ML20056D937
| ML20056D937 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 07/30/1993 |
| From: | Dromerick A Office of Nuclear Reactor Regulation |
| To: | J. J. Barton GENERAL PUBLIC UTILITIES CORP. |
| References | |
| GL-88-20, TAC-M74443, NUDOCS 9308190030 | |
| Download: ML20056D937 (8) | |
Text
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July 30, 1993 Docket No. 50-219 Mr. John J. Barton Vice President and Director GPU Nuclear Corporation Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731
Dear Mr. Barton:
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SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON OYSTER CREEK NUCLEAR GENERATING STATION - INDIVIDUAL PLANT EXAMINATION (IPE) SUBMITTAL -
GENERIC LETTER 88-20 (TAC N0 M74443)
By letter dated August 14, 1992, you submitted the-0yster Creek Nuclear l
Generating Station IPE results for NRC review.
Based on our review of your sub:nittal, we have determined that we need additional information to continue with our review. The enclosed list of questions identifies the information we need. We have found with other licensees that a conference call is very useful prior to submitting your formal response to these questions.
Please review these questions so that wa can schedule a conference call in about 20-30 days to discuss your responses. The purpose of the call is to give your staff and ours an opportunity to understand the. issues and what information the NRC needs to finish this review.
It is likely that the call can result in more direct (possibly shorter) written answers to the questions in-some cases.
We will require your written response within 60 days of the date of this letter.
This requirement affects one respondent and, therefore, is not subject to Office of Management and Budget review under P.L.96-511.
Please contact me should you have any questions regarding this request.
Sincerely, Original _ signed by Alexander W. Dromerick, Sr. Project Manager 9308190030 930730 Project Directorate I-4 PDR ADOCK 05000219 Division of Reactor Projects - I/II' P
pop Office of Nuclear Reactor Regulation
Enclosure:
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WASHINGTON, D.C. 20$55M1 July 30,1993 Docket No. 50-219 Mr. John J. Barton Vice President and Director GPU Nuclear Corporation Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731
Dear-Mr. Barton:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION ON OYSTER CREEK NUCLEAR GENERATING STATION - INDIVIDUAL PLANT EXAMINATION (IPE) SUBMITTAL'-
GENERIC LETTER 88-20 (TAC NO M74443)
By letter dated August 14, 1992, you submitted the Oyster Creek Nuclear Generating Station IPE results for NRC' review. Based on our review of your submittal, we have determined that we need additional information to continue with our review. The enclosed list of questions identifies the information we need. We have found with other licensees that a conference call: is very useful prior to submitting your formal response to these questions.
Please review these questions so that we can schedule a conference call in about 20-30 days to discuss your responses. The purpose of the~ call is to give your' staff and ours an opportunity to understand the issues and what information the NRC needs to finish this review.
It is likely that the call can result in more direct (possibly shorter) written answers to the questions'in some cases.
We will require your written response within 60 days of the date of this' 4
letter.
This requirement affects one respondent and, therefore,-is not subject to Office of Management and Budget review under P.L.96-511.
Please contact me should you have any questions regarding this request'.
Sincerely, r
Alexander W. Dromerick, Sr. Project Manag er -
Project Directorate I -Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
See next page
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Mr. John J. Barton 0yster Creek Nuclear GPU Nuclear Corporation Generating Station.
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Ernest L. Blake, Jr., Esquire Resident Inspector Shaw, Pittman, Potts & Trowbridge c/o U.S. Nuclear Regulatory Commission 2300 N Street, NW.
Post Office Box 445 Washington, DC 20037 Forked River, New Jersey 08731 Regional Administrator, Region I Kent Tosch, Chief U.S. Nuclear Regulatory Commission New Jersey Department of 475 Allendale Road Environmental Protection King of Prussia, Pennsylvania 19406 Bureau of Nuclear Engineering CN 415 BWR Licensing Manager Trenton, New Jersey 08625 GPU Nuclear Corporation 1 Upper Pond Road Mr. John J. Barton Parsippany, New Jersey 07054 Vice President and Director GPU Nuclear Corporation Mayor Oyster Creek Nuclear Generating Station Lacey Township Post Office Box 388 818 West Lacey Road Forked River, New Jersey 08731 Forked River, New Jersey 08731 Licensing Manager Dyster Creek Nuclear Generating Station
' Mail Stop: Site Emergency Bldg.
Post Office Box 388 Forked River, New Jersey 08731 t'
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GPU Nuclear Corporation Oyster Creek Nuclear Generatina Station Docket No. 50-219 Reauest for Additional Information - Oyster Creek Individual Plant Examination G-1.
In addition to NUREG-1150, seven other PRAs were reviewed. However, the insights from these PRA reviews are not included in the submittal, nor compared against findings of the Oyster Creek IPE. How were these insights integrated into the Oyster Creek IPE effort?
G-2.
The rule-based methodology with large logic module is difficult to follow, particularly without event trees for the front-end analysis.
Treatment of special sequences, such as ATWS events, are not transparent and without the event trees, it was not clear how the event sequences could be easily reviewed.
For the proposed plant changes and modifications, please explain how these changes will be incorporated into the model, and discuss the familiarity of the utility staff with this methodology.
G-3.
In conjunction with the above question G-2, please provide an example of actual quantification of the model.
FE-1.
Due to the methodology on the rules-modules approach, the contribution of common cause events to the CDF can not be easily derived.
Please discuss the significance of common cause events with respect to the core damage frequency, and include explicitly the treatment of the diesel generator failures.
FE-2.
Although the bases for all success criteria were provided in the development process of the rules, there is no overall summary of the success criteria and it is not clear where the analyses diverged from the FSAR.
For example, the success criteria for the core spray system in the FSAR chapter 15 analysis require two main pumps and one booster pump be operable.
However, IPE assumes that only one main and one booster pump may be operable.
Per NUREG-1335, please provide the r
bases for the success criteria.
FE-3.
Please explain and clarify the linkages between modules via the event sequence diagram (ESD). The loss of feedwater control module is transferred to ESD LT3c, which does not ey'.st.
FE-4.
Please discuss the long term consequences of the loss of reactor building closed cooling water to the recirculation pump seal cooling and potential for the LOCA during transient, especially involving station blackout. To what extent has induced pump seal LOCA been considered in the station blackout analysis?
FE-5.
The submittal, in its sensitivity study concluded that relaxing assumptions for the EMRVs would reduce total CDF by 18%.
Please clarify and identify the details of the assumptions for the 18%
. reduction.
FE-6.
The submittal stated that the failure data sources for Oyster Creek l
came from the operating experience (surveillance and operating procedures) and plant specific data over a period of approximately 11 years (1978-1989: Tables 4.3.1 through 4.3.7).
However, it appears that the vast majority of the failure data are generic data, and no specific discussions are presented in the submittal on the criteria for using plant specific and generic data, particularly Bayesian update of the generic data. Please identify and discuss your criteria for the use of generic vs plant specific data.
Provide the rationale for instances in which generic data is used in lieu of available plant specific data.
FE-7.
The submittal does not provide rationale or justification for screening out certain events.
Please provide additional information on the screening process of the following events:
a.
core spray failure following small LOCA to the reactor building (page 7.1-19, section 7.1.2.2 of the submittal).
b.
feedwater line breaks outside containment as an initiating event.
c.
line breaks at__the bottom of the vessel, e.g., CRD penetration and instrument lines, as an initiating event.
FE-8.
Please idr,tify those HVAC systems assumed not to be required, and provide justification for the assumption (Section 2.1.4).
Please provide further justification for the assumption that HVAC failures are not initiating events.
[ Initiating events are discussed in Section 2.1.3, guideline #1 of NUREG 1335.]
FE-9.
Please provide your rationale for limiting the discussion of DHR to ultimate heat removal, excluding discussions of injection / core cooling.
[DHR is addressed in Section 2.1.6, guideline #4 of NUREG-e 1335.]
FE-10. What are the size definitions (equivalent hole diameters) for the LOCA categories, including the distinction between water and steam line breaks? [ Initiating events are discussed in Section 2.1.3, guideline
- 1 of NUREG-1335.]
FE-11. Were containment bypass sequences considered in the Master Logic Diagram?
[Section 2.2 of NUREG-1335 discusses containment considerations.] Please explain.
HF-1.
Describe the HRA "in-house" review and per NUREG-1335, the review team members that participated in the HRA review portions of the IPE.
HF-2.
It is not clear why operator actions were evaluated for inclusion or exclusion in the plant and system models.
For example, containment flooding was identified by the independent review as an E0P-required actic, for " bottom breaks" in the comments Section (see p. D-6 in submittal.)
The resolution to this comment implied that flooding "is not required to establish or maintain stable shutdown conditions,....
and is not expected to significantly increase the likelihood of plant damage." However, the operator is instructed to take action by the E0Ps; an action excluded from the analysis. Other IPEs have found that containment flooding can cause early containment failure if molten core breaches the RPV with the RPV with the torus flooded (loss of pressure suppression).
Therefore, the analysis of this operator action helped these plants to identify an area of concern in their E0Ps. Wac the potential down-side of flooding containment identified?
If so, what is the rationale for screening out human actions associated with flooding containment?
From this example, please explain what criteria was used to assure that important operator actions were not excluded from the analysis.
HF-3.
It is a common PRA practice to include pre-initiating events as part of the HRA.
Experience shows that this is an important aspect of the HRA because it provides the opportunity to examine and scrutinize plant maintenance and calibration practices and identify additional areas for potential improvements (Generic Letter 88-20, Section 2, examination process, page 2). The IPE states that the frequency of actions occurring prior to an initiating event are captured in the basic equipment failure rates (p. 6-1) and consequently were not modeled explicitly (p. 5.4.3). However, the NUREG-1150 experience (Peach Bottom) demonstrated that the analysis would have missed significant risk reduction events as well as significant risk increase events (summarized in NUREG/CR-4450, Vol. 4, pgs.1-1 & 4) had they not performed a detailed pre-initiating event analysis. The licensee is requested to assure that important events were not overlooked by omitting a detailed modeling of the pre-initiator human errors.
HF-4.
Per NUREG-1335, "...unless proper justification is provided, all important recovery actions should have written procedures" (Appendix C, Section 9, Response to question 9.1, pg. C-19). The licensee, however, included "non-procedure guided actions." Such an approach has the positive effect of identifying all possible ways of-recovering safety system functions during an accident which is one of the objectives of the IPE. On the negative side, however, modeling many operator recovery actions in a-PRA has the effect of driving the core damage frequency to very low estimates. Unless a sensitivity and an uncertainty analysis is performed, such a method may mask the true plant capability to cope with severe accidents.
a i,
Discuss the significance of non-procedure guided actions modeled in
.f the IPE to the total core damage frequency. Also discuss the measures taken (or available) to assure consistency of the.IPE with the' available procedures and operator training for these actions.-
t BE-1 Please provide a discussion of key uncertainties (phenomenological, systems, and operator actions) and their potential impact on the i
release categories.
BE-2.
Please provide the value for the conditional probability that the containment is not-isolated and a discussion of the process used to estimate its value.
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BE-3.
The IPE does not account for recovery of electrical power following 1
core damage.
Please discuss how the containment failure characterization would change if power were restored.
Identify and discuss any negative effects of restoring electric power.
j BE-4.
What does "such scenarios" refer to in the middle of page 1-4 of.the.
I Level 2 PRA portion of the submittal report?
BE-5.
On page 8-2, first line, Level 2 PRA position of the submittal report, l
it is noted that the PDSs described by "Y" and "Z" (reactor building-not isolated) are not represented.
Please explain why this is the, case.
i BE-6.
Is any credit.taken for " passive mitigation" in the_ reactor building?
It appears.that~ all relevant dominant. release categories have a'"B" designator (note page 11-3 of the' Level 2 PRA), which indicates no credit.
Please explain.
BE-7.
Is the title of Table 12-12, page 12-16, of. the Level 2 PRA portion of
[
the submittal report correct? The tabl.e states that it is;for early--
1 containment failure but it seems to be for late containment failure.
BE-8.
You state (page 4-2 of Submittal Report) that "...the' "no water to l
core debris" end state contributes 8.75% to the total calculated core _
i damage frequency." Further; you suggest that drywell sprays could potentially eliminate at least part of this contribution. How do -
drywell sprays contribute to reducing'the core damage frequency?
BE-9.
It is not clear how the information developed in theilevel _l '(front end) section of the ' report is used in.the-Level.. 2 -(back end) portion.
of the' report.
Please provide the basis for the quantification:of-the CETs for each of the?seven KPDS, and the split fractions used.
j i
BE-10. On page 10-12 of the. Level 2 PRA, you note that procedure-activated dirty venting via the drywell is activated as part of the l
representative sequence for-the KPDS OJAU. The 0JAU KPDS results in a-containment bypass 100% of the' time. What is the source term i
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l associated with this KPDS and what affect does the venting have on the magnitude and timing of the. source term?
q BE-11. On page 7-4, the report states that about half of the CDF is allocated to "No Vessel: Breach," resulting in no radionuclide' releases.. Vessel breach (after core damage) is only prevented by either introducing i
fire protection water when the vessel is under low pressure or-providing' sufficient " control rod drive hydraulic system" flow when j
the vessel is under high pressure.
For both' vessel injection modes,-
operator action is r'auired. Thus, this operator-controlled cooling function appears to have a key bearing on the radiological release profile for Dyster Creek.
Please discuss the sources of the uncertainties associated with core
~
cooling by these means after core damage and the' implications of. these.
uncertainties for the radionuclide release characterization.
Onpage4-2,thereportstates.that"MAAPwasnotlusedtoinvestigate in-vessel recovery under damage core conditions;... where RETRAN/RELAP5 results were mainly used where needed." This' implies -
that the "No Vessel Breach" is related to those sequences where-
' RETRAN/RELAP5 analysis showed that there was adequate core cooling to prevent fuel relocation. This is different from the information'on page 7-4.
Please provide a discussion which clarifies the' two "No.
i Vessel Breach" approaches.
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