room Panels 1.
Control Panel CP1-EC-PRC8 Containment Spray System Cable No.
Train Related Instrument N
FlFeen) undetermined E0139010 A(orange)
Undetenmined 2.
Control Panel CPI-EC-PRCB Reactor Control System Cable No.
Train Related Instrument EG139383 Wgreen)
Reactor r.anual trip switch E0139311 A (orange)
Undetermined 3.
Control Panel CP1-EC-PRCP Chemical & Volume Control System Cable No.
Train Related Instrument EG139335 Wgreen)
LCV-Il2C E0139301 A (orange)
Undetermined 4.
Control Panel CPI-EC-PRCB Auxiliary Feedwater Control System Cable No.
Train Related Instrument E0139753 A(orange)
FK-2453A E0139754 A orang )
FK-2453B l
E0139756 8 green FK-2454A
)
EG139288 B green FK-24548 4
. c.
Electrical Conduit Supports The TRT examined the nonsafety-related conduit support installation in selected seismic Category I areas of the plant. The support installation for non-safety related conduits less than or equal to 2 inches was inconsistent with seismic requirements and no evidence could be found that substantiated the adequacy of the installation for nonsafety-related conduit of any size. According to Regulatory Guide 1.29 and FSAR Section 3.78.2.8, the seismic Category II and nonseismic items should be designed in such a way that their failure would not adversely affect the function of safety-related components or cause injury to plant personnel.
Accordingly, TUEC shall propose a program that assures the adequacy of the seismic support system installation for nonsafety-related conduit in all seismic Category I areas of the plant as follows:
1.
Provide the results of seismic analysis which demonstrate that all nonsafety-related conduits and their support systems, satisfv the provisions of Regulatory Guide 1.29 and FSAR Section 3.7B.2.8.
2.
Verify that nonsafety-related conduits less than or equal to 2 inches in diameter, not installed in accordance with the requirements of Regulatory Guide 1.29, satisfy applicable design requirements.
d.
Electrical QC Inspector Training / Qualifications The TRT examined electrical QC inspectob training and certification files, and requirements for personnel testing, on-the-job training, and recertification. The TRT also interviewed selected electrical QA/QC personnel.
1.
The TRT found a lack of supportive documentation regarding personnel qualificatices in the training and certification files, as required by procedures and regulatory requirements.
Also, the TRT found a lack of documentation for assuring that the requirements for electrical QC inspector recertification were being met. Specific examples are:
One case of no documentation of a high schoc diploma or General Equivalency Diploma.
One case of no documentation to waive the remaining 2 months of the required 1 year experience.
'One case where a QC technician had not passed the required color vision examination administered by a i
professional eye specialist. A makeup test using colored I
pencils was administered by a QC supervisor, was passed, and then a waiver was given.
Two cases where the experience requirements to become a Level 1 technician were only marginally met.
One case of no documentation in the training and certification files substantiating that the person met the experience requirements.
Accordingly TUEC shall review all the electrical QC inspector training, qualification, certific'ation and recertification files against the project requirements and provide the infomation in such a form that each requirement is clearly shown to have been met by each inspector. If an inspector is found to not meet the training, qualification, certification, or recertification requirements. TUEC shall then review the records to detemine the adequacy of inspections madt by the unqualified individuals and provide a statement on the impact of the deficiencies noted on the safety of the project.
2.
The TRT found a lack of guidelines and procedural requirements for the testing and certifying of electrical QC inspectors.,,Specifically, it was found that:
No time limit or additional training requirements existed between a failed test and retest.
No controls existed to assure that the same test would not be given if an individual previously failed that test.
No consistency existed in test scoring.
No guidelines or procedures were available to control the disqualification of questions from the test.
No program was available for establishing new tests (except when procedures changed). The same tests had been utilized for the last 2 yetrs.
Accordingly TUEC shall develop a testing program for electrical QC inspectors which provides adequate administrative guidelines, procedural requirements and test flexibility to assure that suitable proficiency is achieved and maintained.
. The deficiencies identified with the electrical QC inspections have generic implications to other construction disciplines. The implications of these findings will be further assessed as part of the overall programmatic review of QC inspector training and qualification and the results of this review will be reported under the QA/QC category on " Training and Qualification."
II. Civil / Structural' Area i
a.
Unable to Justify Reinforcing Steel Omitted in the Reactor Cavity The TRT investigated a documented occurrence in which reinforcing steel was omitted from a Unit I reactor cavity concrete placement between the 812-foot and 819-foot 1-inch elevations. This reinforcement was installed and inspected according to drawing 2323-51-0572, Revision 2.
~
However, after the concrete was placed.
l Revision 3 to the drawing was issued showing a substantial increase in reinforcing steel over that which was installed. Gibbs & Hill Engineering was infomed of the omission by Brown & Root i
Nonconformance Report CP-77-6. Gibbs & Hill Engineering replied that the omission in no way impaired the structural integrity of the structure. Nevertheless, the additional reinforcing steel was added as a precaution against cracking which might occur in the vicinity of the neutron detector slots should a loss of coolant accident (LOCA) occur. A portion of the omitted reinforcing steel was also placed in the next concrete lift above the 819-foot 1-inch level. This was done to partially compensate for the reinforcing steel omitted in the previous concrete lift and to minimize the overall area potentially subject to cracking.
The TRT requested documentation indicating that an analysis was perfomed supporting the Gibbs & Hill conclusion. The TRT was 1
subsequently informed that an analysis had not been performed.
Therefore, the TRT cannot detemine the safety significance of this issue.until an analysis is perfomed verifying the adequacy of the reinforcing steel as installed.
Accordingly TUEC shall provide an analysis of the as-built condition of the Unit I reactor cavity that verifies the adequacy of the reinforcing steel between the 812-foot and 819-foot 1-inch elevations. The analysis shall consider all required load combinations.
b.
Falsification of Concrete Compression Strength Test Results The TRT investigated allegations that concrete strength tests were falsified. The TRT reviewed an NRC Region IV investigation (IE Report No. 50-445/79-09;50-446/79-09) of this matter that included
. interviews with fifteen individuals. Of these, only the alleger and one other individual stated they thought that falsification occurred, but they did not know when or by whom. The TRT also reviewed slump and air entrainment test results of concrete placed during)the period the alleger was employed (January 1976 to February 1977 and did not find any apparert variation in the unifomity of the parameters for concrete placed during this period.
Although the uniformity of the concrete placed appears to minimize 4
the likelihood that low concrete strengths were obtained, other I
allegations were raised concerning the falsification of records associated with slump and air content tests. The Region IV staff addressed these allegations by assuming that concrete strength test results were adequate. Furthermore, a number of other allegations dealing with concrete placement problems (such as deficient aggregate grading and concrete in the mixer too long) were also resolved by assuming that concrete strength test results were adequate. The TRT agrees with Region IV that, while the preponderance of evidence suggests that falsification of results did not take place, the matter cannot be resolved completely on the basis of concrete strength test results, especially if there is any doubt about whether they may have been falsified.
Due to the importance of the concrete i
strength test results, the TRT believes that additional action by TUEC is necessary to provide confimatory evidence that the reported concrete strength test results are indeed representative of the strength of the concrete installed in the Category I concrete structures.
Accordir. gly,TUECshalidetermineareaswheresafety-relatedconcrete was placed between January 1976 and February 1977, and provide a program to assure acceptable concrete strength. The program shall include tests such as the use of random Schmidt hansner tests on the concrete in areas where safety is critical. The program shall 4
include a comparison of the results with the results of tests per-formed on concrete of the same design strength in areas where the strength of the concrete is not questioned, to determine if any significant variance in strength occurs. TUEC ; hall submit the program for performing these tests to the NRC for review and approval prior to pt:rforming the tests.
c.
Maintenance of Air Gap Between Concrete Structures The TRT investigated ' t requirements to maintain an air gap between concrete structures-
- .ased on the review of available inspection reports and related documents, on field observations, and on discussions with TUEC engineers, the TRT cannot determine whether an adequate air gap has been provided between concrete i
structures. Field investigations by B&R QC inspectors indicated unsatisfactory conditions due to the presence of debris in the air l
l i
l
~
.g-gap, such as wood wedges, rocks, clumps of concrete and rotofoam.
N disposition of the NCR relating to this matter states that the
" field investigation reveals that most of the material has been removed." However, the TRT cannot determine from this report (NCR C-83-01067) the extent and location of the debris remaining between the structures.
Based on discussions with TUEC engineers, it is the TRT's understanding that field investigations were made but that no permanent records were maintained.
In addition, it is not apparent that the permanent installation of elastic joint filler material
("rotofoam") between the Safeguards Building and the Reactor Building, and below grade for the other concrete structures, is consistent with the seismic analysis assumptions and dynamic models used to analyze the buildings, as these analyses are delineated in the Final Safety Analysis Report (FSAR). The TRT, therefore, concludes that TUEC has not adequately demonstrated compliance with FSAR Sections 3.4.1.1.1, 3.8.4.5.1, and 3.7.8.2.8, which require separation of Seismic Category I buildings to prevent seismic interaction during an earthquake.
Accordingly. TUEC shall:
1.
Perform an inspection of the as-built condition to confirm that adequate separation for all seismic category I structures has been provided.
2.
Provide the results of analyses which demonstrate that the presence of rotofoam and other debris between all concrete conditions)(does not result in any sigr. ficant increase in structures as determined by inspections of the as-built seismic response or alter the dynamic response characteristics of the Category I structures, components and piping when compared with the results of the original analyses.
d.
Seismic Design of Control Room Ceiling Elements s
The TRT investigated the seismic design of the ceiling elements installed in the control room. The following matrix designates those ceiling elements present in the control room and their seismic category designation:
i
---,,-,---s
,.w,-
,,,,v-,
e-.,
at,-,
.,-..---------- -,-..,-,-. -- -- - - ~- -
e-
. 1.
Heating, Ventilating and Air Conditioning
- Seismic Category I 2.
Safety-Related Conduits
- Seismic Category I 3.
Nonsafety-Related Conduits
- Seismic Category II 4.
Lighting Fixtures
- Seismic Category II 5.
Sloping Suspended Drywall Ceiling
- Non-Seismic 6.
Acoustical Suspended Ceiling
- Non-Seismic i
7.
Lowered Suspended Ceiling
,Non-Seismic according to Regulatory Guide 1.29 and FSAR Section 3.7B.2.8, the seismic Category II and nonseismic items should be designed in such a way that their failure would not adversely affect the functions of safety-related components or cause injury to operators.
For the nonseismic items (other than the sloping suspended drywall ceiling), and for nonsafety-related conduits whose diameter is 2 inches or less, the TRT could find no evidence that the possible effects of a failure of these items had been i
considered. In addition, the TRT determined that calculations for seismic Category II components (e.g., lighting fixtures) and the calculations for the sloping suspended drywall ceiling did not adequately reflect the rotational interaction with the nonseismic items, nor were the fundamental frequencies of the supported 4
masses determined to assess the influence of the seismic response spectrum at the control room ceiling elevation would have on J
the seismic response of the ceiling elements.
Accordingly, TUEC shall provide:
1.
The results of seismic analysis which demonstrate that the nonseismic items in the control room (other than the sloping suspended drywall ceiling) satisfy the provisions of i
Regulatory Guide 1.29 and FSAR Section 3.78.2.8.
2.
An evaluation of seismic design adequacy of support systems for the lighting fixtures (seismic Category II) and the suspended drywall ceiling (nonseismic item with modification) which accounts for pertinent floor response characteristics of the systems.
3.
Verification that those items in the control room ceiling j
not installed in accordance with the requirements of Regulatory Guide 1.29 satisfy applicable design requirements.
1 4.
The results of an analysis that justify the: adequacy of the nonsafety-related conduit support system in the control room for conduit whose diameter is 2 inches or less.
1
^
- 5.
The results of an analysis which demonstrate that the foregoing problems are not applicable to other Category II and nonseismic structures, systems and components elsewhere in the plant.
e.
Unauthorized Cutting of Rebar in the Fuel Handling Building The TRT investigated an alleged instance of unauthorized cutting of rebar associated with tfie installation of the trolley process aisle rails in the Fuel Handling Building. The claim is that during installation of 22 metal plates in January 1983, a core drill was used to drill about 10 holes approximately 9 inches deep. The TRT reviewed the reinforcement drawings for the Fuel Handling Building and detemined that there were three layers of reinforcing steel in the top reinforcement layer of the slab. This reinforcement layer consisted of a No. 18 bar running in the east-west direction in the first and third layers, and a No. 11 bar running in the north-south direction on the second layer. The review also revealed that the layout of the reinforcement and the trolley rails was such that the east-west reinforcement would interfere with the drilling of holes along only one rail location. However, if 9-inch holes were drilled, both the first and third layers of No.18 reinforcement would be cut.
Design Change Authorization No. 7041 was written for authorization to cut.the uppermost No.18 bar at only one rail location, but did not reference authorization to cut the lower No.18 bar. DCA-7041 also stated that the expansion bolts and base plates may be moved in the east-west direction to avoid interference with reinforcement running in the north-south direction. The information, described in DCA-7041, was substantiated by Gibbs & Hill calculations. If the ten holes were actually drilled 9 inches deep, then the allegation that the reinforcement was cut without proper authorization would be valid.
Accordingly, TUEC shall provide:
1.
Information to demonstrate that only the No. 18 reinforcing steel in the first layer was cut, or l
2.
Design calculations to demonstrate that structural integrity is maintained if the No.18 reinforcing steel on both the first and third layers was cut.
I III. Test programs Area a.
Hot Functional Testing (HFT)
The TRT reviewed a sample of the completed data packages for HFT preoperational test procedures, pertinent startup administrative procedures, NRC inspection reports, and the preoperational test index and its schedule. The TRT also inspected test deficiency reports l
e
. (TDRs) that were generated as a result of test deficiencies found prior to and during HFT.
1.
Chapter 14 of the FSAR and Regulatory Guide 1.68 provide requirements for the conduct of preoperational testing.
In reviewing test data packages, the TRT found that certain test objectives were not met. It appears that the Joint Test Group approved incomplete data packages for at least three preoperational hot functinal tests. These were:
Test Procedure Deficiency ICP-PT-02-12. " Bus Because acceptable voltages Voltage and Load Survey" could not be achieved with the specified transformer taps, they were changed. A subsequent engineering evaluation required returning to the original taps, but no retest was performed.
ICP-PT-34-05, " Steam Level detectors 1-LT-517, 518 l
Generator Narrow Range and 529 were replaced with Level Verification" temporary equipment of a design that was different from that which was to be eventually installed i
1CP-PT-55-05 Level detector 1-LT-461 appeared
" Pressurizer Level to be out of calibration during the Control" test and was replaced after the test.
The retest approved by the JTG was a cold calibration rather than a test consistent with the original test objective, which was to obtain satisfactory data under hot conditions.
Accordingly. TUEC shall review all complete preoperational test data packages to ensure there are no other instances where test objectives were not met, or prerequisite conditions were not satisfied. The three items identified by the TRT shall be included, along with appropriate justification, in the test deferral packages presented to the NRC.
4
. 2.
The TRT noted during a review of HFT completed test data that the JTG did not approve tne data until after cooldown from the test. The tests are not considered complete until this approval
)
is obtained.
In order to complete the proposed post-fueling, i
deferred preoperational HFT, the JTG, or a similarly qualified 1
group, must approve the data prior to proceeding to initial criticality. The TRT did not find any document providing assurance that TUEC is comunitted to do this.
I Accordingly, TUEC shall commit to having a JTG, or similarly qualified group, review and approve all post-fueling preoperational test results prior to declaring the system operable in accordance with the technical specifications.
3.
The TRT pointed out that in order to conduct preoperational tests at the necessary temperatures and pressures after fuel load, certain limiting conditions of the proposed technical specifications cannot be met, e.g., all snubbers will not be operable since some will not have been tested.
Accordingly, TUEC shall evaluate the required plant conditions for the deferred preoperational tests against limiting conditions in the proposed technical specifications and obtain NRC approval where deviations from the technical specifications are necessary.
4.
Data for the thermal expansion tests (which have not yet been approved by the JTG) did not provide for traceability between the calibration of the measuring instruments and the r:anitored locations, as required by Startup Administrative Procedure-7.
The infonnation was separately available in a personal log held by Engineering.
Accordingly, TUEC shall incorporate the information necessary to provide traceability between thennal expansion test monitoring locations and measuring instruments.
TUEC shall also establish administrative controls to assure appropriate test and measuring equipment traceability during future testing.
b.
Containment Intergrated Leak Rate Testing (CILRT) t The TRT reviewed the data package for the CILRT performed on Unit 1, and discussed the conduct of the test with TUEC and NRC personnel who participated in or witnessed it.
i 1
l l
l
. Apparently after repairing leaks found during the first two attempts, the third attempt at a CILRT was successful. It was successfully completed after three electrical penetrations were isolated because the leakage through them could not be stopped.
Though the leaks were subsequently repaired and individually tested with satisfactory results, NRC approval was not obtained to perfom the CILRT with these penetrations isolated. In addition, leak rate calculations were performed using ANSI /ANS 56.8, which is neither endorsed by the NRC nor in accordance with FSAR count tments.
Accordingly. TUEC shall identify to NRC any other differences in the conduct of the CILRT as a result of using ANSI /ANS 56.8 rather than ANSI N45.4-1972. Additionally. TUEC shall identify to NRC all other deviations from FSAR commitments, c.
Prerequisite Testing The TRT reviewed FSAR connitments, startup administrative procedures, prerequisite test records, craft personnel qualification records, and discussed them with startup and craft management personnel. The TRT also observed test support craft personnel at work and interviewed some of them to gain familiarity with their attitudes and capabilities.
The review of test records revealed that craft personnel were signing to verify initial conditions for tests in violation of startup Administrative Procedure-21, entitled: " Conduct of Testing" (CP-SAP-21). This procedure requires this function to be perforined by System Test Engineers (STE). Startup management had issued a memorandum improperly authorizing craft personnel to perfom these verifications on selected tests.
Accordingly, TUEC shall rescind the startup memorandum (STM-83084),
which was issued in conflict with CP-SAP-21, and ensure that no other i
memoranda were issued which are in conflict with approved procedures, d.
Preoperational Testing l
The TRT assessed the preoperational test program by reviewing i
administrative procedures, interviewing startup personnel, and examining test records, schedules, system assignments, subsystem definition packages, and the master data base.
Problems found with test data are addressed in section III.a of this enclosure. The TRT also found that STEs were not being provided with current design information on a routine, controlled basis, and had to update their own material when they considered it appropriate.
Accordingly, TUEC shall establish measures to provide greater i
assurance that STEs and other responsible personnel are provided with current controlled design documents and change notices.
t
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j y.7 NUCLEAR REGULATORY COMMISSION 2
REoioN w 611 RYAN PLAZA DRIVE, SUITE 1000 k,... 8 ARUNGToN, TEXAS 70011 July 18, 1984 In Reply Refer To:
Docket: 50-445/84-21 ATTACH 1ENT B Texas Utilities' Electric Company Attn:
M. O. Spence, President, TUGC0 Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201 Gentlemen:
This refers to the inspection conducted by Mr. W. F. Smith of this office during the period June 14-16, 1984, of activities authorized by NRC Construction Permit CPPR-126 for the Comanche Peak Facility, Unit 1, and to the discussion of our findings with Mr. J. T. Merritt and other members of your staff.
Areas examined during the inspection included preoperational test witnessing and test procedure review. Within these areas, the inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observations of testing in progress by the inspector.
During this inspection, it was found that certain of your activities were in violation of NRC requirements. Consequently, you are required to respond to these violations in writing, in accordance with the provisions of 2
l Section 2.201 of the NRC's " Rules of Practice," Part 2, Title 10, Code of,
Federal Regulations. Your response should be based on the specifics contained in the Notice of Violation enclosed with this letter.
I
^
Details of this inspection will be included in a report to be issued in the near future and identified as NRC Inspection Report 50-445/84-21.
This office calls your attention to the fact that similar violations were addressed during the exit meeting of June 1,1984, which will be documented in NRC Inspection Repcrt 50-445/84-18. This indicates that actions taken thus far do not appear to have been effective with regard to procedure compliance.
i i
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-.-,--,-,.-,-,,,.-.,.,__+y
,w-
-y., _ _. - -,,,
Texas Utilities Electric Company July 18, 1984 The response directed by this letter and the accompanying Notice is not subject
.to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, P1 96-511.
Should you have any question concerning this inspection, we will be pleased to discuss them with you.
Sincerely, lC tc dv tL 1 Richard L. Bangart, Dir '
or Olvision of Radiation S ety and Safeguards
Enclosure:
Appendix - Notice of Violation cc w/ enclosure:
Texas Utilities Electric Company ATTN:
H. C. Schmidt, Manager Nuclear Services Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201 Texas Utilities Electric Company ATTN:
B. R. Clements, Vice Presicent, Nuclear Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201
')
APPENDIX-NOTICE OF VIOLATION E
Texas Utilities Electric Company Docket: 50-445/84-21
- Comanche Peak Steam Electric Station Construction Permit: CPPR-126 3
' Based on the results of an NRC inspection conducted'during the period of i
June 14-16, 1984,'and in accordance with the NRC Enforcement Policy (10 CFR Part 2, Appendix C), 49 FR-8583, dated March 8, 1984, the following violations were identified:
1.
Criterion V of Appendix B to 10 CFR 50 states, in part, " Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings...."
1 Con *.rary to the above, on June 16, 1984, an operator proceeded to partially open Station Service Water Chlorination Valve XSW-042 in violation of Step 5.4.1.6 of System Operating Procedure SOP-501A (Rev. 0),
i
" Station Service Water System," which requires XSW-036 to be opened. The l
operation was aborted and the valve restored to the shut position only after j
the NRC inspector pointed out the procedure violation.
Subsequently, it was determined that the procedure was in error, thus was changed accordingly:
and the operation resumed by opening Valve XSW-042.
j This is a Severity Level IV Violation.
(Supplement II-0) (445/8421-01)
I 2.
Criterion XI of Appendix B to 10 CFR 50 states in part, "... the test i
program shall include, as appropriate, proof test prior to installation, preoperational tests, and operational tests during nuclear power plant or fuel reprocessing plant operation of structures, systems, and components.
Test procedures shall include provisions for assuring that all prerequisites for the given test have been met,
)
a.
Contrary to the above, during the performance of the Olesel Generator Control Circuit Functional and Start Test, 1CP-PT-29-02 RT-1, the NRC inspector noted that there was no prerequisite in the test procedure to provide for station service air so that Step 7.1.6.7 can be per-formed to operate the barring device, which requires service air to function. This became apparent to the NRC inspector when he noticed j
the service air piping was not connected to the barring device.
In lieu of service air, the STE utilized temporary air from a portable air compressor, which is not addressed by the procedure.
7
+
i L
t i
b.
Contrary to the above, the station service water flow balancing test procedure, ICP-PT-04-01, had no prerequisite requirement to ensure the flow gages used during Step 7.8 (Flow Adjusteent) were properly
. filled and vented.
Failure to fill and vent these detectors just prior to flow adjustment can cause erroneous flow gage indications.
This can place the flow data in question. As a result, during conduct of Step 7.8 of the' test, the service water flow gage for.
containment spray was pegged high with no flow.
It was evident that the gage was malfunctioning due to air binding or other mechanical problem.
This is a Severity Level IV Violation.
(Supplement II-E) (445/8421-02)
Pursuant to the provisions of 10 CFR 2.201, Texas Utilities Electric Company is hereby required to submit to this office within 30 days of the date of this Notice, a written statement or explanation in reply, including: (1) the corrective steps which have been taken and the results achieved; (2) corrective steps which will be taken to avoid further violations; and (3) the date when full compliance will be achieved. Consideration may be given to extending your response time for good cause ' hown.
s Dated:
July 18,1984
' 9 Caa p[" "'%,
UNITE) STATES k'
t NUCLEAR REGULATORY COMMISSION y
e,,
,. I REoioN IV
[
sit RYAN PLAZA drive SulTE 1000
,8 ARLINGTON. TEXAS 70011 August 14, 1984 In Reply Refer To:
Docket:
50-445/84-21 ATTACINENT C Texas Utilities Electric Company ATTN:
M. D. Spence, President, TUGC0 Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201 Gentlemen:
This refers to the inspection conducted by Messrs. D. L. Kelley and W. F. Smith of this office during the period June 1-30, 1984, of activities authorized by NRC Construction Permit CPPR-126 for the Comanche Peak Facility, Unit 1, and to the discussion of our findings with Messrs. B. R. Clements and J. C. Kuykendall and other members of your staff at the conclusion of the inspection.
Areas examined during the inspection included: (1) plant procedures inspection; (2) preoperational test witnessing; and (3) plant tours.
Wit'11n these areas, the inspection consisted of selective examination of procedures and representative records, interviews with personnel, and observations by the inspectors.
These findings are documented in the enclosed inspection report.
l During this inspection, it was found that certain of your activities were in violation of NRC requirwments.
The two violations reported in paragraph 3 of the enclosed inspection report were forwarded to you by our letter and Notice of Violation, dated July 18, 1934; therefore, this letter does not require further written response.
In accordance with 10 CFR 2.790(a), a copy of this letter and the enclosure will be placed in the NRC Public Document Room unless you notify this office, by telephone, within 10 days of the date of this letter, and submit written application to withhold information contained therein within 30 days of the date of this letter. Such application must be consistent with the requiremerits of 2.790(b)(1).
r Texas Utilities Electric Company August 14, 1984 Should you have any questions concerning this inspection, we will be pleased to discuss them with you.
Sincerely, (bCl(( L Ba.ty.>f l
Richard L. Bangart, Dir tor RIV Comanche Peak Tack Force
Enclosure:
Appendix - NRC Inspection Report 50-445/84-21 cc w/ enclosure:
Texas Utilities Electric Company ATTN:
H. C. Schmidt, Manager Nuclear Services Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201 Texas Utilities Electric Company ATTN:
B. R. Clements, Vice President, Nuclear Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201
APPENDIX U. S. NUCLEAR REGULATORY COMMISSION REGION IV NRC Inspection Report: 50-445/84-21 Construction Permit: CPPR-126 Docket: 50-445 Category:
A2 Licensee: Texas Utilities Electric Company (TUEC)
Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas
.75201 Facility Name:
Comanche Peak Steam Electric Station (CPSES), Unit 1 Inspection At: Glen Rose, Texas Inspection Conducted: June 1-30 1984 Inspectors:
/
9 D.L.Kelley,5fnior/ResidentReactorInspector Dft/
I (SRRI) (paragraphs 1, 3, 4, and 5)
M/
dike W. F. Smith, Resident Reactor Inspector (RRI)
Date (paragraphs 1, 2, 3, and 5)
Approved:
k f!
8f D. M. Hunnicutt, Team Leader Task Force Date Inspection Summary Inspection Conducted: June 1-30, 1984 (Report 50-445/84-21)
Areas Inspected:
Routine, announced inspection of (1) plant procedures inspection; (2) preoperational test witnessing; and (3) plant tours.
The inspection involved 180 inspector-hours onsite by two NRC inspectors.
Results: Within the areas inspected, two violations were identified and were transmitted under separate cover to the licensee on July 18, 1984, as Severity Level IV Violations, 445/8421-01, Supplement II-D, and 445/8421-02, Supplement II-E.
2 DETAILS 1.
Persons Contacted Licensee Personnel
- B. R. Clements, Vice President, Nuclear Operations
- J. C. Kuykendall, Manager, Nuclear Operations
- R. A. Jones, Manager, Plant Operations J. T. Merritt, Assistant Project General Manager
- J. H. Roberts, Construction Startup Turnover Surveillance Supervisor
- T. P. Miller, Lead Startup Engineer
- R. B. Seidel, Operations Superintendent
- H. A. Lancaster, Startup Quality Assurance Specialist
- J. C. Smith, Quality Assurance
- T. L. Gosdin, Support Services' Superintendent
- D. E. Deviney, Operations Quality Assurance Supervisor
- S. M. Franks, Startup Special Projects R. R. Wistrand, Administrative Superintendent J. Moorefield, Office Services Coordinator CPSES D. C. Hisey, System Test Engineer J. A. Van Gulik, System Test Engineer K. B. Becker, System Test Engineer K. E. Hemmila, System Test Engineer S. E. Harvey, Assistant Shift Supervisor R. L. Fortenberry, Shift Supervisor M. S. Harris, System Test Engineer M. Smith, Shift Supervisor R. Beck, System Test Engineer Others The NRC inspectors also interviewed other licensee employees during this inspection period.
- Denotes those present during the exit interview.
2.
Plant Procedures Inspection The objective of this inspection was to determine that the scope of the plant procedures system is adequate to control safety-related operations within applicable regulatory requirements and to determine the adequacy of management controls in implementing and maintaining a viable procedure system.
The first segment of this inspection module was accomplished during the period March 1 through April 30, 1984.
The results of the inspection are
. described in NRC Inspection Report 50-445/84-15, dated July 3, 1984. The plant procedures inspection was completed June 30, 1984.
Detailed operating, emergency, and maintenance procedure inspections will be conducted separately and reported in subsequent inspection reports.
The following procedures were reviewed during this inspection 1
period:
4 STA-605 (Rev.3)
" Clearance and Safety Tagging" STA-707 (Rev.1)
" Safety Evaluations" STA-606 (Rev.3)
" Maintenance Action Requests" STA-608 (Rev.5)
" Control of Measuring and Test Equipment" STA-616 (Rev.0)
" Control Room and Observation Area Access" SOP-609A (Rev.0)
" Diesel Generator System" SOP-501A (Rev.0)
" Station Service Water System" 00A-202 (Rev.2)
" Preparation of System Operating Procedures" i
00A-301 (Rev.3)
" Operating Logs" The RRI verified that responsibilities have been assigned to assure that site procedures such as those listed above are reviewed, updated, approved, and that 10 CFR 50.59 considerations are included in the review.
In addition, the NRC inspector verified that when special orders are used, administrative controls have been established that provide a mechanism for their review, issuance, distribution, and limitations for use.
The RRI interviewed a reactor shift supervisor to determine whether or not I
he understood the systems established for controlling temporary changes to procedures.
Several pertinent questions were asked, and all of the answers provided by the shift supervisor were correct.
As the NRC inspector witnessed the daily progress of preoperational testing, he noted that the shift supervisors as a group were sufficiently aware of the established systems.
The NRC inspector reviewed the above listed procedures to ensure that' The review, approval, and updating had been done in accordance with station administrative requirements.
The issuance and superseding of the procedures were done in accordance with the established controls.
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I l The procedures were formatted properly.
k The procedures were free of typographical errors, conflicts, or l
editorial errors.
The procedures were adequate for the intended purpose and scope.
The working copy at key locations such as the control room is the latest revision.
Upon completion of the above review, the NRC inspector did not identify any significant deficiencies; however, the following comments were offered to the licensee for consideration to reduce the possibility of problems in the future:
J STA-605 Section 4.1.9 does not require independent verification of danger tags for nonsafety-related systems that do not affect the safe operation of the plant. Discussion with shift supervisors revealed a tendency on their part to be conservative in actual practice and require such verification checks on most, if not all, tagouts. The NRC inspector stated to representatives of the licensee that some plants require independent verification checks in those situations where operating system pressure, temperature, electrical, or radiological conditions could result in equipment damage or injury to personnel. Usually the pressure is defined as greater than 150 pounds per square inch gage and/or temperatures greater than 200 degrees Fahrenheit. The licensee agreed to consider this matter for the next revision of STA-605.
Section 4.1.9 states that the hanging of danger tags "should not normally" be done simultaneously with the independent verification check. The NRC inspector recommended that the statement be changed to "shall not." The NRC inspector was concerned that the power of suggestion in watching a tag being hung on a component could lead the verifier into believing the component was correct instead of the verifier independently determining the component was correct.
This can defeat the concept of independent verification.
The NRC inspector noted that Attachment 1 of STA-605, " Clearance Report,"
did not have a column for the independent verification of tag removal and restoration of each component in accordance with Section 4.2.2.3.
This action is not documented except by a single signature. Such a column would be a good tool to help the verifier check off each item, and would provide better assurance to the snift supervisor that none were inadvertently overlooked.
. The above comments were discussed with the licensee's representative, who indicated that the comments are under consideration and that some of them are already incorporated into a major rewrite of STA-605 that is currently underway.
In particular, the attachment, such as the clearance report, has been improved significantly to incorporate such features as verification columns discussed above.
STA-707 Section 4.1.5 does not clearly implement the requirement of 10 CFR 50.59(b) to publish a periodic (at least annual) report of changes made as permitted by 10 CFR 50.59(a). During discussion between the RRI and the licensee's representative, two major points relative to this report were brought out by the RRI.
IE Circular 80-18 dated August 22, 1980, clarifies the NRC requirements for the report.
In short, the Circular points out that, for all cases requiring a written safety evaluation, the safety evaluation must set forth the bases and criteria used to determine that the proposed change does not involve an "unreviewed safety question." Though the annual report can be brief, a simple statement of conclusion in itself is not sufficient. The regulation requires a summary of the safety evaluation.
Changes made under the authority of 10 CFR 50.59(a) are reportable and should appear in the annual report, if a change in the facility or procedure generates a need for revision of any of the text or drawings in the current Safety Analysis Report (SAR).
i In addition, tests or experiments not described in the current SAR shall also be reported if they are to be added to the SAR.
Section 4.1.5 of l
STA-707 should more clearly address the requirements of 10 CFR 50.59.
00A-202 Section 4.2.6.1 of 00A-202, Revision 2, requires the " Instructions" section of system operating procedures to be subdivided into specific evolutions of operation.
Examples are, "Startup," " Normal Operations,"
" Shutdown," and " Draining." Because of the differences between systems it is not practical to use the subsections specified by 00A-202.
Consequently, some standard operating procedures (SOPS)do not follow the o
formatting requirement, such as 50P-503, " Surface Water Pretreatment System," SOP-607A, "118 VAC Distribution System and Inverters," SOP-706,
" Digital Radiation Monitoring System," and 50P-710. "Incore i
Instrumentation System." The NRC inspector discussed this with the l
. licensee, who stated that there is a major rewrite of SOPS in progress, and formatting problems such as this are being corrected.
Since these procedures are scheduled for NRC review after publication, the NRC inspector indicated that this area would be reinspected at a later date.
There were no other concerns, deficiencies, or violations noted during the procedures inspection.
1 3.
Prooperational Test Witnessina a.
ICP-PT-37-01. RT-1. " Auxiliary Feedwater System (Motor Driven Pumps)"
The purpose of this retest was to verify those items which remained open items to ICP-PT-37-01, Rev. O, and to retest certain items whien required retest due to rework. The items to be tested and reason for retest were:
(1) Auxiliary feedwater valves control logic due to major rework of control boards, analog racks, relay racks, and cable spread room wiring.
(2) Motor-driven auxiliary feedwater pumps 1-1 and 1-2 control logic due to major rework of control boards, analog racks, relay racks, and cable spread room wiring.
(J) Motor-driven auxiliary feedwater pumps 1-1 and 1-2 ' hydraulic performance due to redesign of test line, mini-flow orifice, and a ICP-PT-37-01, Rev. O, open item.
(4) Auxiliary feedwater system response time because system response time was not determined during performance of ICP-PT-37-01, Rev. O.
(5) Auxiliary feedwater pumps 1-1 and 1-2 endurance test due to redesign of test line orifice due to unsatisfactory operation of originai test line orifice.
The NRC inspector noted that during the section of the test to verify item (1) above, the timing of the feedwater valves was not within the range specified in the test. An error in the calibration of the timing logic was found to be the problem.
The logic was recalibrated and the test section was reperformed and the results were satisfactory.
During the 48-hour performance run of pump 1-1, a high temperature condition developed in the pump outboard bearing.
This resulted in having to stop the pump and check for bearing misalignment or other problems.
There were no apparent problems.
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. Several attempts were made to re-initiate the 48-hour run. These were unsuccessful until it was determined that there was too much oil in the outboard bearing. The level was adjusted and the 48-hour performance run of pump 1-1 was successfully completed.
b.
ICP-PT-49-01. Rev. 3. "Letdcwn. Charaina and Seal Water System" The purpose of the test was:
(1) To verify proper operation of control and interlock functions for various valves arid pumps in the Chemical and Volume Control System (CVCS) (see Section 2.0 for components tested).
(2) To verify response of various CVCS valves and pumps to Solid State Protection System (SSPS) signals such as safety injection, inclujing response times of valves.
(3) To verify hydraulic performance of the positive displacement charging pump.
(4) To verify proper operation of the volume control tank diversion valves.
The NRC inspector observed portions of the last phases of this test.
A review of the completed portion of the test was also performed, including a review of the test log entries, test procedure deviations and test deficiency reports, if any. The witnessing of this test was in addition to the preoperational tests preselected by the NRC inspectors for observation.
c.
ICP-PT-57-02, RT-1. " Centrifugal Charaina Pump Test" The purpose of this test was to verify proper operation of control i
and interlock functions for various valves in the CVCS which are related to the centrifugal charging pump high head injection flowpaths.
The retest was required as a result of electrical rework for train separation criteria and walkdown deficiencies.
The NRC inspector witnessed portions of the test performance from the control room and hot shutdown panel. There were no problems i
encountered with the test.
The NRC inspector, however, noted that when a transfer switch is operated to transfer control of a device from the control room to the i
hot shutdown panel the device being transferred will go to the position dictated by the control switch on the hot shutdown panel.
This will
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result in valves changing position unless the hot shutdown panel
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. valve control switches are matched to the actual valve positions prior to operating the transfer switches. This concern was discussed with the licensee, who indicated that procedural or hardware changes are under consideration. The NRC inspector will followup on this
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during subsequent inspections.
d.
3CP-PT-04-01. RT.1. " Station Service Water (SSW)"
The purpose of this test was to verify the operating characteristics and to demonstrate the capability of each train of the SSW to supply adequate flow to each of the components served.
As independent inspection effort, the NRC inspectors witnessed the performance of this preoperational test over a period of several days. There were no major problems associated with obtaining 1-satisfactory test results. The system performed as expected.
I However, the NRC inspector observed problems which resulted in two violations:
l (1) During the flow balancing of the SSW system in accordance with ICP-PT-04-01, it was necessary to place the SSW Chlorination System in operation in accordance with System Operating j
Procedure 50P-501A, " Station Service Water System."
Step 5.4.1.6 of SOP-501A directs the operator to open SSW Chlorination Valve XSW-036. The operator, in the presence of the System Test Engineer (STE), noticed that what appeared to be the correct valve was labeled "XSW-042."
Instead of halting the l
test to determine whether the valve label or the procedure was in error, the operator proceeded to open the valve. When the l
NRC inspector brought his attention to the procedure violation, j
the operation was aborted and the valve restored to the shut l
l position.
Subsequently, it was determined the procedure was in i
error; thus, it was changed accordingly and the operation
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resumed.
Prior to issuance of this inspection report, the Notice of Violation was transmitted to the licensee as Severity Level IV Violation 445/8421-01. This is the second violation issued in recent weeks pertaining to lack of procedure compliance. The previous violation was identified as 445/8418-01 and contains three examples of failure to follow procedures.
The licensee was made aware by the resident inspectors of the importance of decisive permanent corrective action by senior management to prevent future procedure violations as the pace of testing and operations increases at CPSES.
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(2) During the flow balancing of the SSW system, when the procedure required the STE to record the flow of service water to
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conta),iment spray cooling, the installed gage was pegged high with or without flow.
It became evident that the gage was malfunctioning due to air binding. There was no prerequisite in the 1CP-PT-04-01 to provide for filling and venting of the installed instruments used for this test, just prior to the test. Without such a prerequisite, the data is subject to question. because air in the instrument lines will cause j4 erroneous readings. This is contrary to Criterion XI of Appendix 8 to 10 CFR 50, and Notice of Violation was transmitted to the licensee prior to this inspection report in which the violation was identified as a Level IV Violation 445/8421-02.
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ICP-PT-29-02. kT-1. " Diesel Generator (DG) Control Circuit Functiona;
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and Start Test" l
l The purpose of this test was to functionally demonstrate electrical and pneumatic control circuit operability in the manual mode of l
operation for Train A diesel generator.
The NRC inspector witnessed parts of this test to verify that the testing was conducted in accordance with approved procedures, that the observations recorded by the STE were consistent with the observations of the NRC inspecter, that test results were adequately
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documented, and that the procedure is adequate to accomplish the intended purpose.
The test was conducted in a professional efficient manner. There were no problems observed by the NRC inspector with regard to the above
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attributes; however, as the NRC inspector observed the interlock testing associated with the DG barring device (the "Barring Device" is an air-operated jacking mechanism installed on the DG for the nurpose of rotating the crankshaft during maintenance), he noticed that service air was not connected.
Instead, a temporary air hose was connected from a portable diesel air compressor outsioe.
1CP-PT-29-02 did not have a prerequisite requiring service air (or a temporary source of air) to conduct the test. This left the STE to his own devices to perform the test and as such is contrary to Criterion XI of Appendix 8 to 10 CFR 50.
f.
In addition to the above tests that were completed during this reporting period, these three additional tests were started but are still in progress. These tests are:
ICP-PT-48-02, " Containment Spray System ICP-PT-02-02, "118 VAC RPS Inverters" 1CP-PT-34-01, Rev.1, " Main Steam Isolation Valves"
1
. No violations or deviations were found during witnessing of the above three operational tests.
4.
Plant Status The following is a status of TUEC (TUGCO) manning levels for operations and plant testing activities as of June 30, 1984.
a.
Operations Mannino Status l
Authorized Personnel Level (including maintenance, operations, administration, quality assurance, and engineering) - 553 Number Presently on Board - 482 b.
Plant Testina Status The present status of the NRC preoperational testing phase inspection program is approximatly 60 percent complete.
The licensee preoperational testing program is as follows:
Test Completion Status Preoperation Tests-136 Acceptance Tests-64 5.
Exit Interview An exit interview was conducted July 6, 1984, with licensee representatives (identified in paragraph 1).
During this interview, the SRRI and RRI reviewed the scope and discussed the inspection findings.
The licensee acknowledged the findings.
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y kh y-UNITED STATES j..., 5" ( I,I "g
.JUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20665 AUG 171984 Docket No.: 50-445 ATTACit1ENT D Mr. M. D. Spence s
I President Texas Utilities Generating Company 400 N. Olive Street L. 8. 81 Dallas, Texas 75201
Dear Mr. Spence:
Subject:
Acceptance of Preoperational Test Deferrals for Comanche Peak Steam Electric Station, Unit 1 The staff has completed its review of the following preoperational tests requested by letters dated May 29, June 5, June 8 and June 15, 1984 from B. R. Clearents:
1.
Containment Cooling Systems 2.
Safety Injection System Check Valve Leakage 3.
Turbine Drive Auxiliary Feedwater Pump Steam Supply Line Check Valve and Drain Pot Leve? Control Valve 4.
Reactor Coolant Pump Seal Performance 5.
Thermal Expansion Testing 6.
Control Room Ventilation System Enclosed are the staff's evaluations which are the proposed findings for inclusion in a future SER supplement. These proposed findings indicate that the requested deferrals are acceptable. Therefore, the Unit 1 Oper-ating License will contain license conditions consistent with your commit-ments on conducting the tests prior to initial criticality.
Sincerely, w
au B.' J.oungbloo Chief Li niingBranc No. 1 Divi,s' ion of Lic'nsing e
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Enclosure:
As stated cc: See next page I
i-
l COMANCHE PEAK AUG 171934 l
Mr. M. D. Spence President Texas Utilities Senerating Company 400 N. Olive St., L.B. 81 Dallas, Texas 75201 cc: Nicholas S. Reynolds, Esq.
Mr. James E. Cummins v
Bishop, Liberman, Cook, Resident Inspector /Ccmanche Peak Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W.
c/o U. S. Nuclear Regulatory Washington, D. C. 20036 Connission P. O. Box 38 Robert A. Wooldridge, Esq.
Glen Rose, Texas 76043 Worsham, Forsythe, Sampels &
Wooldridge Mr. John T. Collins 2001 Bryan Tower, Suite 2500 U. S. NRC, Region IV Dallas, Texas 75201 611 Ryan Plaza Drive Suite 1000 Mr. Homer C. Schmidt Arlington, Texas 76011 Manager - Nuclear Services Texas Utilities Generating Company Mr. Lanny Alan Sinkin Skyway Tower 114 W. 7th, Suite 220 400 North Olive Street Austin, Texas 78701 L. B. 81 Dallis, Texas 75201 B. R. Clements Vice President ~ Nuclear Mr. H. R. Rock Texas Utilities Generating Company Gibbs and Hill, Inc.
Skyway Tower 393 Seventh Avenue 400 North Olive Street New York, New York 10001 L. B. 81 Dallas, Texas 75201 Mr. A. T. Parker Westinghouse Electric Corporation William A. Burchette, Esq.
P. O. Box 355 1200 New Hampshire Avenue, N. W.
l Pittsburgh, Pennsylvania 15230 Suite 420 Washington, D. C.
20036 Renea Hicks, Esq.
Assistant Attorney General Ms. Billie Pirner Garde Environmental Protection Division Citizens Clinic Director P. O. Box 12548, Capitol Station Government Accountability Project Austin, Texas -78711 1901 Que Street, N. W.
Mrs. Juanita Ellis, President Citizens Association for Sound David R. Pigott, Esq.
Energy Orrick, Herringtor. & Sutcliffe 1426 South Polk 600 Montgomery Street Dallas, Texas 75224 San Francisco, California 94111 Ms. Nancy H. Williams Anthony Z. Roisman, Esq.
CYGNA Trial Lawyers for Public Justice 101 California Street 2000 P. Street, N. W.
San Francisco, California 94111 Suite 611 Washington, D. C. 20036
ENCLOSURE SUPPLE!1 ENTAL SAFETY EVALUATION REPORT DEFERRAL OF CERTAIN PREOPERATIONAL TESTS COMANCHE PEAK UNIT 1 Texas Utilities Generating Company in letters from B. R. Clements to H. R. Denton, NRC, dated May 29, June 5, June 8 and June 15, 1984', requested approval to defer six preoperatio.nal tests until after fuel loading. The testing would be completed prior to initial criticality with the exception of a portion of the thermal expansion test. This test requires heatup and return to cold shutdown conditions for completion and is setieduled at the completion of the 30 percent power plateau.
A.
ICP-PT-45-06, Containment Cooling Systems The applicant has requested that this test be repeated after fuel loading. Testing of the containment cooling systems were performed during"the normal preoperational test program; however, test deficiencies were identified requiring system modifications which could not be
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retested prior to the scheduled fuel loading, t
The repeat of this test after fue! loading is acceptable because only limited portions of the system require retesting, no technical specification exceptions are required and, for operation to continue, the system must still meet technical specifications temperature limits in critical areas.
B. '1CP-PT-57-09, Check Valve and Hot Functional Safety Infection The applicant has requested that this test be repeated after fuel loading.
During the initial test, a. number of check valves leaked in excess of their acceptance criteria. These valves have been repaired or replaced. The repeat testing of these valves would be ' performed as required-by the technical specifications surveillance tests for check valves.
It is acceptable to defer repeating portions of this test until after fuel loading, but before criticality, because (1) it is consistent with the technical specifications which control normal operation anc define check valve operability and (2) presents nc safety problem because retesting is completed prior to criticality.
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C.
ICP-PT-37-03, Turbine Driven Auxiliary Feedwater Pump Steam Supply Line Check Valve and Drain Pot Level Control Val ~ve In performing this preoperational test, a faulty level switch and corroded and bent disks in the steam supply line check valve were discovered. The applicant has made the necessary repairs and requests approval to complete the test after fuel loading (which will be the next scheduledheatup). This request is acceptable because there will have been no power operation prior to the retest and repeat of the necessary portions of this test will'be completed prior to criticality.
D.
ICP-PT-55-09, Reactor Coolant Pump (RCP) Test and ICP-PT-49-02, Seal Water and Letdown Flow Performance Several test deficiencies relating to the RCP seals were identified during the performance of these preoperational tests. Modifications to correct the deficiencies have been completed. The applicant proposed to incorporate the portions of the tests to be repeated into the appropriate startup test procedures to be performed after fuel loading, but prior to criticality. This schedule for retesting of the RCP seals is acceptable because it would be consistent with nonnal operating maintenance and test procedures and prior to initial criticality these systems are not required for plant safety.
E.
ICP-PT-55-11, Thermal Expansion Preoperational Test During the performance of the thermal expansion test, a number of test deficiencies were noted pertaining to snubbers, springs and supports.
These deficiencies were of three categcries:
(1) installed items did not meet acceptance criteria; (2) installed items removed due to interferences, and; (3) items not installed for the test.
The applicant will have corrected these deficiencies and proposes that the test be repeated after fuel loading when the next plant heatup is completed for initial criticality. Final cold setting of retest items would be acccmplished at the shutdown scheduled at the end of the 30%
power plateau.
The deferral of the thermal expansion retest is acceptable because it is I
consistent with approved industry practice on other plant test programs.
Furthermore, compliance with Technical Spacifications relating to piping supports will be required for plant operation to proceed.
F.
Control Room Ventilation System During performance of the Control Room Ventilation System preoperational test, it was detennined that the system provided more than adequate air supply to the control room area for Unit 1, but less than design air flow was supplied to Unit 2 control room area. The applicant is proceeding with modifications to the ventilation system to correct the design deficiency. The applicant plans to start retesting the modified system, but anticipates not being able to complete the testing prior to scheduled Unit 1 fuel loading. The applicant, therefore, requests deferral of completien of the test until after fuel loading.
Based on the condition that this deferral is a retest of a system which was already determined to be acceptable for the Unit I control area, we find the deferral of the retesting of the Control Room Ventilation System until completion of the initial fuel loading of Unit 1.(and before initial criticality) to be acceptable.
In sumary, the deferral of these six preoperational tests represent retesting of modifications made to correct identified system deficiencies in the respective systems.
Retesting these systems after initial fuel loading, but prior to initial criticality, will pose no safety problem will be s
controlled by the plant Technical Specifications and are consistent with
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other plant test programs. On this basis, the requested deferrals are approved.
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pf UNITED STATES 8
NUCLEAR REGULATORY COMMISSION a
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ATTACHMENT E Docket Nos. 50-445 AUG 28 1!B4 L
and 50-446
.c Mr. M. D. Spence l
President Texas Utilities Generating Company 400 N.' Olive St., L. B. 81 Dallas, Texas 75201
Dear Mr. Spence:
l
Subject:
Request for Exemption from a Portion of General Design Criterion 4 of Appendix A to 10 CFR Part 50 Regarding the Need to Analyze Large Primary Loop Pipe Ruptures as the Structure Design Basis for Comanche Peak Steam Electric Station (Units.1 and 2)
By letter dated October 31, 1983, Texas Utilities Generating Company (TUGCO) provided Westinghouse Report MT-SME-3135 (proprietary) as the technical basis in support of its request for exemption from a portion of the requirements of General Design Criterion (GDC) 4 of Appendix A to 10 CFR Part 50. The Westing-house report addressed the " leak-before-break" concept as an alternative to providing protective devices against the dynamic effects of postulated ruptures in the primary coolant loops. My letter to R. J. Gary dated March 2,1984, re-quested responses to questions and comments raised by the staff based on its review of Westinghouse Report MT-SME-3135 and its generic review of Westing-house Generic Report WCAP-10456 (proprietary), which provided an analysis of the fracture toughness of piping materials under thennal aging conditions.
l Your letter to H. R. Denton dated April 23, 1984, submitted a revision to West-l inghouse Report MT-SME-3135, identified as Westinghouse Report WCAP-10527 (pro-l prietary), which responded to the questions and cormnents furnished by my letter dated March 2, 1984. In a separate letter to H. R. Denton (also dated April 23, 1984), you provided a value-impact analysis, associated with your exemption request.
Since the Westinghouse Report WCAP-10527 provided supporting analyses encom-passing other structures in both Units 1 and 2, and seemed to be in conflict with the scope of the exemption requested in an earlier TUGCO. letter dated February 17,1984, TUGC0 was requested to clarify this apparent inconsistency.
H. C. Schmidt's letter to me, dated June 7,1984, provided clarification stat-ing that the exemption requested from the GDC 4 requirements was limited to the installation of jet impingement shields associated with postulated pipe breaks in eight (8) locations per loop in Unit 1, as speci-fied in Section 4.0 of the value-impact analysis submitted by TUGC0 letter to H. Denton dated April 23, 1984 s
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. O Mr. M. D. Spence AUG 2 81984 On the basis of the staff's evaluation of these.ubmittals, the Commission has granted your exemption request for Comanche Peak, Unit 1, which 'is enclosed.
The exemption pertains only to the installation of jet impingement shields as reflected in Mr. H. C. Schmidt's letter to me dated June 7,1984. The enclosed exemption is.being forwarded to the Office of tne Federal Register for publi-cation. We are also processing your request for amendment of the construction permit for Unit 1 to reflect this exemption.
Sincerely, 1
7
(
. Yo ngblood Chief Lice i g Branch No.1
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Div i of Licensing
Enclosure:
As stated cc: See next page l
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l I
l COMANCHE PEAK Mr. M. D. Spence President Texas Utilities Generating Company 400 N. Olive St., L.B. 81 l
Dallas, Texas 75201
- cc: Nicholas S. Reynolds, Esq.
Mr. James E. Cunnins Bishop, Liberman, Cook, Resident Inspector / Comanche Peak Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W.
c/o U. S. Nuclear Regulatory Washington, D. C. 20036 Commission P. O. Box 38 l
Robert A. Wooldridge, Esq.
Glen Rose Texas 76043 Worsham, Forsythe, Sampels &
Wooldridge Mr. John T. Collins 2001 Bryan Tower, Suite 2500 U. S. NRC, Region IV Dallas, Texas 75201 611 Ryan Plaza Drive Suite 1000 Mr.- Homer C. Schmidt Arlington, Texas 76011 Manager - Nuclear Services Texas Utilities Generating Company Mr. Lanny Alan Sinkin Skyway Tower 114 W. 7th, Suite 220 400 North Olive Street Austin, Texas 78701 L. B. 81 J
i Dallas, Texas 75201 B. R. Clements Vice President Nuclear Mr. H. R. Rock Texas Utilities Generating Company Gibbs and Hill, Inc.
Skyway Tower 393 Seventh Avenue 400 North Olive Street New York, New York 10001 L. B. 81 Dallas, Texas 75201 Mr. A. T. Parker Westinghouse Electric Corporation William A. Burchette, Esq.
P. O. Box 355 1200 New Hampshire Avenue, N. W.
Pittsburgh, Pennsylvania 15230 Suite 420 Washington, D. C.
20036 Renea Hicks, Esq.
Assistant Attorney General Ms. Billie Pirner Garde Environmental Protection Division Citizens Clinic Director P. O. Box 12548, Capitol Station Government Accountability Project Austin, Texas 78711 1901 Que Street, N. W.
Mrs. Juanita Ellis, President Citizens Association for Sound David R. Pigott, Esq.
Energy Orrick, Herrington & Sutcliffe 1426 South Polk 600 Montgomery Street Dallas, Texas 75224 San Francisco, California 94111
~
Ms. Nancy H. Williams Anthony Z. Roisman, Esq.
CYGNA Trial Lawyers for Public Justice 101 California Street 2000 P. Street, N. W.
San Francisco, California 94111 Suite 611 Washington, D. C. 20036 s.
7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION t
In the Matter of L,
TEXAS UTILITIES GENERATING COMPANY Docket Nos. 50-445 and 50-446 (Comanche Peak Steam Electric Station, Units 1 and 2)
EXEMPTION
.I :.
On July 20, 1973, theTexasUtilitiesGeneratingCompany(theapplicant)
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tendered an application for licenses to construct Comanche Peak Steam Electric Station, Units 1 and 2 (Comanche Peak or the faciFity) with the Atomic Energy Commission (currently the Nuclear Regulatory Commission or the Commission).
Following a public hearing before the wLtomic Safety and Licensing Board, the Commission issued Construction Permit Nos. CPPR-126 and CPPR-127 permitting the construction of Units 1 and 2, respectively, on December 19, 1974. Each Unit.of the facility is a pressurized water reactor, combining a Westinghouse Electric Company nuc. lear steam supply system, located at the applicant's site in Somervell/ Hood Counties, Texas, approximately 40 miles southwest of Fort j
Worth, Texas.
On February 27, 1978, the applicant tendered an application for Operating Licenses for each Unit of the facility, currently in the licensing review process, with Unit I licensing to occur in the near term.
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7590-01 II.
i The Construction Permits issued for constructing the facility provide, in pertinent part, that the facility Units are subject to all rules, regulations and Orders of the Comission. This includes General Design Criterion (GDC) 4 of Appendix A to 10 CFR 50. GDC 4 requires that structures, systems and components important to safety shall be designed to accommodate the effects of and to be compatible with the environment'al conditions associated with the nomal operation, maintenance, testing and postulated accidents, including loss-of-coolant accidents. These structures, systems and components shall be appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, discharging fluids that may result from equipment failures, and from. events and conditions outside the nuclear power unit.
1 i
By a submittal dated October 31, 1983, the applicant requested an exemp-tion from a portion of the requirements of GDC 4 to: (1) eliminate the need to postulate circumferential and longitudinal p'ipe breaks in the Reactor Coolant System (RCS) primary loop (hot leg, cold leg and cross-over leg piping);
(2) eliminate the need to install pipe whip restraints and jet impingement shields associated with previously postulated breaks in the RCS primary loops and; (3) to elimina'te 'the need to consider dynamic effects and loading condi-tions associated with previously postulated pipe breaks in the RCS primary loop, including jet impingement loads, cavity pressure loads, blowdown loads in the RCS and attached piping, and subcompartment pressure loads.
In support of this exemption request, the applicant's submittal enclosed Westinghouse Re-port MT-SME-3135 (Reference 1) containing the technical basis for their request.
i 7590-01 Based on its review of the applicant's submittal, the NRC staff requested
. I additional information and provided comments on the reports (References 1 and 9) which were transmitted to the applicant in tbe form.of questions by NRC letter
~ :.
dated March 2,1984, (Reference 2).
By a submittal dated April 23, 1984, the applicant responded to the staff's questions (Reference 2) and provided a revision to the Reference 1 report iden-tifiedasWestinghouseReportUCAP-10527(Reference 3).
In a separate submittal, also dated April 23, 1984, the applicant provided a value-impact analysis which, together with the technical information contained in the Reference 3 report, provided a comprehensive justification for requesting a partial exemption from the requirements of GDC 4.
From the deterministic fracture mechanics analysis contained in the technical information furnished, the applicant s-tated that the postulated double-ended f
guillotine breaks (DEGB) of the primary loop coolant piping will not occur in Comar.:he Peak Units 1 and 2.and, therefore, need not be considered as a design basis for installing protective structures, such as pipe whip restraints and
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jet impingement shields, to guard against the dynamic effects associated with such postulated breaks.
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7590-01 By letter dated June 7,1984 (Reference 10), the applicant clarified the scope of its request for exemption from GDC 4 requirements. Since the Westinghouse
. Report WCAP-10527 provided analyses encompasping other structures in both Comanche Peak Units 1 and 2, and seemed to be in conflict with the scope of the exemption requested in an earlier letter dated February 17,1984 (Refer-ence 11), the applicant stated in the Reference 10 letter that, although the analyses contained in the Report WCAP-10527 encompassed relief from the need to install pipe break protective devices 'in both Units 1 and 2, the exemption being requested pertained solely to the installation of jet impingement shields associated with such breaks in eight (8) ' locations per loop in Comanche Peak Unit 1, as~speci.fied in Section 4.0 of the value-impact analysis submitted by the applicant's letter dated April 23, 1984.
III.
The Commission's regulations require that appifcants provide protective measures against the dynamic effects of postulated pipe breaks in high energy fluid system piping. Protective meas 0'res include physical isolation from postulated pipe rupture locations if feasible or the installation of pipe whip restraints, jet impingement shields or compartments.-
In 1975, concerns arose l
as to the asymmetric l'oads on pressurized water reactor (PWR) Nessels and their internals which could result from these large postulated breaks at discrete locations in the main primary coolant loop piping. This led to the establishment of Unresolved Saf'ty Issue (USI) A-2, " Asymmetric Blowdown Loads e
on PWR Primary Systems."
~~
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7590-01 The NRC staff, after several review meetings with the Advisory Committee.
on Reactor Safeguards (ACRS) and a meeting with the NRC Committee to Re-view Generic Requirements (CRGR), concluded that for certain facilities an exemption from the regulations would be acceptable as an alternative for reso-lution of USI A-2 for sixteen facilities owned by eleven licensees in the West-inghouse Owner's Group (one of these facilities, Fort Calhoun has a Combustion Engineering nuclear steain supply system). This NRC staff position was stated in Generic Letter 84-04, published on February 1,1984 (Reference 4). The generic letter states that the affected licensees must justify an exemption to GDC 4 on a plant-specific basis. Other PWR applicants or licensees may request similar exemptions from the requirements of GDC 4 provided that they submit an acceptable technical basis for eliminating the need to postulate pipe breaks.
~
The acceptance of an exemption was made possible by the development of advanced fracture mechanics technologyr These advanced fracture mechanics techniques deal with relatively small flaws in piping components (either postulated or real) and examine their behavior under various pipe loads. The objective is to demonstrate by deteniiinistic analyses that the detection of small flaws by either inseIvice inspection or leakage monitoring systems is assured long before the flaws can grow to critical de unstable sizes which could lead to large break areas such as the DEGB or its equiva' lent. The concept underlying such analyses is referred to as " leak-before-break" (LBB).
T There is no implication that piping failures cannot occur, but rather that
, improved knowledge of the failure modes of piping systems and the application of appropriate remedial measures, if indicated, can reduce the probability of catastrophic failure to insignificant values.
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7590-01 Advanced fracture mechanics technology was applied in topical reports (References 5, 6 and 7) submitted to.the staff by Westinghouse on behalf of the licensees belonging to the USI A-2 Owners Group, Although the topical reports were intended to resolve the issue of asymmetric blowdown loads that resulted from a limited number of discrete break locations, the technology advanced in these topical reports demonstrated that the probability of breaks occurring in the primary coolant system main loop piping is sufficiently low such that these breaks need not be co'nsidered as a design basis for l
requiring installation of pipe whip restraints or jet impingement shields.
The staff's. Topical Report Evaluation is' attached as Enclosure 1 to Reference 4.
1 l
1 Probabilistic. fracture mechanics studies conducted by the Lawrence Liver-1 more National Laboratories (LLNL) on both Westinghouse and Combustion Engineer-ing nuclear steam supply system main loop piping (Reference 8) confirm that both the probability of leakage (e.g., undetected flaw growth through the pipe wall by fatigue) and the probability of a DEGB are very low. The results given in Reference 8 are that the best-estimate leak probabilities for Westinghouse nuclear steam supply system main loop piping range from 1.2 x 10-8 to 1.5 x 10-7 per plant year and the best-estimate DEGB probabilities range from 1 x 10-12 to 7 x 10-12 per plant' ye'ar. Similarly, the best-estimate leak probabilities for Combustion Engineering nuclear steam supply system main loop piping range from 1 x 10-8 per plant year to 3 x 10-8 per plant year, and the best-estimate DEGB probabilities range from 5 x 10-14 to 5 x 10-13 per plant year. These results do not affect core melt probabilities in any significant way.
7590-01 i
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During the past few years it has also become apparent that the requirement for installation of large
- 1ve pipe whip restraints and jet impingement shields is not necessarily the most cost effective way to achieve the desired i-level of safety, as indicated in Enclosure 2, Regulatory Analysis, to Reference 4.
'E'ven for new plants, these devices tend to restrict access for future inservice inspection of piping; or if they are removed and reinstalled for inspection, 1
there is a potential risk of damaging the piping and other safety-related components in this process. If installed in op'erating plants, high occupational radiation exposure (ORE) would be incurred while public risk reduction would be very-low. Removal and reinstallation for inservice inspection also entail significant ORE over the life of a plant.
l IV.
i The primary coolant system of Comanche Peak Units 1 and 2, described in l
Reference 3, has four main loops each comprising a 33.9 inch diameter hot leg, a 36.2 inch diameter crossover leg and 32.14 inch diameter cold leg piping.
l The material in the primary loop piping is cast stainless steel (SA 351 CF8A).
i In its review of Reference 3, the staff evaluated the Westinghouse analyses with regard to:
the location of maximum stresses in the piping, associated with the combined loads from normal operation and the SSE; potential cracking mechanisms; size of through-wall cracks that would leak a detectable amount under normal loads and pressure; stability of a " leakage-size crack" under normal plus SSE loads and _ the expected margin in terms of ' load;
7590-01 margin based on crack size; and the fracture toughness properties of thermally-aged cast stainless steel piping and weld material.
The NRC staff's criteria for evaluation'of the above parameters are delin~-
eated in its Topical Report Evaluation, Enclosure 1 to Reference 4, Section 4.1, "NRC Evaluation Criteria", and are as follows:
4 (1) The loading conditions should includ.e the -static forces and moments (pres-sure, deadweight and thermal expansion) due to normal operation, and the forces and moments associated with the safe shutdown earthquake (SSE).
These forces and moments should be located where the highest stresses, coincident 'with the poorest material properties, are induced for base matiir~ials, weldments and safe-ends.
(2) For the piping run/ systems under evaluation, all pertinent i*rmation which demonstrates that degradation or failure of the piping resulting from stress corrosion cracking, fatigue. or water hammer is not likely, should be provided. Relevant operating history should be cited, which includes system operational procedures; system or component modification; water chemistry parameters, limits and controls; resistance of :aaterial to various forms of stress corrosion, 'and performance under cyclic loadings.
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(3) A through-wall crack should be postulated at the highest stressed locations i
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determinedfrom'(1)above. The size of the crack should be large enough -
so that the leakage is assured of detection with' adequate margin using the minimum installed leak detection capability when the pipe is subjected to normal operational loads.
(4) 'It should be demonstrated that the postulated leakage crack is stable under normal plus SSE loads for long periods of time; that is, crack growth, if any, is minimal during an earthquake. The margin, in terms of applied loads, should be determined by a crack stability analysis, i.e., that the leakage-size crack will not experience unstable crack growth even if larger loads (larger than design loads) are applied. This analysis should demon-strate that crack growth is stable and the final crack size is limited, such that a double-ended pipe break will not occur.
(5) The crack size should be determined by comparing leakage-size crack to e <
critical-size cracks. Under normal plus SSE loads, it should be demon-s f
.strated that there is adequate mi~rgin between the leakage-size crack and the critical-size crack to account for the uncertainties inherent in the analyses, and leakage detection capability.
A' limit-load analysis may suffice for this purpose, however, an elastic-plastic fra'cture mechanics (tearing instability) analysis is preferrable.
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(6) The materials data provided should include types of materials and materials specifications used for base metal, weldments and safe-ends, the materials properties including the J-R curve used.in the. analyses, and long-tenn
" L, effects such as thennal aging and other limitations to valid data (e.g.
J maximum,
.ximumcrackgrowth).
Based on its evaluation of the analysis contained in Westinghouse Report WCAP-10527 (Reference 3), the staff finds' that the applicant has presented an acceptable technical jtstification, addressing the above critaria, for not
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installing, protective devices to deal with the dynamic effects of large pipe ruptures in the main loop primary coolant system piping of Comanche Peak, Units 1 and 2.
This finding is predicated on the fact that each of the para-meters evaluated for Comanche Peak is. enveloped by the generic analysis per-fonned by Westinghouse in Reference (5), and accepted by the staff in Enclo-sure I to Reference 4.
Specifically:
(1) The loads associated with the highest stressed location in the main loop.
primary system piping are considerably lower than the bounding loads used by Westinghous'e t'n Reference 5, or those established by the staff as limits (e.g. a moment of 42,000 in-kips in Enclosure 1 to Reference 4).
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7590-01 11-(2) For Westinghouse plants, there is no history of cracking failure in reac-tor primary coolant system loop piping. The Westinghouse reactor coolant system primary loop has an operating history which demonstrates its inherent stability. This includes a low susceptibility to cracking failure from the effects of corrosion (e.g. intergrannular stress corrosion cracking), water hammer, or fatigue (low and high cycle). This operating history totals over
'400 reactor-years, including five plants each having 15 years of operation and 15 other plants with over 10 years of operation.
(3) The results of the leak rate calcula'tions performed for Comanche Peak.
using an initial through-wall crack are identical to those of Enclosure 1 toReference'(4). The Comanche Peak plant has an RCS pressure boundary leakdetectionsystemwhichisconsistentwiththeguidelinesofRegula-tory Guide 1.45, and it can detect leakage of one (1) gpm in one hour.
i The calculated leak rate through-the postulated flaw is large relative to the sensitivity of the Comanche Peak plant leak detection system.
l l
(4).The expected margin in terms of Itad for the leakage-size crack under 1ormal plus SSE loads is within the bounds calculated by the staff in Section4.2.3ofEnclosure(1)toReference4.-Inaddition,thestaff found a significant margin in terms of loads larger than normal plus SSE loads.
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(5) The margin between the leakage-size crack and the critical-size crack was calculated. Again, the results demonstrated that a significant margin exists and is within the bounds of Section 4.2,3 of Enclosure 1 to Refer-ence 4.
(6). As an integral.part of its review, the staff's evaluation of the material properties data of Reference 9 is enclosed as Appendix 1 to this Exemption.
In Reference 9, data for ten (10) p1' ants,' including the Comanche Peak Units,
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are presented, and lower bound or " worst case" materials properties were
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identified an'd used in the analysis performed in the Reference 3 report 2
by Westinghouse. The staff's upper bound of 3000 in-lb/in on the applied J (r.efer to Appendix 1, page 6) was not exceeded; the applied J for Comanche Peak in Reference 3 was substantially less than 3000 in-lb/in,
2 1
In view of the analytical results--presented in the Westinghouse Report for Comanche Peak (Reference 3) and the staff's evaluation findings related above, the staff concludes that the probability or likelihood of large pipe breaks occuring in the primary coolant syst& loop of Comanche Peak Units 1 and 2 is j
sufficiently low such that such pipe breaks need not be considered as a design.
basis for requiring protective devices. However, the pipe whip restraints have already been inst'alled in Unit 1, and the applicant has 1-imited the scope of its exemption request to the installation of jet impingement shields in Unit 1 only. The requested exemption from GDC 4 is limited to exemption from the i
need to install jet impingement' shields at specified locations in Unit 1.
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F 7590-01 The staff also reviewed the value-impact analysis provided by the appli-cant for not providing protective structures against postulated reactor coolant system loop pipe breaks to assure as low as. reasonably achievable (ALARA) expo-l sure to plant personnel. Consideration was given to design features for reduc-i-
ing doses to personnel who must operate, service and maintain the Comanche Peak instrumentation, controls, equipment, etc. Normally, facilities and equipment are designed to save person-rems; however, the Comanche Peak value-impact anal-ysis shows that the addition of protective devices for RCS pipe breaks will cost about 2 person-rems annually due to the slowing down of nonnally anticipated I
j work, and increasing the scope of routine maintenance in radiation areas that i
j would be involved. The analysis provides a reasonable estimate for this addi-i l
tional radiological cost.
In view of the very low probability of pipe breaks at l
the specified locations covered by this exemption,- the reduction of occupational i
exposure resulting from this exemption outweighs the potential accident exposure reduction that might result from installation of the jet impingement barriers.
VI.
In view of the staff's evaluation findings, conclusions, and recommenda-
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tions above, the Commission has detennined that, pursuant to 10 CFR 50.12(a),
this Exemption is authorized by law and will not endanger life' or prop-erty or the common defense I
's otherwise in the public inter-est. The Commission he' ited exemption from GDC 4 o' nsee not to install 1
q 7590-01 jet impingement shields associated with postulated pipe breaks of the eight (8) locations per loop in the Comanche Peak Unit 1 primary coolant system, as specified in Section 4.0 of the value-impact: analysis submit by the applicant's letter dated April 23, 1984. This Exemption does not pertain to the installation of pipe whip restraints, already installed in Unit 1, or to the installation of pipe whip restraints and jet impingement shields in Comanche Peak Unit 2.
The portion of the request concerning Unit 2 will be dealt with in a separate NRC action.
The Commission has determined that the issuance of the exemption will have no significant environmental impact on the environment (49 FR 33945
).
FOR THE NUCLEAR REGULATORY COMMISSION Odb Frank Miraglia$.Dh:y Director Division of Licensing Office of Nuclear Reactor Regulation Datedatpthesda, Maryland this Jayof 4
1984 G
7590-01 REFERENCES (1) Westir.ghouse Report MT-SME-3135, " Technical Bases for Eliminating Large Primary Loop Pipe Ruptures as the Structural Design Basis for Comanche Peak Units 1 and 2," October 1983, Westinghouse Class 2 proprietary.
(2) Letter to R. J.' Gary of Texas Utilities' Generating Company, " Request for Additional Information Concerning Leak-Before-Break Analysis for Comanche Peak Steam Electric Station (Units 1.and 2)," dated March 2, 1984.
(3) Westinghouse Report WCAP-10527, " Technical Bases for Eliminating Large Primary Loop Pipe Rupture as the Structural Design Basis for Comanche Peak Units 1 and 2," April 1984, Westinghouse Class 2 proprietary.
(4) NRC Generic Letter 84-04, " Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Breaks in PWR Primary Main Loops," February 1, 1984.
(5) Mechanistic Fracture Evaluation of Reactor Coolant Pipe Containing a Postulated Circumferential Throughwall Crack, WCAP-9558, Rev. 2, May 1981, Westinghouse Class 2 proprietary ~.
._ (6) Tensile and Toughness Properties of Primary Piping Weld Metal for Use in Mechanistic Fracture Evaluation, WCAP-9787, May 1981, Westinghouse Class 2 proprietary.
(7) Westinghouse Response to Questions and Coments Raised by Members of ACRS Subcomittee on Metal Components During the Westinghouse Presentation on September 25, 1981, Letter Report NS-EPR-2519, E. P. Rahe to Darrell G.
Eisenhut, November 10,1981, West 4nghouse Class 2 proprietary.
(8) Lawrence Livennore National Laboratory Rpeort, UCRL-86249, " Failure Prob-ability of PWR Reactor Coolant Loop Piping," by T. Lo, H. H. Woo, G. S.
Holman and C. K. Chou, February 1984 (Preprint of a paper intended for publication).
(9) ~Westinghouse Report WCAP+10456, "The Effects of Thermal Aging on the Structural Integrity of Cast Stainless Steel Piping for Westinghouse Nuclear Steam Supply Systems," November 1983, Westinghouse Class 2 proprietary.
(10) Texas Utilities Generating Company letter TXX-4197, "Requ'est for Partial Exemption" (H. C. Schmidt to B. J. Youngblood) dated June 7,1984.
(11) Texas Utilities Generating Company letter TXX-4118, " Request for Partial Exemption," (R. J. Gary to B. J. Youngblood) dated February 17, 1984.
Notes: See next page 1
7590-01 REFERENCES NOTE. Non-proprietary versions of References 1, 3, 5, 6, 7 and 9 are available in the NRC Public Document Room as follows:
MT-SME-3136 WCAP 10528 r"
WCAP 9570 (6
WCAP 9788 (7
Non-proprietary version attached to the Letter Report (9
WCAP 10457 1
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APPENDIX 1 Evaluation of Westinghouse Report WCAP 10456, "The Effects of Thermal Aging j
on the Structural Integrity of Cast Sta$n1_ess l
Steel Piping for Westinghouse Nuclear Steam t
I Supply Systems" j
q INTRODUCTION i
The primary coolant piping in some Westinghouse Nuclear Steam Supply i
Systems (NSSS) contain cast stainless steel base metal and weld metal.
The base metal.and weld metal are fabricated to produce a duplex structure of delta (6) ferrite in an austenitic matrix. The duplex structure pro-l
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duces a material that has a higher yield strength, improved weldability and greater resistance to intergranular stres's-corrosion cracking-than a single phase austenitic material. However, as.early as 1965 (Ref.1),
it was recognized that loag time t.hermal aging at primary loop water I
temperatures (55d*F650'F)couldsihnificantlyaffecttheCharpyimpact toughness of the duplex structured alloys.
Since the Charpy impact test is a measure of a material's resistance to fracture, a loss in Charpy impact toughness could result in reduced structural stability in the piping system.
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The purpose of Report WCAP 10456 is to evaluate whether cast stainless steelbasemetalandweldmetalcontainingpostuiatedcrackswillbe sensitive to unstable fracture during the 40 year life of a nuclear power plant.
In order to determine whether a piping system will behave I
e 1
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'in such a fashion, the pipe materials' mechanical properties, design j-criteria and method of predicting failure must be established.
In this i
evaluat,fon, we assess the mechanical properties of thermally aged cast stainless steel pipe materials, which are reported in Report WCAP 10456.
i DISCUSSION
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Weld Metal
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Report WCAP 10456 refers to test results reported in a paper by Slama,
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st.al. (Ref. 2) to conclude that the weld metal in primary loop piping -
would not-be ' overly sensitive to aging and that the aged cast pipe base l
metal material would be structurally limiting.
In the Slama report eight (8) welds were evaluated. The tensile properties were only slightly affected by aging. The Charpy U-notch impact energy in the most highly sensitive weld decreased from 7daJ/cm2 (40 ft-lbs) to near 4daJ/cm2 (24 ft-lbs) after aging for 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> at 400*C (752*F).
l This change was not considered significant.
The relatively small effect of aging on the weld, as compared to cast pipe material was reported to be caused by a difference in microstructure and lower levels of ferrite in the weld than in the cast pipe material.
! l l
2.
Cast Stainless Steel Pipe Base Metal Report WCAP 10456 contains machanical property test results froe a number of heats of aged cast stainless steel material and a metallurgical study, which was performed by Westinghouse, to support a statistically based model for predicting the effect of thermal aging on the Charpy impact test properties of cast stain-less steel.: As a result of these tests and the proposed model, Westinghouse concludes that the fracture toughness test results from one heat of material tested represents end-of-life conditions
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for the ten (10) plants surveyed. The ten (10) plants surveyed
~~
are ider.tified as Plants A through J.
a.
Mechanical Property Test Results Reported in WCAP 10456 r
Mechanical property test results on aged and unaged cast stain-less steel materials, as reported in papers by Landerman and Bamford (Ref. 3), Bamford, Landerman and Diaz (Ref. 4), Slama et al.
(Ref. 2), were discussed in Report 10456.
In addition, Westinghouse performed confirmatory Charpy Y notch and J-integral tests on aged cast stainless steel material, which was tested and evaluated by Slama et al.
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4 The results of these tests indicate that:
, (1) The faLigue crack growth rates of aged or unaged material in air and pressurized water reactor environments were equivalent.
(2) Tensile properties were essentially unaffected except for a slight increase in tensile strength and a decrease in ductility.
l 1
l l
(3) J-integral test results indicate that the J and tearing.
1C modulus, T, are affected by aging.
b.
Mechanism Study in WCAP 10456 I
The tests and literature survey conducted by Westinghouse indicate that the proposed mechanism of aging occurs in the range of operating temperatures for pressurized water reactors and the data from accelerated aging studies can tie used to l
predict the behavior at operating temperatures.
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c.
Cast Stainless Steel Pipe Test The materials data discussed in the previous section of this evaluation were obtained from small specimens. As a consequence, the J-R results are limited to relatively short crack extensions.
To investigate the behavior of cast stainless steel in actual piping geometry, Westinghouse performed two experiments, cne of which was with thermally aged cast stainless steel and the other test was identical except that the steel was not thermally 1
aged.
Each pipe tested contained a throughwal circumferential crack to the extent specified in WCAP 104b6. The pipe sections were closed at the ends, pressurized to nominal PWR operating pressure and then bending loads were applied.
The results of the tests were very similar, in that both pipes displayed extensive ductility, and stable crack extension. There was no observed unstable crack extension or fast fracture.
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. The results of the Westinghouse pipe experiments indicate that cast stainless steel, both aged and unaged, can withstand crack extensions well beyond the range of the J-R results with small specimens. However, if crack extension is predicted in an actual applicstion of thermally aged cast stainless steel in a piping system, we believe that it is prudent to limit 2
the applied J to 3000 in-lbs/in or less ualess further studies and/or experiments demonstrate that higher values are tolerable.
Loss of initial toughness due to thermal aging of cast stainless steels at noimal nuclear facility operating temperatures occurs slowly over the course of many years; therefore, continuing study of the aging phencmenon may lead to a relaxation of this position.
Conversely, in the unlikely event that the total loss of toughness and the rate of toughness are greater than those projected in this evaluation, the staff will take appropriate action to limit the values to that which can be justified by experimental data.
Because the aging is a slow process, the staff believes there would be sufficient time for the staff to recognize the problem and to rectify the situation. However, the staff believes this situation is highly unlikely because the staff has accepted only the lower bounds of data that were gathered among ten plants l
encompassing the range of materials in use.
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7-d.
Effects of Thermal Aging on Westinghouse Supplied Centrifuga11y Cast _ Reactor Coplant Piping Reported in WCAP 10456 The reactor coalcnt cast stainless steel piping materials in the plants identified in WCAP 10455 as A through J, were produced to the specification SA-351. Class CF8A as outlined in ASME Code Section II, Part A and also to Westinghouse Equipment Specification G-678864,' as revised.
For these materials, Westinghouse has calculated the predicted end-of-life Charpy U-notch properties, based on their proposed model. The two (2) standard deviation end of-life lower limit value for all the plants surveyed was greater than the Charpy V notch properties of the aged reference materials, which Westinghouse indicates represents end-of-life properties for all the plants. As a result, Westinghouse con-cluded that the amount of embrittlement in the aged reference material exceed the amount projected at end-of-life for all cast stainless steel pipe materials in Plants A through J.
Conclusions Based on our review of the information and data contained in Westinghouse Report WCAP 10456, we conclude that:
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Weld metal that is used in cast stainless steel piping system is initially less fracture resistant than the cast stainless steel base metal. However, the weld metal is less susceptible to thermal aging than the cast stainless steel base metal. Hence, at end-of-life the cast stainless steel base metal is anticipated to be the least fracture resistant material.
2.
The Westinghouse proposed model may be used to predict the relative amount of embrittlement on a heat of cast stainless steel material.
The two standard deviation lower confidence limit for this model will provide a useful engineering estimate of the predicted end-of-life Charpy impact properties for cast stainless steel base metal.
r 3.
Since there is considerable scatter in J-integral test data for the heats of material tested, lower bound values for J aad T 1c should be used as engineering estimates for the fracture resistance of the aged reference caterial. We believe these 'talues should also provide a lower bound for the fracture resistance of aged and unaged weld metal.
If crack extension is predicted in an actual application l
of cast stainless steel in a piping system, we conclude that the applied J should be limited to 3000 in-lbs/in2 or less unless further
[.
studies and tests demonstrate that higher values are tolerable. The Westinghouse pipe tests demonstrate that this may be possible.
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Since the predicted end-of-life Charpy impact values for the materials in Plants A through J are greater than the value measured for the aged-reference material, the lower bound fracture properties for eged reference material may be used to determine the fracture resistance for the cast stainless steel material in Plants A through J.
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REFERENCES (1)
F. H. Beck, E. A. Schoefer, J. W. Flowers, M. E. Fontana, "New Cast High Strength Alloy Grades by Structural Control," ASTM STP 369 (1965)
(2)
G. Slama, P. Petrequin, S. H. Masson, T. R. Mager, "Effect of Aging on Mechanical Properties of Austenitic Stainless Steel Casting and Welds," presented at SMIRT 7 Post Conference Seminar 6 - Assuring Structural Integrity of Steel Reactor Pressure Boundary Components, l
August 29/30, 1983, Monterey, Ca.
(3)
E. I. Landerman and W. H. Bamford, " Fracture Toughness and Fatigue Characteristics of. Centrifuga11y Cast Type 316 Stainless Steel After Simulated Thermal Service Conditions.
Presented at the Winter Annual Meeting of the ASME, San Francisco, Ca.
December 197P (MPC-8 ASME) l l
(4)
W. H. Bamford, E. I. Landerman and E. Diaz, " Thermal Aging of Cast Stainless Steel and Its Impact on Piping Integrity." Presented at ASME Pressure Vessel and Piping Conference, Portland, Oregon, June 1983.' To be published in ASME Trans.
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e-ATTACHMENT F 35058 Federal Regisler / Vol. 49. No.173 / Wednesday. September 5.1984 / Notices Alternative Use r>f Resources Nuclear Reactor Regulation for This action does not involve the use of treatment pursuant to 10 CFR 2.206 of The Construction Permits issued for resources not previously considered in the Commission's regulations and a fina.,
connection with the Fmal Er.vironmental Director s Decision has been issued by constructing the facility provide. in Statement Relating to Operation of the Director denying the Petitioner a pertinent part. that the facility Units are Millstone Unit 2.
request. The reasons for this dental are subject to all rules. regulations and Orders of the Commission.This includes explained in the Director s Decision Agencies andPersons Consulted under to CFR 2.206"(DD-84-20), which General Design Criterion (CDC) 4 of The NRC staff reviewed the licensee's is available for inspecti3n in the App ndix to 10 50 C 4
request and did not consult other Commission a Public Do ument Room.
9 agencies or persons' 1717 H Street. NW., Washington, D.C.
components important to safety shall be 20555 and at the Local Public Document designed to accommodate the effects of Finding of No Significant Impact Room at the Robert E. Kennedy 1.ibrary.
and to be compatible with the
'Uvironmental conditions associated The Commission has determined not California Polytechnic State University.
to prepare an environmental statement San Luis Obispo. Cahfornia 93407.
with the normal ope ~ ration. maintenance.
for the proposed relief.
A copy of the decision will be filed testing and postulated accidents.
Based upon the foregoing with the Secretary for Commission including loss-of.ccolant accidents.
environmental assessment. we conclude review in accordance with to CFR These structures. systems and that the proposed action will not have a 2.20G(c). As provided in 10 CFR 2.206;c).
components shall be appropriately significant effect on the quality of the the decision will become the final action protected against dynamic effects.
human environment.
of the Commission 25 days after including the effects of missiles pipe For further details with respect to this issuance. unless the Commissicn. on its whipping. discharging fluids that may action. see the application for relief own motion, takes review of the result from equipment failures, and from dated May 4.1984. which is available for decision within that time.
r.ents and condi' ions outside the nuclear power unit.
public inspection al the Commission's Dated at Bethesda. Maryland. this 20th day Public Document Room.1717 H Street.
of August 1984.
By a submittal dated October 31.1983.
NW Washington. D C. and at the For the Nuclear Regulatory Commission.
the apphcant requested an exemption Waterford Public Library. Waterford.
from a portion of the requirements of Harold R.Denton, Director. Office ofNucl ear Reactor postulate circumferential and Dated at Bethesda. Maryland this :sth day g
uf,"
of August 1984.
longitudinal pipe breaks in the Reactor I"'
" " "I For the Nuclear Regulatory Coramission.
Coolant Svetem (RCS) pnmary loop (hot
^*'
t Gus C. t.ainas.
leg. cold leg and cross-over leg piping):
D: ision of Lscensing Officr ofNuclear l2) eliminate the need to install pipe whip restramts and jet impingement Beactor Regulation IDocket Nos. 50-445 and 50-4461 shields associated with previously irn um s me r,wd noe eu..
su.mo cooe isso
.as Texas Utilities Generating Co.
postulated breaks in the RCS primary (Comanche Peak Steam Electric loops and. (3) to ehm:nate the need to Station, Unita 1 and 2); Exemption consider dynamic effects and loading (Docket No. 50-2751 conditions associated with previously I
postulated pipe breaks in the RCS pacific Gas and Electric Co. (Diablo Canyon Nuclear Power Plant. Unit 1).
On July 20.1973. the Texas Utilities primary loop. it cluding jet impringement a in t R an at he i rg Unde FR 2'2 endere an pp ca o fo icen e to construct Comanche Peak Steam and subcompartment pressure loads. In Notice is hereby given that the Electric Station. Units 1 and 2 support of this exemption request. the Director. Office of Nuclear Reactor (Comanche Peak cr the facility) with the apphcant's submittal enclosed l
Regulatian. has issued a decision Atomic Energs Commission (currently Westinghouse Report MT-SME-J135 l
roncernine Petitions dated February 2.
the Nuclear R'egulatory Commission or (Reference 1) containing the technical j
March 1. March 23. April 12. May 3. June the Commission). Following a public basis for their request.
- 21. juni.22. July 11. July 16. and July 23.
hearing before the Atomic Safety and Based on its review of the applicant's 1984 filed by the Government ~
Licensing Board, the Corr. mission issued submittal. the NRC staff requerted Accountability Project on behalf of the Construction Permit Nos. CPPR-120 and additional information and provided San Luis Obispo Mothers for Peace. The CPPR-127 permittmg the construction of comments on the reports (Refe ences 1 Petitioner requested that the Units 1 and 2. respectively, on December and 9) which were trahsmitted to the I
Commission defer all heensing decisions 19.1974.Each Unit of the facility is a applicant in the form of questions by I
on the Diablo Canyon Nuclear Power pressurized water reactor, combining a NRC letter dated March 2.1984.
Plant. Unit I until a number of specified Westinghouse Electric Company nuclear (Reference 2).
actions were taken including. inter olio.
steam supply system. located at the
- By a submittal dated April 23.19M.
a comprehensive third. party apphcant's site in Somervell/ Hood the applicant responded to the staffs reinspection of all safety.related Counties. Texas, approximetely 40 miles questions (Reference 2) and provided a equipment, an independent management southwest of Fort Worth. Texas.
revision to the Reference 1 report audit and a fullinvestigation of On February 27,1978, the applicant identified as Westinghouse Report questions of harassment. The Petitioner tendered an application for Operating WCAP-10527 (Reference 3). in a a!!eged numerous siolations of Licenses for each Unit of the !?cility, separate submittal also dated April 23.
Commission requirements as the basis currently in the licensing review 1984. the applicant provided a value-l for its request.The petitions were process, with Unit 1 licensing to occur in impact analysis which together with the i
referred to the Director. Office of the near term.
technicalinformation cootained in the I
,+
l
l Federal Register / Vol. 49. No.173 / Wednesday. September 5.19M / Notices 35059 Reference 3 report. provided a exemption from the regulations would on both Westinghouse and Combustion z
comprehensive lustification for be acceptable as an alternative for Engineering nuclear steam supply m*
requesting a partial exemption from the resolution of 1/S! A-2 for sixteen system mam loop piping (Reference 8) requirements of GDC 4.
facilities owned by eleven licensees in confirm that ooth the probability of From the deterministic fracture the Westinghouse Owner's Group (one leakage (e g.. undetected flaw g owth t
mechanics analysis contained in the of these facilities. Fort Calhoun has a through the pipe wall by fatigue) and the 4
techrucalirformation furni*hed. the Combustion En!nneering nuclear steam probability of a DEGB are very low. The y
applicant stated that the postulated supply system).This NRC. staff position results given in Reference 8 are that the double-ended guillotme breaks (DECB) was stated in Gene tc Letter 84-04, best. estimate leak probabilities for of the primary loop coolant piping will published on February 1.1984 Westmghouse nuclear steam supply not occur in Comanche Peak Units 1 and (Reference 4). The generic letter states system main loop piping range from t.2 4
2 and, therefore. need not be considered that the affected licensees must justify x 10-
- to 1.3 x 10-8per plant year and M
[
as a design basir for installing an exemption to CDC 4 on a plant-the best-estimate DEGB probabilities j-i protective structures such as pipe whip specific basis. Other PWR applicants or range from 1 x 10-" to 7 x 10-'8per y.
restraints and jet impingement shields, licensees may request similar plant year. Similarly, the bes& estimate
]
to guard against the dynamiceffects exemptions from the requirements of leak probabihties for Combustion 2 1 associated with such postulated breaks.
GDC 4 prosidet, that they submit an Engineenng nuclear steam supply
=
By letter dated June 7.1984 (Reference acceptable technical basis for system main loop pipeg range from 1 x 2,
~
10), the applicant clanfied the scope of elimmating the need to postulate pipe l0' "per plant year to 3 x 10-* per plant its request for exemption from CDC 4 breaks.
year. and the best-estimate DEGB requirements. Since the Westmghouse The acceptance of an exemption was probabilines range from 5 x 10-"to 5 x Q.
Report WCAP-10527 provided at 'yses made possible by the development of 10"2 per plant year. These results do not V
encompassing other structures in coth advanced fracture mechanics affect cora melt probabilities in any y.
Comanche Peak Units 1 and 2. and techr. ology. These advanced fracture sigmficant way.
~
seemed to be in conflict with the scope mechanics techniques deal with During the past few years it has also of the exemption requested in an earber relatively small flaws in piping become apparent that the requirement letter dated February 17.1984 components (either postulated or real) for installation of large. massive pipe (Reference 11), the applicant stated in and examine their behavior under c
whip restraints and jet impingement M
the Reference 10 letter that. although the various pipe loads. The, objective is to shields is r.ot necessarily the most cost analyses contained in the Report demonstrate by determmistic analyses effectise way to achieve the desired WCAP-10527 encompassed relief from that the detection of smatl flaws by level of safet'v. as indicated in Enclosure L
the need to mstall pipe break protective either inservice inspection or leakage
- 2. Regulatory Anatvsis to Reference 4.
r t,
devices in both Units 1 and 2. the monitoring systems is assured long Even for new plant's. these devices tend 7
f exemption being requested pertam, ed before the flaws can grow to critical or to restrict access for future inservice 3
solely to the mstallation of jet unstable sizes which could lead to large impmgement shields associated with treak areas such as the DEGB or its inspection of piping: or if they are i
removed and remstalled for inspection.
i e
such breaks in eight (a) locations per equivalent. The concept underlying such loop in Comanche Peak Unit 1. as analyses is referred to as " leak-before*
there is a potential nsk of damasing the specified in Section 4.0 of the value-break" (LBBl. There is no implication p ng I d ~
^-
c ponen p
ss ifinstalled impact analysis submitted by the that piping failures cannot occur, but p n P""g p
p~
l applicant's letter dated Apnl 23.1984.
rather that improved knowledge of the fadiP {o
/-
til failure modes of piping systems and the p
E
'd be
$s application of appropnate remedial incurred while pubhc r:sk reduction 3
L The Commission's regulations require measures. ifindicated. can reduce the w uld be very low Re nova 1 and 5
that applicants provide protective probability of catastrophic failure to mmstaHahon for msuuce mspecum I
measures apinst the dynamic effects of insignificant values.
also entail sunmcant ORE over the hie 7
l postulated pipe breaks in high energy Advanced fracture mechanics of a plant.
?
.(
fluid system piping. Protective measures technology was applied in topical IV j
inc!cd.e physical isolation from reports (References 5. 6 and 7) submitted
.y postulated pipe rupture locations if to the staff by Westinghouse on behalf The primary coolant sWem of feasible or the installation of pipe whip of the licensees belonging to the USI Comanche Peak Units 1 and 2. described
=
g[
4 restraints. let impingegnent shields or A-2 Owners Group. Although the top'.al in Reference 3. has four mairi loops each compartments. In 1975. concerns arose reports were intended to resolve the compnsing a 33.9 in-h diameter hot leg.
as to tne asymmetnc loads on issue of asymmetric blowdown loads a 36.2 inch diam. _. crossover leg and f,
y pressurized water reactor (PWR) vessels that resulted from a limited number of 32.14 inch diameter cold leg pipmg. The and their internals which could result discrete break locations. the technology materialin the primary !oop pipma is E,
from these large postulated breaks at advanced in these topical reports cast stainless steelISA 351 CF8Al. In its d
s discrete locahons in the main primary demonstrated that the probability of review of Reference 3. the staff
(
coolant loop piping. This led to the breaks occurnng in the primary coolant evaluated the Westinghause analyses establishment of Unresolved Safety system main loop piping is sufficient!y with regard to:
e lisue (USI) A-2. " Asymmetric low such that these breaks need not be
-The locatien of maximum stresses m Illowdown Loads on PWR Pnmary considered as a design basis for the piping. associa'ed with the Systems.,
requinng installation of pipe whip combined !aads from normal 4
1 5
The NRC staff. after several review restramts or jet impingement shields.
-7 meetings with the Advisory Committee The staffs Topical Report Evaluation is E"*
8i l
on Reactor Safeguards (ACRS) and a attached as Enclosure 1 to Reference 4.
-Potential cracking mechanisms:
h meeting with the NRC Committee to Probabilistic fracture mechanics
-Size of throuith wall crar.ks that would d
p Review Genenc Requirements (CRCR).
studies conducted by the Lawrence leak a detectable amount under concluded that for certa % facilities an Livermore National Laboratories (LLNL) normalloads and pressure
'i I
b d
i W
4 i1 2
T
e O
35060 Federal Register / Vol. 49. No.173 / Wednesday. September 5.198-1 / Notices
-Stability of a " leakage size crack" and the critical-size crcck to account for 1.45 and it can detect leakage of one (1) under normal plus SSE loads and the the uncertainties inherent m the gpm in one hour. The calculated leak espected margm in terms ofload; analyses and leakage detection rate through the postulaid flaw is large
-! irgin based oa crack size: and capability. A limit-lcad analysis may relative to the sensitivity of the
-The fracture toughness properties of suffice for this prupose. however. an Comanche Peak plant !eak detection thermally-aged cast stainless steel elastic-plastic fracture mechanics smem.
pipmg and weld material.
(tearing instability) analysis is (4)'T he npected margin in terms of The NRC staffs criteria for evaluation preferable.
loS6 for the leakage-size crack under of the abose parameters are delineated (GlThe materials data provided normal plus SSE loads is within the should include types of materials and bounds calculated by the staff m Section in its Topical Report Evaluation, to Reference 4. Section 4.1.
materials specificatons used for base t0.3 of Enclosure (1) to Reference 4 la metal. weldments and safe-ends. the addition. the staff found a significant "NRC Evaluation Criteria", and are as materials properties includmg the J-R ma.pr. in terms of loads larger than follows; curve used in the analyses. and long-m.nal plus SSE loads.
(11 The loading conditions should term effects such as thermal aging and f 511 he margin between the leakap include the static forces and moments other hmitations to valid data (e.g.
sm crack and the critical-size crack (pressure. deadweight and thermal I maximum. maximum crack growth) was caiculated. Again. the results expansion) due to normal operation, and demonstrated that a significent margin the forces and mome.ts associated with V the sJe shutdown eachquake (SSE).
Based on its evaluation of the analysis b"
These forces er d morm nts should be contained in Westinghouse Report located where the highe. I stresses.
WCAP-10527 (Reference 3), the staff (6) As an integral part of its revgew, comcident with the poorest material finds that the apphcant has presented the staffs evaluation of the matenal properties. are induced for base an acceptable technical justification.
prnerties data of Reference 9 ts materials. weHments and s. fe-ends.
addressing the above entena, for not enc! sed as Appendix 1 to this (2) For the piping run/syste..v under installing protective devices to deal with Exem; tion. In Reference 9. data for ten evaluation, all pertment mformation tn= b.=mic effects of large pipe (10 p!ar.ts. including the Comanche w..ich demonstrates that degradation or ruptures in the main loop primary Peak Units. are presented. and lower failure of the piping resulting from stress coolant system piping of Comanche bound or " worst case matenals corrosion cracking. fatigue or water Peak. Units 1 and 2. This finding is hammer is not likely shouid be predicated on the fact that eacn of the P' P'Tli'5.were identified and used m, provided. Relevant operating history parameters evaluated for Comanche the analysis performed in the Reference 3 report by Westinghouse.The staffs should be cited, which includes s3 stem Pe.k is enveloped by the generic operational procedures: system or analysis performed by Westmghouse in upper bound of 3000 in-lb/in'on the component modification: water Reference (5), and accepted by the staff 8 'vd j (refer to Appendix 1. page 6) w..
ct exceeded: the applied J for chemistry parameters. hmits and in Enclosure 1 to Ref erence 4.
Comanche Peak in Reference 3 was controls: resistance of material to specir,c,1;y venous for;ns of stress corrosion, and (1)The loads assor ated with the substantially less than 3000 in-!b/in' performance under cyclic loadmgs-highest location in the main loop in siew of the analytical results (3) A through-wall crack should be primay system piping are considerably presented in the Westinghouse Report postulated at the highest stressed lower than the boundmg loads used by for Comanche Peak (Reference 3) and locations determined from (1) above.
Westinghouse m Reference 5. or those the staffs evaluation findings related The size of the crack shculd be large estabbshed by the staff as hmits (e.g., a above. the staff concludes that the enough so that the leakage is assured of moment of 42.000 in kips in Enclosure 1 probability or likehhood oflarge pipe detection with adequate margin using to Reference 4).
breaks occurrmg in the primar) coolant tl e mmimum installed leak detection (2) For Westmghouse plants. there is system loop of Comanche Peak Units 1 capat,ihty when the pipe is subjected to no history of cracking failure in reactor and 2 is sufficiently low so such that normal operationalloads.
primary coolant systemloop piping.The such pipe breaks need not be considered (4) It should be demonstrated that the Westinghouse reactor coolant system as a design basis for requirmg protective postulated leakage crack is stable under primary loop has an operating h'istory devices. However, the pipe whip normal plus SSE loads for long periods which demonstrates its in'.ierent restraints have already been installed m of time. that is. crack growth. if any. is stability. This includes a low Unit 1. and the applicant has hmited the minimal dunng an earthquake. The susceptibihty to cracking failure from scope of its exemption request to the margin. in terms of applied loads. should the effects of corrosion le g.
installation of jet impingement shields m be determmed by a crack stability intergranular stress corrosion crackingl.
Urnt 1 only.The requested exemption analysis,i e that the leakage size crack water hammer, or fatigue (low and high from GDC 4 as limited to exemption from wdl not experience unstable crack cycle).This operating history totals over the need to install jet impingement
{
growth esen iflarger loads (larger than 400 reactor-years, including five plants shields at specified locations in Umt 1 design loads) are applied. This analy sis each havmg 15 years of operation and 15 The staff also reviewed the value-should demonstrate that crack growth is other planta with oser 10 years of impact analysis provided by the j
siable and the final crack size is limited.
operation.
applicant for not providing protective such that a double-ended pipe break (3)The results of the leak rate structures against postulated reactor wdl not occur.
calculations performed for Comanche coolant sistem loop pipe breaks to 1
(5) The cra:k size should be Pesk. using an initial through-wall crack assure as low as reasonably achievable j
determined by comparing leakage-size are identical to those of Enclosure 1 to (ALARA) exposure to plant personnel.
I cracks to critical stre cracks. Under Reference (4). the Comanche Peak plant Consideration was given to design normal plus SSE loads. it should be has an RCS pressure boundary leak features for reducing doses to personnel demonstrated that there is adequate detection system which is consistem who must operate. service and maintain margin between the leakage site crack with the guidelines of Regulatory Guide the Comanche Peak instrumentation.
l Federal Register / Vol. 49. No.173 / Wednesday. September 5.19M / Notices 35061 J
ccntrols, equipment. etc. Normally.
Addit onal inferrnanon C.,cernma ink.
specifed by the provisions of to CFR
{r f.it hties and equipment are designed to Before-Break An.h s:s for L nanche Paak 50.55alb) to Wisconsm E!e-'nc Power sa: e person ret.s: howeser. the Steam Electrc Station (Units 1 'nd 21. ' da:ed Company (the hcensee). for the Point 2
Comar.che Peak ulLe.mpact ana!)us Beach Nui ar Plant Unit No.1. locahd nouse Report WCAP-10527 shows that the addinon of pro'ectne
" Technical Bases for Ehmmanng Lage in the Towr. of Two Creeks. Manitowoc des ites for RCS pipe breaks will cost P-maa loop Pipe Rup. re as the Structural County. Wisconsm.
about 2 person rems annuaMy due to the Desyn Basis for Comanche Peak Units 1 and Environmental Assessment s owing down of normaily anticipated 2." Apnl 1984. W estmshouse Cl ass 2 L
work. Jnd increasing the scope of propretary.
identdication of Propased Acnon T
routme maintenance in radianon areas Al NRC Cenenc letter S4-04. '.,.ufety I
that would be mvolved. The analysis Evalcanon of W estinghouse Topic.i Reports De actioil would provide relief from h
provides a reasonab;e est: mate for this N'hrg wnh Ehminanon of Postulaied the requirnment to perform surface "O Md*
P' examinations of the safety miec: ion additional radiclogical ccst. In view of he s ery low probabihty of p:pe breaks tc[an reducer-to-safe end weids as required Fractu e Esaluation of it 'he spec:f<ed locations covered b)
Reactor Cooiant P're Conia nina a Pos'u!a'ed by Section XI of the ASME Boner and l
th:s exemption. the redwhon of Circumferennal Throuanwa'I Cr'ack. W CAP _
Pressure Vessel Code wh:ch,has been k
m "upahenal esposure resultmg from 9558. Rev. 2. May 1981[W estmahouse Cbss 2 incorpor' ted by reference m the k
'r s exemphon outweighs the potential proor'e t a ry.
require...ents of 51 CFR 50 55a rela me s
a cident exposure reduction that might
- 16) Tensde and Twahnes Propert es c' to inservice Ir.spection of Safety Re.ated t
asult from mstallation of the let Pnmary P' ping W eld Metal for Use a Components. Volumetnc exam.na aor.s Mechanist:c Fracture Es aluahon. WCAP-inpingement barriers.
of these welds would be performed 9*87. May 1981. Westinghouse Class 2 eVEry 10 vears as required.
\\l propr: eta ry in Wunnahouse Reponse to @nnons The Need for the Proposed Actwn In view of the staffs evaluation and Comments Raised by Members of ACRS
'mdings. conclusions, and Subcommittee on Metal Component Dunng
'he proposed relief ir required racommendation above, the Commission the Westmahouse Presentanon on Scotember because surface examinations of these has determined that, pursuant to 10 CFR 25.1981. Letter Report NS-EF R-2519. E.P-welds are not possible due to the M 12(a). this Exemption is authonzed by Rahe to Darrell G Eisenhut. Nosember 10.
maccessibility of the weld surfaces The J
- sw md will not endanger hfe or 1981 Wesunghouse Class 2 propnetary.
welds are located between the reactor property or the common defense and
]e 0g
- rmgre Nat on w nnce vess and the biological sh:e'd wail.
i are secunty. and is otherwise m the pubbc p
mterest The Commission hereby f
y Q
- l^
Enuronmentalimpact.s of the Propowd 4
C 9
n
^
approses the requested limited OU and C K. Chou t' arm v :w ;Prepnnt of a nemption fron GDC 4 of Appendix A paper intended fo wacan n).
The proposed reliefis aHowed by the to 10 CFR Part 50. to permit the licensee t91 % esonghouse > m A CAP-1NM provisions of 10 CFR 50.55ag)(6){i) not to install et impint ement shields "The Effects of Thermal Asma on the l
~
where the tests or exammations associated mth pstulated pipe breaks Structuraiin centy of Cast Simniess Steen of the eight (81 locations per loop m the P!pma for Westmphouse %ciear Steam required by the code are determmed f
Comanche Peak Unit 1 pnmary coolant Supply Systems." Nos emt>er 1983.
impractical to perform. As the surfaces sy stem. as specified in Sectiori 4 0 of the Wnnna5 use Class 2 propne'a v of the welds m question are value-impact analy sis submit by the M "" U"hnes Gene ancg Company inaccessible. a surface examinahon has un
-4
. Request for Parnal been determined by the hcensee and apphcant's letter dated Apnl 23,1984.
I-ev luated by the Commission as g
This Exemption does not pertam to the oune ood ed n 19 imPract:ral to perfccm. The staff has 9
mstallation of pipe whip restraints.
(1U Texas Unhties Generanna Company j{
deteramed that the required salumemc already mstalled m Unit 1. or to the letter TXX4118. " Request for Partial mstallation of p<pe whip restraints and Exempnon." (R l Gary to B l. Yourgbiood) mspect:on of the welds once esery to iet impmgemert shields m Comanche dated February 17.1944.
years will provide adequate assurance d h mwal egg M & wh e
Peak Umt 2. The portion of the aquest g,f.e,nc.,
Unit 2 wdl be dealt with m a separate Ide itical relief to that requested for Unit N '* N n-PT P"8 rv ' '' n ' o f 1 was provided for Pomt Beach Umt 2 f
NRC action h*fs"RN by the Commission's Safety Evaluahon g
The Comenission has determmed that c
umen Room as 1 s
and letter of March 29,1064.
s:
the issuance of the exemption wdl have-(H Nfr-SE3136. (3) WCAP 10528. (5) no sigmficant environmental impact on WCAP 9570.16) WCAP 9788. (7) Non.
Consequently, as the Commission has the env.runment (49 FR 33945).
propnetary ser ion attached to the Letter determined that the welds will retain Report. (9) W CAP 10457-adequate structural mtegnty utihzmg the IF Dated at Bethesda. Maryland this 28th day of Aueust 19a4 miD* a-msi N S+a m e licensee's proposed alternate For the Nuclear Regulators Commiss:on.
sesso coot rssiHn-es examination (volumetnc examination Frank Miragba, once every 10 years), the probabihty of g
W UUN W
Droun Direcror Omsmn of Lcensmg. Omce tnocy,,nn,$n, ssl o' Nuclear Reactor R<yviarmn.
significantly and the consequences of References Wisconsin Electric Power Co.;
p st weld failure radiological releases i
Environmental Assessment and will not be greater than previously (1) W.estmghouse Report VT-SME-3135.
determmed nor does the requested relief Technical Baws for Ehmmai'ne large Finding of No Significant Impact g
ral oma [he P7a The U S. Nuclear Regulatory effluents. Thuefore. the Commission E
e7Ba e for t
n
- 2. ' October 1983. Westmghouse Class 2 Commission (the Commission) is has determined that there are no
[
propnetary considenng issuance of relief from the sign ficant radiological environmental
=
(2) t.etter to R l Gary of Texas Unlities requirements of Section XI of the ASME impacts associated with the requested i
.l Generanng Company. Request for Boiler and Pressure Vessel Code as re hef.
a
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UNITED STATES
! %,_ g,,g
' NUCLEAR REGULATORY COMMISSION
/
p a.3.u%gNye /
v wsa$roororo,o.c.2osss
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1 SEP 211984 ATTACHMENT G Docket Nos.: 50-445 and 50-446 Mr. M. D. Spence President Texas Utilities Generating Company 400 N. Olive Street Lock Box 81 Dallas, Texas 75,201
~
Dear Mr. Spence:
Subject:
Request for Additional Informat' ion Concerning the Handling of Heavy Loads at the Comanche Peak Steam Electric Stations (Un.its 1 and 2)
As a part of its continufng review of the Comanche Peak FSAR, the staff has requested that the following additional information be provided regarding the handling of heavy loads:
1.
Describe the m2ans provided to assure the integrity of concrete structures, lifting eyes, and any other heavy loads so that they.
will not fall apart while being handled during refueling should the lifting eye fail or the plug impact other structures.
2.
Alternatively, describe the consequences of failure of concrete.
structures or other heavy loads during handling.
Your response should contain an evaluation to confirm that uhacceptable fuel damage or damage to safety related equipment will not occur.
Your staff was advised of this forthcoming request by telephone.
This letter serves to fonnally document the staff's r? quest.
Please advise Mr. John Stefano of my staff whe1 we nay expect to receive ycur response upon receipt of this letter.
- Sincejel,
[
hkk{
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B'. J. Y ungbl d, Chief Licedsi 9 Br'.ch No. 1 Division of Licen-ing cc: See next.page
Luv.AWCHE PEAK ~
Mr. M. D. Spence-President Texas Utilities Generating Company 400 N. Olive St., L.B. 81 Dallas, Texas 75201 cc: Nicholas S. Reynolds, Esq.
Mr. James E. Cummins Bishop, Liberman, Cook, Resident Inspector / Comanche Peak Purcell & Reynolds Nuclear Power Station 1200 Seventeenth Street, N. W.
c/o U. S. Nuclear Regulatory Washington, D. C. 20036 Commission P. O. Box 38 i
Robert A. Wooldridge, Esq.
Glen Rose, Texas 76043 Worsham, Forsythe, Sampels &
Wooldridge Mr. John T. Collins 2001 Bryan Tower, Suite 2500 U. S. NRC, Region IV Dallas, Texas 75201 611 Ryan Plaza Drive Suite 1000 Mr. Homer C. Schmidt Arlington, Texas 76011 Manager - Nuclear Services Texas Utilities Generating Company Mr. Lanny Alan Sinkin Skyway Tower 114 W. 7th, Suite 220 400 North Olive Street Austin, Texas 78701 L. B. 81 DalTii, Texas 75201 B. R.'Clements Vice President Nuclear Mr..H. R. Rock Texas Utilities Generating Company Gibbs and Pell, Inc.
Skyway' Tower
~~
l 393 Seventh Avenue 400 North Olive Street New York, New York 10001 L. B. 81 Dallas, Texas 75201 Mr. A. T. Parker Westinghouse Electric Corporation William A. Burchette, Esq.
P. O. Box 355 1200 New Hampshire Avenue...N. W.
Pittsburgh, Pennsylvania 15230 Suite 420 Washingto6, D. C.
20036 Renea Hicks, Esq.
Assistant Attorney General Ms. Billie Pirner Garde Environmental Protection Division Citizens Clinic Director P. O. Box 12548, Capitol Station Government Accountability Project Austin, Texas. 78.711 1901 Que Street, N. W.
Washington, D. C.
20009 Mrs. Juanita Ellis, President Citizens Association for Sound David R. Pigott, Esq.
Energy Orrick, Herrington & Sutcliffe 1426 South Polk 600 Montgomery Street Dallas, Texas 75224 San Francisco, California 94111 Ms. Nancy H. Williams Anthony Z. Roisman, Esq.
CYGNA Trial Lawyers for Public Justice 101 California Street 2000 P. Street, N. W.
San Francisco, California 94111 Suite 611 Washington, D. C. 20036
.s.
- _ ~
t l
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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TEXAS UTILITIES ELECTRIC
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Docket Nos. 50-445-1 COMPANY, et al.
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and 50-446-1 (Comanche Peak Steam Electric
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Station, Units 1 and 2)
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CERTIFICATE OF SERVICE Ey my signature below, I hereby certify that true and correct copies of CASE's Answer to Applicants' 9/13/84 Supplement to Motion for Authorization Pursuant to 10 C.F.R. 50.57(c); and CASE's Answer to Applicants' Reply to CASE's Answer to Applicants' Motion for Summary Disposition Regarding Consideration of Friction Forces have been sent to the names listed below this 34 hayof October,1984,
by:
Express Mail where indicated by *de Hearings where indicated by **
and First Class Mail elsewaere.
Hand-delivered at operating Lice
- Administrative Judge Peter B. Bloch *
- Nicholas S. Reynolds, Esq. (both hand-del.
U. S. Nuclear Regulatory Commission Bishop, Liberman, Cook, Purcell & mailed) 4350 East / West Highway, 4th Floor
& Reynolds Bethesda, Maryland 20814 1200 - 17th St., N. W.
Washington, D.C.
20036
- Ms. Ellen Ginsberg, Las Clerk U. S. Nuclear Regulatory Commission *
- Geary S. Mizuno, Esq.
4350 East / West Highway, 4th Floor Office of Executive Legal Bethesda, Maryland 20814 Director U. S. Nuclear Regulatory
- Dr. Kenneth A. McCollon, Dean Commission Division of Engineering, Maryland National Bank Bldg.
Architecture and Technology
- Room 10105 Oklahoma State University 7735 Old Georgetown Road Stillwater, Oklahoma 74074 Bethesda, Maryland 20814
- Dr. Walter H. Jntdan Chairman, Atomic Safety and Licensing 881 W. Outer Drive Board Panel Oak Ridge, Tennessee 37830 U. S. Nuclear Regulatory Commission Washington, D. C.
20555 i
1 I
Chairman Renea Ricks, Esq.
Atomic Safety and Licensing Appeal Assistant Attorney General Board Panel Environmental Protection Division U. S. Nuclear Regulatory Commission Supreme Court Building Washington, D. C.
20555 Austin, Texas 78711 John Collins Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 611 Ryan Plaza Dr., Suite 1000 Arlington, Texas 76011 Lanny A. Sinkin 114 W. 7th, Suite 220 Austin, Texas 787C1 Dr. David H. Boltz 2012' S. Polk Dallas, Texas 75224 Michael D. Spence, President Texas Utilities Generating Company Skyway Tower l
400 North Olive St., L.B. 81 Dallas, Texas 75201 Docketing and Service Section (3 copies)
Office of the Secretary U. S. Nuclear Regulatory Commission Washinge.on, D. C.
20555 J
ff/. L s
/ d-fjVa.) Juanita Ellis, President CASE (Citizens Association for Sound Energy) 1426 S. Polk Dallas, Texas 75224 214/946-9446 2
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