ML19206A130
| ML19206A130 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/05/1977 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Grossman M NRC Office of the Executive Legal Director (OELD) |
| References | |
| NUDOCS 7904180269 | |
| Download: ML19206A130 (6) | |
Text
,g ~
-1 y,
- c
_c
~
W MEMORANDUM FOR: Milton Grossman, Hearing Division, Director and Chief Counsel, Offir.e of tile Executive Legal Director FROM:
D. B. Vassallo, Assistant Dirqctor for Light Water Reactors, Division of Projeqt Management
SUBJECT:
BOARD HOTIFICATION - THREE MIy! ISLANC 2 I recoinnend notifying the Three Mile Island 2 Board of two recent matters involving (1) steam generator tube investigations and (2) additional analyses on the steam line break accident.
The first matter covers a prcgram being undertaken by the applicant in connection with steam generator tube degradation. The Board notification is for their information. The staff expects to gain useful information from this overall program and we will review the additional instrumentation and tube sleeving required for the testing to assure that the facility safety is not cc::: premised.
The second c:atter concerns a new worst-case analysis for the period inmediately following a steant line break wherein preliminary staff calculations indicate excessive radioactive releases.
Proposed Board letters are enclosed for each of these matters.
Original sipd ':y D.B.Yassslia D. B. Vass_ allo, Assistant Director for Light Water Reactors Division of Project Management
Enclosures:
2 Proposed Letters cc w/ enclosure-Distribution:
ase R. Boyd S-) Central Files f.
D. Vassallo R. Mattson F. Williams H. Denton H. Smith R. DeYoung J. Stolz BNP File M u_.
K. Kniel O. Parr 3, y V_ W 11e K. G611er
/'#
- "T=,qaggg, n LMRiTC LWR 90 - tWW F#91(as DVassallo &/t r
....a fl.
l 6 vw
.m.
12/
/77 12/ V /77 NRC FORM 318 (9 76) NRCM 0240 I? u. s. move==esauf emnvime orwicas ie7e-eawa4 4S 130
~19 6 11 9,.a (/7 7
u ORAFT FROPOSED 30ARD LETT_E.R._
This is to inform you of actions being taken by the applicant at TMI-2 in connection with Babcock & Wilcox (3&W) steam generator tube degradation.
In the recent past, some 5.5W once-through steam generators in ocer-ation at other facilities have shown indications of aoparent tube degradation. These indications appear to be clustered in the " flow lane," between the upper tuce support plate and the upper tube sneet, although some indications have been evicent in other areas. At eacn facility as appropriate, inspection and corrective action nas been performed in accordance with Regulatory Guide 1.83 anc 1.121.
We have been discussing this situation with the applicants involved and with 3&W, and various investigatory programs are underway.
Since TMI-2 is not yet operational, tne applicant has elected to undertake three investigatory programs to help define the cause of tne problem and to examine the effects of one possible fix, as descrioed further below.
1.
Steam Generator Instrumentation Procram The applicant proposes to install accelercmeters, strain gages, and pressure sensors in selected tubes and other po:itions to provide greater knowledge of actual ooerating parameters relative
,o* r~ ~
n
. to structural and thermodynamic responses.
In a recent meeting, the program was described; it was noted that tne instrumentation would be removed after the first cycle, and it was stated that the instrumentation would not affect the safety of the plant.
We requestea appropriate documentation; thus far a safety evaluation has been received, but various suppcrtive information is still forthcoming. We will review all pertinent information when received, and will assure that tne program can produce Jseful results and that the safety of the facility is not compromised.
2.
Steam Generator Tube Sleeving Programs In a meeting on November 22, 1977 and in a letter dated.'iovember 21, the applicant stated his intention to initiate a tube sleeving test program utilizing sleeves designed by botn Br.! anc C;m ustian Engineering.
Bench tests have indicated tnat sleeves wnich stiffen the tubes in the area where difficulties have been excerier.ced have produced ceneficial results.
The program would include different sleeve arrangements instailec in a number of tuces at IMI-2, approcriate instrumentation, and cost-test tube examination. Descriptive documentation and safety analyses are expected in early Decemoer. We will review this information to assure the safety of the facility, tot a
J v
a preliminary assessment of the information presently availacie indicates no significant concerns.
3.
Baseline Tube Examination In addition to an eddy current examination of a small numcer of tubes prior to hot functional testing (HFT), t1e applicant decided to examine all tubes of both steam generators after HFT prior to actual operation. The "A" generator examination was comoleted quite recently.
Preliminary information indicates eight tubes with greater than 40% wall reduction, approximately 40 tubes with about 20~ wall recuction, 400 tubes witn indenta-tions of 1 to 3 mils, and six tuoes v.hich did not allow passage of the inspection probe. The indications seem rancomly located, except tnat the more severe wall reductions appear located between the bottom tube sheet and tne first suoport plate.
The applicant intends shortly to remove one tube frca each defect category for furtner detailed examination, and to croceed witr scdy current examinc' ion of tne "B" steam generator.
This complete base line examination has not oeen creviously performed on any oths r B&W steam generator nor is it present'.y required by NRC, but we bel'ieve tnat it will provide a much imoroved basis for e valuating any future tube damage.
4r133
n.
4_
As ciscussed above, the staff is following the details anc safety inclications of this investigatory program tnrougn meetings anc documentation requests. We anticipate tnat tne program will yielc useful information regarding the general area of tu::e degradation and will follow the results carefully.
Lk,7 su-
o.
DRAF-m a
PROPOSED _ LETTER 70 30ARD As requested, the applicant has submitted additional analyses of :ne steam line break to deme.: strate that such an event during tne first year of operation would not present an undue risk to the health and safety of tne public. The applicant is committed to incorporate appropriate safety grade equipment at the first refueling.
Staff concerns had centered on possible fuel damage during a subcritical return to power. The analyses indicate that no additional fuel failures are predicted to occur during this time frame.
In addition, the applicant has introduced a new worst-case analysis for the period immediately after the break. The results of tnis analysis indicate that 15 percent of the fuel rods undergo a DN5R less than 1.3 following reactor trip.
Preliminary staff calculations of tais event indicate that unacceptable radioactive releases would result.
The acplicant maintains that neither the releases nor fuel camage c:culo be excessive.
We are pursuing resolution of this matter witn the applicant and Daccock & Wilcox, the reactor vendor. When accro:riate,
,,e wil' re: Ort changes in status of this problem to :ne Board.
ti$1 I.3Ei