ML20024F260
| ML20024F260 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 08/25/1983 |
| From: | Hodgdon A NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | LEWIS, M. |
| Shared Package | |
| ML20024F258 | List: |
| References | |
| NUDOCS 8309090140 | |
| Download: ML20024F260 (7) | |
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8/25/83 UtilTED STATES OF AliERICA NUCLEAR REGULATORY C0 lit!ISSION 7
i BEFORE THE AT0!11C SAFETY Af;D LICENSING BOARD i
In the liatter of
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PilILADELPlilA ELECTRIC C0!!PANY
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Docket Nos. 50-352 i
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50-353 (Limerick Generating Station,
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Units 1 and 2)
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NRC STAFF RESP 0llSE TO IliTERVEliOR LEWIS'S " THIRD AND FINAL SET OF IfiTERP0 GAT 0 RIES TO THE NRC STAFF At D LICENSEE" 1.
IllTRODUCTION 4
Pursuant to the Licensing Board's Special Prehearing Conference Order of June 1,1982 and Itemorandum and Order Confirming Schedules Estcblished During Prehearing Confercnce of 11ay 16,1983, Intervenor 4
1 Marvin Lewis Propounded his " Third end Final Set of Interrogatories to the NRC Staff and Licensee" on August 1, 1983.
The llRC Staff's responses to fir. Lewis' Interrogatories are set forth below.
II. RESPONSES TO INTERR0GATORIES /
I Interrogatory 1[A]
1.
In the "fiRC Response to the first set of interrogatories on contention I-62", the staff states, "the basic conditions under which BWRs operate make it cuch less likely for BURS than for PWRs that the simultaneous rapid cooling and high pressure necessary to create a PTS will occur."
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Interrogatories 2 and 6 are directed to the Applicant and are, therefore, not addressed in these responses.
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The response leaves open the question, "Are there cny set of conditions wherein a Bl R can experience the sinultaneous rcpid cooling and high pressure necessary to create a significant PTS?"
Further, can a BWR experier.ce a cooling (rapid or not) which will prcduce significance stresses at tenperatures close to rtndt in any temperature - pressure boundary?
Please answer above two questions regarding cny set of conditions that can produce'a PTS in a BWR.
NRC Staff Response A BUR operates with the primary system " saturated".
That is, the steam-water coolant mixture inside the pressure vessel is at the pressure determined by boiling cnd steam formation in the core region where the nuclear fuel is providing a heat source. Any ccid water introduction into such a systen will reduce the steam formation or result in steam condensation, which will lower the pressure. Thus, it is not possible to postulate conditions for such a system that will result in sinultaneous cooling and pressurization.
The above applies to cooling - rapid cr otherwise.
The effect is cost preacunced for rapid cooling events, for which significant thermal stress (a necessary ingredient for the PTS concern) could be developed in the pressure vessel. But the effect discussed above prevents occurrence of high pressure for such rapid cooling events, and high pressure is another necessary ingredient for a significant PTS concern (i.e. one cust have both rapid cooling resulting in thermal stress and high pressure).
For "not rapio" coolins events there is no significant thertal stress, therefore no significcnt PTS concern.
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Interrogatory 1[B]
Also refer to IIC Meacrandum UNANALYZED REACTOR VESSEL THERMAL STRESS DURIllG C00LDOWN from Eisenhut for Ccamissioners dated April 12, 1983.
Discuss whether this type of stratification could cccur in a PWR and its effect upon thermal stresses.
Include in your discussion any staff concerns on the operability of safety relief valves which could complicate the thernal picture. Also discuss hcw natural circulation can fail or be reduced.
Dccument request related to above interrogatory.
Please provide GE HEE 24988 P, " Analysis of Generic EUR Safety Relief Valve Operability Test Results."
I do not know the date nor availability as I have only read a short synopsis.
NRC Staff Response Pursuant to an agreement reached in a telephone conversation between Intervenor Lewis and Staff Counsel, this question is addressed in the letter from Staff Counsel which covers these responses.
Interrogatory 3 Recently the NRC released an order concerning the ongoing IGSCC in BURS.
I have attempted to get further information on this without success. Therefore, I am submitting the folicwing interrogatory on the relationship of IGSCC to PTS in BURS.
Are any structures which receive neutron irradiation subject to IGSCC7 Discuss the feedwater nozzle and piping as mentioned on page 5.3-7 of the LGS FSAR, Para. 5.3.1.5.3.
a.
Has all "new" information on unresolved safety issues been factored into the PTS problem as the information has become available?
b.
Have all synergistic or cumulative effects from other USI been factored into the PTS problem and its consideration by the Licensee and flRC staff? Give specific examples. Discuss USI of " cold overpressurization" in your answer.
c.
Are there any other concerns not covered in the USI which can or do' have an effect on the consideration of PTS been considered adequately at LGS? Discuss measurenent of neutron flux (PECo Boyer to Eisenhut, April 15,1983, Page 4, last paragraph) and difficulties of ultrasonic testing (URC Schwer.cer to PEco Bauer, June 3,1983, liEB enclosure Page 250-8 Paragraph 250.4 and 250.3.A.)
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_4 liRC Staff Response Scae structures which receive neutron irradiction are subject to IGSCC (intergranular stress corrosion cracking).
Recent NRC releases concerning IGSCC in BWRs were for BUR recirculation piping which was fabricated fron, austenitic grade 304 stainless steel. This caterial was sensitized during welding and became susceptible to IGSCC. The internal core support structures inside the Limerick reactor vessels are constructed of austenitic grade 304 stainless steel and receive considerable neutron irrediation.
Inservice examination of BUR internal core support structures at other facilities, which is performed in accordance with Table IWB-2500 of Section XI, ASME Code, indicates IGSCC is not a problem for such structures. The BUR recirculation piping outside the vessel receives insignificant neutron irradiation.
Feedwater nozzle and piping do not contain sensitized austenitic
_ grade 304 stainless steel and are not located in the reactor vessel beltline.
Hence, feedwater nozzle and piping are not susceptible to either IGSCC or significant neutron irradiation.
a.
New information is being and will continue to be factored into resolving USI A-49.
b.
The Staff 'is exploring in USI A-49 all such effects associated with PTS.
Pressurized Thermal Shock (PTS) is the name given to certain accident conditions which result in high thermal and pressure stresses.
" Cold overpressurization" is an event which occurs at Icu tepperatures, has high pressure stresses and no thermal stresses.
/.s a result, a PTS event would be r: ore limiting for reactor vessel integrity than a " cold overpressurization" event.
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The Staff believes that all concerns regarding PTS have been adequately considered for Limerick. The amount of neutron flux affects the emount of embrittlement. This is discussed in Commission Report SECY 82-465, " Pressurized Thermal Shock," which has previously been provided to you. Ultraso,nic inspection is a method used to determine whether cracks exist in a vessel. The minimum size of defect that can be detected depends on the ultrasonic inspection method. Since the minimum size of detectable crack varies with inspection method, all size cracks are
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considered in USI A-49.
Interrogatory 4 4.
In the NRC Staff Response to Intervenor Lewis's First Set of Interrogatories in PTS Contention (I-62), the staff states, "For BWRs plants including Lin.erick, the location of fluid systems injection does not result in direct impingement on the vessel wall." Page 3.
What does fluid system impinge directly upon? Are any of these structures able to fail in a PTS situation or during a transient? Are any of these structures made of a material which will change RTndt with neutron flux?
Also provide a drawing of a jet pump and a description of its functicn and material of construction.
NRC Staff Response The " fluid systems" (high pressure emergency cooling system) inject i
directly into the core region inside the core shroud and through the feedwater sparger radially inward away from the vessel wall..The stainless steel and fuel cladding materials in that region are designed to withstand the resulting thernal shock.
PTS involves failure of the vessel due to combined thermal and pressure induced stresses. Thermal shock of the core is not related to PTS.
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The injection location for the " fluid systems" through the feedwater spcrgers, which are located above the dcunconer cutside the core shroud, allcws considerable mixing of the cold water with the warmer water cutside the core shroud before the nixture contacts the pressure vessel vali, thus precluding a significant PTS concern.
Drawings of the jet pumps and descriptions cf the material used in the reactor cooling system may be found in the FSAR.
s Interrogatory 5 To the staff : Document request: Please provide CE I:EDO 10029, "An Analytical Study of Crittle Fracture of GE EWR Vessels Subject to a Design Basis Accident."
NRC Staff Response See Response to Interrogatory 1[B].
Interrogatory 7 In my Interrogatory 6 of my first set of Interrogatories I was looking for a yes or no answer.
Please provide a yes or no answer to the following repeat interrogatory.
Have any "tect coupons" of affected materials been irradiated and tested from BWRs of design similar to Limerick?
a.
Reference E-7 has not yet been published.
I therefore cannot make a document request. Hopefully you can provide sorte information nonetheless.
Please provide the number of Charpy test which were used to l
determine the Guthrie trend curve, standard deviation, and 2 sigma upper bound.
If possible, provide the statistical value of " confidence" for the above data. Provide the reference from which the statistical values were developed.
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liRC-S_taffReshense Irradiated test coupens have been tested from BURS of design similar to Limerick.
a.
The Guthrie trend curve was developed frcm 136 data points.
There were approximately 15 Charpy V-flotch tests per data point. The e
Guthrie trend curve has a standard deviation of 124.3 F and a 2 sigma of 1 48.CF. The Guthrie data were published in f'UREG-CR 2805 Vol. I, HEDL-Tl4E 82-18, Quarterly Progress Report Jan.-liarch 1982.
This publicationdoesnotidentifythe" confidence"forthea$ovedata.
Respectfully submitted, iO y
Ann P. Mcdsdon i
Counsel for tiRC Staff i
Dated at Bethesda, liaryland
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this 25th day of August 1983 t
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