ML17256B101
| ML17256B101 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/21/1982 |
| From: | Zudans Z FRANKLIN INSTITUTE |
| To: | Russell W Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17256B100 | List: |
| References | |
| RTR-NUREG-0821, RTR-NUREG-821 NUDOCS 8207130330 | |
| Download: ML17256B101 (7) | |
Text
PP Franklin Research Center A Division of The FranMin In&tute Z. ZUDANS, PH.D.
Senior Vice President end ChiefopereVnti OAi'cer June. 21, 1982 Mr.
W. Russell SEP Project Manager/Technical Coordinaror U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Subject:
Revie~ of SEP Integrated Plant Safety Assessment Report
Reference:
NUREG-0821, Integrated Plant Safety Assessment, Systematic Evaluation Program R.E. Ginna Nuclear Power Plant, Rochester Gas and Electric Corporation, Docket No. 50-244
Dear Hr. Russell:
In accordance with your request I have reviewed the Ginna SEP program and offer the following comments.
At present, the SEP program appears to be well organized and well managed.
The referenced document summarizing Ginna SEP is comprehensive with respect to the arguments leading to'esolution of various SEP topics for Ginna.
The procedure followed during Ginna SEP review, Figures 1 to 3, is the same as the one used for Palisades.
As it can be seen from Figures 1 to 3, the procedure is generally well defined and.at the completion should lead to the satisfaction of the Coauaission's goals ror the SEP program.
Following this procedure, Figure 1, 24 topics exit at (1) identified as generic items related to USI and THI Action Plan, 21 topics exit at (2) because these are not applicable to Ginna and 92 topics reach (3) where the actual review of SEP topics for Ginna begins.
For each of 92 topics, a Safety Evaluation Report was issued documenting the comparison with current 1'censing criteria and identifying areas of potential backfitting.'oth Method 1 and Method 2, Figure 2, were utilized in this process.
However, at the step A (Disposition of Topics), all but 27 were left for backfit candidacy, the remaining 58 having been put in one of the categories 1 to 3..
None of the topics fell in the category 4 (i.e., safety significa'nt departure),
requiring prompt action.
I find that tecnnical arguments leading to distr"'bution of topics to.various categories have sound engineering base.*
t"Topic disposition categories used by Staff differ slightly from those shown under A, Figure 2.
Categories 1 and 2 of Figure 2 are included in Staff's category 1 and categories 4 and 5 of Figure 2 are contained 'n Staff's category 3.
Staff's category 2 is identical to category 3 in'Figure 2.
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W. Russell USNRC June 21, 1982 With respect to topics in the group of Integiated Assessment (Sections 4.1 to 4.31 of NUREG-0821),
NUREG-0821 represents the Draft Integrated Assessment Repozt (DIAR).
Since Licensee response and resolution for most Integrated Assessment Topics is already contained in NUREG-0821, it also represents
'Final Integrated Assessment Report (FIAR).
My overall impression oz Ginna SEP review is that a considerably amount of sound engineering effort has been put in SEP review, in particular, in ter~s of proper understanding of design, processes and consequences involved.
The
- process, ho~ever,.is not complete until all open items. are resolved in an inte-grated manner.
The supplement to NUREG-0821 will discuss the 'Bfl and USI integration.
With respect to the specific topics reviewed in NUREG-0821, I offer the following additional comments.
~ For 9 of 27 topics slated for Integrated Assessment, risk assessment by Sandia (SAI) (usixg REP Crystal River-3 PRA and HASH 1400 PNL Combination) provided logical support for engineering judgment in complicated situations.
In this context V-5 (RCP Boundary Leak Detection) was identified as high risk item (two diverse systems are required to identify 1 gpm leak within one. hour),
VIII-3B (DC Power Supply System and voltage monit'Gring) as medium importance to risk (justifying recommendation to install battery current meters for resolu-tion of VIII-3B).
Similarly, an extensive use of the plant operating e~erience in support of engineering judgment was made by the SEP staff.
In this context, it was established that frequent problems affecting the plant availability were mostly the result of equipment malfunction or failure, human error was an infrequent contributor.
Of '83 LERs, however, 58/ were either caused or complicated by human error.
Relative to Topic VI-7.B (Eng'neered Safety Feature Switchover from Injection to Recirculation Mode), licensee has agreed to install redundant RWST level transmitter and indicator.
Action also has been initiated to expedite the switchover procedure.
Current Ginna procedures,. if followed correctly, do not provide sufficient margin in switchover time available.
Westinghouse (under contract to licensee) is to review procedures (in terms of uncertainties asso-ciated with boil off calculations) and to implement changes in coordination with T."iZ Action Plan Item I.C.1.
I~find this topic has been properly analyzed'and proposed solution represents an improvement.
Integrated assessment,'ection 4', addresses 27 topics (foui of which
-III-5.B, III-6, VI-7.B, VI-4 and IX-6, were commented upon above).
Summary
~'f the inte'grated assessment is shown in Table 4.1 of NUREG-0821.
The presentation of the results is well organized and easy to follow.
Hr.
W. Russell USNRC June 21, 1982 An integrated three step ef ort to resolve structural design related issues (II-2.A, ZZI-2, III-4.A, 111-7.B, iZI-6, ZZI-5.A, ZZI-5.B and ZX-6) proposed by the licensee is well defined and includes important aspects such as design parameters and acceptance criteria for structural upgrade as well as value-impact assessment, followed by performance of structuzal analysis and engineering desi'gn and installation.
In view of the limited stzuctural capability of a number 'of structures important to safety (as illustzated by the wind load capability, Table 4,2,
'.lUREG-0821), major structural mod'fications may be re-quirec if current loads are imposed (250 mph tornado winds, 1.5 ps'g different'al pressure).
It is of interest to see what specifications and acceptance cziter'a licensee proposes (due Dec.
1982) to resolve these topics.
Topic iII-5.A (Effects of Pipe Break on Structures,
- Systems, and Components Inside Containment) al'ows fracture mechanics evaluation in lieu oz remedial modification in case of impracticality of the latter.
I believe that if the affected components are required to mitigate the conseouences of the pipe break (such as foz example pressurizer surge line break) physical protection should be provided.
With respect to containment liner insulation and the thermal compressive stress for non-insulated portion of the liner (topic III-7.B), it should be noted that liner buckling is not synonymous with liner failure..
What one 'should guard against is ailure of liner anchorages adjacent to a buckled liner panel.
Removal of the dome insulation above the spring line will not help here.
Common issue resulting in inudation of auxilxary in turbine buildings and
.(
screenhouse is addressed in topics II-3.B, II-3.B.1, and II-3.C.
Stafz's estimated maximum flood level of 275.4 ft. was based on probable maximum, flood flow of 38,700 cfs.
Using standard project Plood Determination Procedure of the U.S.
Army Corps of Engineers, flood peak discharge was estimated at 15,000 czs.
The largest recorded normalized peak discharge flow was approxi-
= rely 1/3 oz the capacity of Dee" Creek to convey water without over low.'ng the ba."'-: (~12,000 c=s). it is difficult for this rev'ewer to get overly concerned about the va'e of additional rotection recuired by sta
= beyond the standard Pro'ect.Plood levels unless the probability oz occurrence of sucn flooding is stated and it exceeds an acceptable value.
Classification of structures, components, and systems (topic IIZ-1) iden-t'fies a number oz issues resoluton or which depencs on additional information to be supplied by the licensee.
The radiography or volumetric inspection of welded joints 'n safety related components and piping is important and should be done (iz not performed previou'sly).
Modifications relative to tendon suzvei'lance program (topic iIZ-7.A)'re important.
The tendon relaxation experience of Ginna requi".es explanation and reasons for it must be understood (not discussed in NUREG-0821).
Mr.
W. Russell USNRC June 21, 3.982 Reactor primary coolant system inventory is addressed in Topic V-5.
Although the resolution requires no backfitting, the procedures for RCPB inventory must be able to provide operator with good data on the state of the system.
Resolution of TMI Action Plan Items II.F.X,'I.F.2, and II.F.3 should provide the required confidence in determination of RCPB inventory during normal operation and during accidents.
Staff's-requirement to have overpressurization protection system (OPS) in service prior to RHR operat'on is proper.
Question in my mind relates to relief valve capability to limit the pressure in the RHR system below the failure pressure even with OPS operational.
USI A-17 (Systems Interactions in Nuclear Power Plants) should address this issue in greater detail.
Topic VIII-3.B (DC Power System Bus Voltage Monitoring and Annunciation) resolution requires additional indications so that the operator is informed on the functional status of the DC power system.
Qualification level of this back-fitting item is not defined in NUREG-0821.
Topic IX-3.(Station Service and Cooling Water System) addresses an important
~link between the safety related components and the ultimate heat sink.
Tt is this reviewer's opinion that the importance of heat removal systems can never be overstressed and that the availability of an adequate service water must not depend on some complex procedure to be followed.
In spite of low impact found by PRA, Staff's requirement to install additional water level indications in CCW surge tank is proper.
Very truly yours, ces enons Zudans enior Vice President and Chief Operating Officer
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