Text

Provides Response to Request for Addl Info Re Implementation of ASME Code N560 at Plant
	ML20236X182
Person / Time
Site:	Vermont Yankee
Issue date:	07/31/1998
From:	Leach D; VERMONT YANKEE NUCLEAR POWER CORP.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	BVY-98-114, NUDOCS 9808070182
	Download: ML20236X182 (60)
	v • d • e

. _

VERMONT YANKEE i

~

<G p NUCLEAR POWER CORPORATION 185 Old Ferry Road, Brattleboro, VT 05301 7002 (802) 257-5271 i

July 31,1998 BVY 98-114 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington,DC 20555

References:

(a) Letter, USNRC to VYNPC," Request for Additional Information Regarding Implementation of ASME Code Case N560 at Vermont Yankee Nuclear Power Station (TAC NO. M99389)," NVY 98-32, dated March 11,1998 (b) Letter, VYNPC to USNRC, " Implementation of ASME Code Case N560 at Vermont Yankee Nuclear Power Station", BVY 97-99, dated August 6, 1997

Subject:

Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)

Response to Request for AdditionalInformation Regarding Implementation of ASME Code Case N560 at Vermont Yankee Nuclear Power Station Reference (a) requested Vermont Yankee Nuclear Power Corporation (VYNPC) to provis.

additional information in response to questions outlined in the enclosure to that letter. Attached for your use is a written reply to each of the staffs questions.

We trust that the information provided will enable you to complete your review of Reference (b);

however, should you have any questions on this matter, please contact this office.

Sincerely, VERMONT YANZE NUCLEAR POWER CORPORATION *

~

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,, Don M. llach ' }J s

1,v u' Vice President, Engineering Attachments cc: USNRC Region I Administrator USNRC Resident Inspector- VYNPS USNRC Project Manager-VYNPS Vennont Department of Public Service 9808070182 990731 PDR ADOCK 05000271 p PDR

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TABLE OF CONTENTS RA I Q U E S TI O N 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . .. . . . . . . . . . 2 RA I Q U E S Tl O N 2 . . . . . . . . . . .. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . .. . . . . . . . . . . . . . . 6 RA I Q U E S TI O N 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .. . . . . . . . . .. . . . . . . . . . . . . . . . 7 RA I Q U E S TI O N 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . .. . . . . . . . . . . . . . . 8 RA I Q U E S TI O N 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 RA I Q U E S TI O N 6 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2 RA I Q U E S Tl O N 7 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . 1 4 RA I Q U E S TI O N 8 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . 1 5 RA I Q U E S TI O N 9 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . 1 7 RA I Q U E S TIO N 10 . ......... . . .... . . . .. . ......... .. .. ............. ....... .. .. ... .. ..... . ... .... . ...... .. ... .. 17 RA I Q U E S Tl O N 1 1 .... .. .. . .. . . . .. . .. .. . . .. .. .. .. . . .. . . . ....... .. . . . .. .. ... . .... . . . . .. . . . ............. .. .. . .. 19 RA I Q U E STIO N 12 . . . . . .. . .. ... . . . . .... . ..... . . .. . .. . . . . . .... ... . . . . ..... .. . . .. ... . ... . . ....... .... ... . .... . . 19 RAI Q U E STIO N 13 . . . ..... ... . . . . ..... . .. .. .. .. .. . ... .... . .. . . . . . . . . ..... . . . . .. .... .. . ... . .. ... .... . . . .. .. . .. 21 i

RAI Q U E STIO N 14 . . ...... . . . .... . .. . . . . . . . . .. .. .. ... . ..... . . . ... .. . .. . .. .. . . . . .. . .... . .. . . . ... . . . . . . . . . . . . . . 21 RA I Q U E S TIO N 1 5 ... .... ... . .. . . . . . ... .... ... ... .. . . . .. . . . . . . . . . . . .. . . . . ..... . ... . . . . . . . . . . . . . . . ... . .. .. .... 2 2 l RA I Q U E S TIO N 16 .. . . . . . .... ... ... .... . ... . ... . ..... ...... . . . .. .. . . . . . . .. . . .. . .. . . . . . . . . .. . . ..... . . . . . . . . . .. 2 6 I

f 1

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l

REOUEST FOR ADDITIONAL INFORMATION The following information was requested by the staffin order to allow them to complete their review.

Ouestion 1 - RAI G/11/98)

N-560 requires an evaluation ofindirect effects of postulated pressure boundary failures, including the spatial effects of flood, spray and pipe whip on mitigating systems, and these effects have generally been addressed by the licensee. However, for breaks outside the Drywell, the efTect of a steam environment on Motor Control Centers (MCCs) and other electrical components should be evaluated. Show that the analysis of the result of breaks outside containment has considered the effect of a steam environment on MCCs or other electrical switchgear.

Answer Steam impact on MCCs and other electrical components in the reactor building is based on the

~ VY IPE analysis of LOCAs outside the Drywell (IPE Section 3.1.3). The only piping outside the drywell that does not have a HIGH consequence rank is steam piping in the steam tunnel (main steam, HPCI, RCIC, and main steam drains). Thus, this piping, which has a MEDIUM

- consequence, is described further below. The following summarizes the analysis of steam pipirs in the steam tunnel for the failure to isolate case which is judged to provide the more severe '

envimnmental conditions:

e LOCAs in the steam tunnel are assumed to disable HPCI and RCIC (e.g. a large break will depressurize the RPV causing loss of motive power to the HPCI/RCIC turbine),

feedwater (e.g. steam impact in the turbine building), and the main condenser (e.g. ,

' MSIV isolation due to high steam tunnel temperature). Only low pressure makeup is credited in the analysis.

LOCAs in the steam tunnel will propagate into the turbine building via blowout panels. Open areas, a door and smaller blowout panels, also allow propagation into the reactor building Elevation 252.

The environmental impacts on electrical equipment in the reactor building are assumed to be minor because: 1) equipment is not in the immediate, direct propagation path, and 2) Elevation 252 of the reactor building is large, and vents directly to higher elevations through an equipment hatch and stairs. Overpressure in the reactor building is relieved by blowout panels on the refueling floor establishing a 1- flowpath to the atmosphere. Low pressure ECCS injection is allowed to be successful based on the judgment that severe environmental conditions are unlikely to impact 3 equipment in the reactor building. Note that for the breaks in the steam tunnel, j depressurization is assumed to be successful.

2 i

I m____________ ____m___ . - _ _ _ _

The largest of the steam piping in the steam tunnel is main steam (18 inch).

l The inside drywell MSIV closure reliability is not affected by the steam environment. The potential for impact on low pressure ECCS due to steam impact on electrical equipment is judged unlikely.

Table 1 summarizes major low pressure ECCS equipment (high pressure ECCS equipment I is not credited for those breaks), its location, electrical dependencies, IPE evaluation summary for pipe breaks in the steam tunnel, and equipment (MCC) qualification relative to maximum temperatures expected from HPCI and main steam (MS) pipe breaks. As shown, HPCI steam breaks are more limiting. MS piping isolates immediately on high temperature without delay, whereas HPCI has a minimum 30 minute delay for smaller breaks, and the corresponding temperatures are judged more representative oflarge breaks with failure to isolate. For the MS failure to isolate case, temperatures in the reactor building Elevation 280 are notjudged to increase above the 200F qualification temperature.

Therefore, as described above and in Table 1, the N560 evaluation has addressed the environmental effects of breaks outside of the drywell.

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Ouestion 2 - RAI G/11/98)

The example in Section 2.2 for the Initiating Event Impact Group Assessment includes l the probability of failure to isolate a break in the reactor water cleanup system due to

failure of the inner isolation valve to close on low reactor pressure vessel (RPV) water level (4E-3), but Note 3 of Table 2-1 indicates that the initiating event frequency of loss-of-coolant acci&nts (LOCAs) outside containment already includes the probability of inner isolation valve failure. This would change the ranking based on conditional core damage probability (CCDP) to high rather than medium. Provide further clarification and an evaluation to determine any changes in the ranking of LOCA-OC events.

For this example (and the' analysis that considered isolation success), provide an evaluation showing that the reactor water clean-up (RWCU) motor-operated valve (MOV) would be able to close in the event of blowdown due to the break.

- Answer The CCDP shown in Table 2-1 for LOCA-OC, versus the final CCDP estimate as part of the RI-ISI analysis documented in Table 4-1, is clarified with the folfowing example of how LOCA-OC was included in the VY PRA:

CDFm.oc = IEFm.oc

CCDPm.oc = [Pr

ISO]

CCDPm.oc where: CDFm.oc = Core Damage Frequency due to LOCA-OC initiator ;

IEFm.oc = Initiating Event Frequency for LOCA-OC (given in Table 2.1) -

Pr = pipe failure frequency ISO = failure of an inner isolation valve (MOV) to close on demand CCDPm.oc = unavailability of mitigation capability given LOCA-OC initiator (given in Table 2-1)

Because of the nature of the evaluation (analysis of the potential pipe breaks) the RI-ISI analysis assumes the pipe failure (P,) as the initiating event and includes an inner isolation valve failure in the plant response. In this case, the following summarizes the equation:

CDFm.oc = P, * [ ISO

CCDPm.oc] = P, * [ Final CCDPma.]

The RI-ISI analysis assumes a pipe break, Pr =1.0, and includes an inner isolation valve

- failure in the final CCDP estimate for the case where impacts of MOV failure are assessed (Table 4-1). Thus, for the RI-ISI assessment the actual CCDP estimate includes the valve failure probability and the CCDPm.ac from Table 2-1. The following also I summarizes above discussion (also see Section 2.2):

6

Figure 1 P, MOV Impact CCDP Total pipe break Closes Table 2-1 Table 2-1 CCDP 1.0 1.0 T <1 E-6 <1 E-6 4.0E-03 LOCA-OC 1.0E-02 4.0E-05 1

Detection and isolation of RWCU pipe breaks outside the Drywell is a design requirement to protect electrical equipment from a RWCU line break (FSAR Page 4.9-4, Revision 14). Based on calculation VYS-1459 (Revision 1)" Determination of 1

~ Blowdown Isolation Valve Stroke Times for Generic Letter 89-10", RWCU isolation MOV15 blowdown conditions achieve a 1130 psig line pressure and differential pressure of 1166 psid. The above calculation demonstrates that the valve will close under these ,

conditions.

]

l Ouestion 3 -RAI (3/11/98L Item 2 under the System Impact Group Assessment regarding determination of the number of unaffected backup systems / trains includes a confusing statement: "When considering the consequences, given an isolation failure, the number of available backup trains includes isolation.". A similar statement appears in the Combination Impact Group Assessment discussion. Clarify what is meant by these statements.

Answer Figure 1 in the response to RAI 2 provides an example of how isolation success and failure are assessed when a pipe break causes an initiating event. As shown in the figure:

e isolation success - the probability of successful isolation is included (in this case 0.996 or ~1.0), the impacts associated with success are included (T initiator),

and the associated backup trains (CCDP) given the impacts is included (CCDP,

<1E-6 for T initiator in Table 2-1).

e isolation failum - the probability ofisolation failure is included (in this case, l 4E-3), the impacts associated with the failure to isolate are included (LOCA-OC initiator), and tae associated backup trains (CCDP) given the above impacts is also included (~1E-2 CCDP for RWRBI LOCA-OC in Table 2-1).

l For the case where the pipe break does not cause an initiating event, it is assumed that the l pipe failure occurs during a demand on the piping and the analysis is slightly different l

(System Impact Group Assessment). The following example is provided to clarify the

' analysis (refer to Demand Configuration of Line 24-RIIR-29, the segment between MOV25B & Drywell, in Table 4-1 of the August 1997 submittal).

7

LOCA Operator Exposure impact Backup Equipment Total Demand Isolates Time Trains l CCDP CCDP 1E-2/yr ~1.0 1 yr LPCIB 2.5 1.E-05 1.E-07 1.E-02 1 yr All ECCS 0 1.0 1.E-04 f

. Isole+ ion success - the probability of successful isolation is included (in this case,0.99 or ~1.0). The successful isolation results in the loss of only one train of ECCS (LPCI B). The LOCA demand of LPCI in Table 2-2 is an unexpected frequency of challenge (1E-2), the exposure time is all year, and there are 2.5 backup trains in Table 2-2 (CCDP-1.0E-5).

e Isolation failure - the probability ofisolation failure is included (in this case, 1E-2). When the operators fail to isolate the break before LPCI B pumps the suppression pool into the Reactor Building, the impacts are assumed to be much more significant (loss of all ECCS). The LOCA demand of LPCI in Table 2-2 is an unexpected frequency of challenge (1E-2), the exposure time is all year, and there are no backup trains (CCDP = 1) except for the probability of the operator to fail to isolate (in this case,1E-2).-

For both cases (i.e. this pipe break causes an initiating event or does not), the likelihood ;

of successful isolation and unsuccessful isolation is assessed. Automatic isolation is generally more reliable than remote manual isolation. Manual isolation may only be credited when there is sufficient time, detection and a practical means ofisolation.

l Ouestion 4 - RAI (3/11/98)

Table 2-2 of the submittal, which was used as a guideline for assigning consequence categories to pipe failures that result in a loss of systems / trains without an initiating event, differs from the version in N-560 (Table 1-5) and the one in the EPRI methodology. In some cases, using the consequence categories of Table 2-2 would result in less conservative rankings than using the corresponding table in N-560 (e.g., for a system responding to an infrequent event _with a long AOT and only I backup train, the l

consequence category would be high in the N-560 table but medium in Table 2-2).

Although the basis for Table 2-2 is explained in RAI Response III.8, this would seem to be a departure from the original guidance. Please explain why it was necessary to depart

- from the original guidance.

i l

8 l l

l t _ - --- _. _ _ - - _ _ _ - _

I

a. With regard to your response to RAI III.8 (10/23/97), explain how the unavailability limits were chosen for each number of backup trains available, and whether the resulting mean values that were used are conservative with respect to values ;

obtained from the VY PRA as shown in Table 3-2 of the submittal. l Answer In Table I-5 (N560), the exposure times defined as "long" and "short" correspond to the "all year" and "between test" exposure times in Table 2-2.

Table I-5 Table 2-2 Exposure Time Long (1 year) All Year (1 year) i Short (1 month) Between Tests ( l to 3 months)

Table I-5 does not use "long and short AOT" exposure times, because it is not expected that those would apply in the N560 evaluation (this is a conservative assumption because, in some cases, the break could be detected before a demand and an ACT would be entered). Basically, detection of the break, before a demand, is not credited in the N560 Code Case. Therefore, for the infrequent demand, and one backup train:

Table I 5 Table 2-2 Long H H All Year Exposure Exposure Short M H Between Test Exposure Exposure This conservative difference is because in Table I-5 "Short Exposure Time" corresponds to one-month test periods, and in Table 2-2 "Between Test Exposure Time" corresponds to three-month test periods.

a. Unavailability limits given in response to RAI III.8 (10/23/97), are chosen so that the geometric mean of the upper and lower limit corresponds approximately to the mean value (in a lognormal distribution - the median value is a geometric mean of symmetrical bounds). Therefore, for example, limits [3E-2,3E-1] correspond to:

I

/3E-02+3E-Ol = lE-0 l 9

To show how the resulting mean value compares to the values in Table 3-2, Table 1 is given below:

Table 2: Actual System Unavailabilities vs. Credited Backup Unavailabilities Credited Credited Actual l

System Backup Backup System Trains Unavailability Unavailability CR-control rod insertion 2.5 1E-5 1.0E-5 VS-vapor suppression 2 1E-4 1.1E-4 LP-LPCI(all support available) 1.5 IE-3 2.5E-4 LP-LPCI(1 LPCI loop) 1 1E-2 9.2E-3 CS-CS (all support available) 1.5 1E-3 3.9E-4 CS-CS (1CS loop) 1 1E-2 1.1E-2 PI-ECCS LP interlock 1.5 1E-3 4.2E-4 TC-torus cooling (2 train RHR) 1.5 1E-3 5.2E-4 TC-torus cooling (1 train IUIR) 1 1E-2 1.0E-2 VT-containment vent 1.5 lE-3 1.9E-3 Al-alternate injection (2 trains) 2 1E-4 9.9E-5 Al-alternate injection (1 train) 1 1E-2 1.0E-2 DS-Drywell spray % 1E-1 1.0E-1 OD-depressurization (MLOCA) % 1E-1 9.9E-2 OD-depressurization (SLOCA) 1 1E-2 4.7E-3 lip-HPCI (all support available) % 1 E-1 8.8E-2 RC-RCIC(all support available) % 1 E-l 1.1 E-1 HP & RC -HPCI and RCIC 1 1E-2 1.3 E-2 FW-feedwater 1 1E-2 1.5 E-2 AD-ADS (MLOCA) 1.5 1E-3 1.1 E-3 AD-ADS (SLOCA/ Transient) 1.5 1E-3 3.6E-4 CN-condensate 1 1E-2 1.3 E-2 RM-recover main condenser 2 1E-4 1.1E-4 TB-turbine bypass (transients) 1 1E-2 5.3 E-3 As can be seen from Table 2, there is agreement between credited unavailabilities for VY mitigating systems and actual system unavailabilities. This is to be expected, since the number of credited backup trains is determined based on the actual system unavailability, as described in the response to RAI III.8.

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Ouestion 5-RAI (3/11/98)

Table 2-2 is not based on CCDP from the PRA calculations, but instead relies on estimates based in part on Unavailability Mean Values corresponding to the number of backup trains. It is not clear how the quantitative estimates account for dependent failures between trains of the same system or different systems. Please clarify how common cause failures and spatial impacts (if any) are accounted for. Please also clarify how support system dependencies are addressed. For example, Figure 3-1, Note 2 (for the medium LOCA success criteria) indicates use of the condensate system as a low pressure makeup source is dependent upon feedwater success. Similarly, residual heat removal (RHR) train "A" for decay heat removal is dependent upon LPCI train "A".

Please explain how dependencies are treated while generating the quantitative estimates and provide some example calculations.

Answer The quantitative estimates take into consideration dependent failures between two trains, by using PRA values for the corresponding system. For example, in Table 3-2, LP injections trains (1 LPCI or 1 CS Loop) are each credited as one train. Two trains of LPCI or CS are not credited as two trains, but as 1.5 trains. This is based on the actual system unavailability value, which acccunts for dependency and common cause between trains.

Spatial effects due to the pipe break are explicitly analyzed for each case, and effects on multiple systems are evaluated.

Support system dependency and common cause among different systems are considered only ifinfluenced by the pipe break. Otherwise, it is considered that they will have a minor effect on the determination of the number of citigating trains. Support system dependencies could have a significant effect if a suppod system (s) is analyzed or impacted by the pipe break and, in this case, support system dependencies would be evaluated. No such cases were identified as part of the N560 evaluation. Also, it should be noted that the loss of a major support system almost always leads to an initiating event, and the corresponding CCDP would account for these dependencies.

Direct dependency between systems is always considered. For example, Condensate dependency on FW would be evaluated for all pipe breaks that would effect FW, if Condensate is going to be credited as a backup train in the analysis (Note: Condensate was not credited as a LP makeup system in this analysis) .

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The same is valid for RHR and LPCI dependency. If the pipe break could disable both systems, then the " worst case" consequence is analyzed. For example, the break which disables LPCI will impact the " Low Pressure Makeup" function. Based on Figure 3-1, the remaining backup trains are two CS trains, credited as 1.5 trains (Table 3-2). Based on Table 2-2, for " unexpected frequency of challenge" (<102 /yr), "all year" exposure time (1 year), and 1.5 backup trains (10-8), the consequence category would be

_ (<10-2/yr

1 year

10-3 < l.0E-5) in " Medium." If the same break also disables both !

RHR trains, it will impact the " Heat Removal Function" and " Shutdown Cooling." For the " Heat Removal Function," based on Figure 3-1, the remaining backup train is Containment Vent with Alternate Injection, credited as 1.5 trains (Table 3-2). Based on Table 2-2, with the same input as in the previous evaluation, the consequence category is l also " Medium." Both the " Low Pressure Makeup" and " Heat Removal" functions result in the same consequence rank, and a " Medium" rank would be assigned to the analyzed segment (before considering shutdown cooling).

In order to check the use of Figure 3-1 and Table 2-2, two PRA runs were performed using the VY IPE PRA model. For example, for the loss of one LPCI train, CCDP

[CCDP (LPCI failed)- CCDP (base case)) was quantified to be 1.1E-6, which is in agreement with the Table 2-2 evaluation.

Out of 65 consequence segments, Table 2-2 was used to evaluate only 12. Most of those segments ('7) have already been evaluated as having a "High" consequence, due to other considerations (mainly LOCAs outside containment). Four of the five remaining consequence segments are evaluated as having a " Low" consequence. The pipe break in those segments would result in a loss of one CS or LPCI train; the break is inside the Drywell, upstream from the check valve (no LOCAs). The last consequence segment from this group is in the SLC System. As can be concluded from the above summary, the dependency between different mitigating systems does not play an important role in the N560 Code Case evaluation.

Question 6-RAI(3/11/98)

Several questions remain regarding determination of the number of backup trains based on the critical safety functions success criteria as depicted in Figure 3-1 of the submittal:

a. Please clarify the relationship between the " Qualitative Basis" in Table 4-1 and the success criteria in Figure 3-1. Please verify that the success criteria in Figure 3-1 reficct all the entries in 4-1, or provide a set of figures similar to Figure 3-1 with the appropriate success criteria.

i b. Note 3 of Figure 3-1 indicates that the VY PRA does not credit the recovery of the main condenser for MLOCA. Identify why recovery of the main condenser is credited as a backup in this figure.

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[ c. Success criteria for a small LOCA indicate reactor core isolation cooling (RCIC) is sufficient backup for high pressure makeup. Most BWR studies consider that there are some small LOCAs that RCIC alor mould be insufficient to make up. Clarify why this success criteria conflicts with the discussion for LOCAs in Section 4.2 of the submittal since, in that section, "small LOCA" is defined as the pipe size that, if broken, still allows normal makeup capabilities from 1 CRD pump and RCIC to maintain reactor water level.

Answer

a. Figure 3-1 is used primarily to communicate key success criteria. When a pipe break causes only an initiating event (Initiating Event Impact Group Assessment), the CCDP in Table 2-1 provides the consequence rank and by definition accounts for the success criteria communicated in Figure 3-1. The use of Figure 3-1 for the System Impact Group and Combination Group Assessments is summarized below:

System Impact Group Assessment (Table 2-2) - the number of backup trains are determined using Figure 3-1 and Table 3-2. For example, assuming that the pipe failure impacts the low pressure makeup function (e.g., core spray), the remaining backup low pressure makeup function trains are estimated from Figure 3-1 and Table 3-2. Also, a number of PRA quantifications were conducted to verify the results.

Combination Impact Group Assessment - the number of backup trains are determined using Figure 3-1 and Table 3-2. For example, assuming that the pipe failure impacts the high pressure makeup function (e.g., RCIC), the remaining backup high pressure makeup function trains are estimated from Figure 3-1 and Table 3-2. Also, the PRA was quantified to verify the results.

b. Recovery of main condenser (heat removal function) for MLOCA could be removed from Figure 3-1. It was shown for completeness and is not credited in the assessment of MLOCAs in Table 4-1.

c. The calculation that was conducted to determine ASME SXI line size exemptions was used to define the upper limit for small LOCA (SLOCA). The success criteria in this calculation is based on defining the break sizes for water and steam pipe breaks which is equivalent to 450 gpm makeup capacity, i.e. 400 gpm from RCIC and 50 gpm from CRD. Thus, these systems have the capability to provide makeup for the break without loss oflevel and subsequent ECCS actuation. In the VY PRA, the lower MLOCA limit (also the upper SLOCA limit) is defined as that break for which RCIC alone can not keep the core covered and prevent core damage. These two definitions for the SLOCA upper limit do cause some confusion. The PRA success criteria is l judged to be reasonable because it neglects CRD and the success criteria is preventing core damage versus maintaining break flow. Also, VY's response to NRC questions 13

on the IPE (Letter BVY 95-114, dated October 27,1995) indicato that RCIC success criteria is consistent with NUREG/CR-4550.

Question 7- RAI (3/11/98)

In Table 4-1, it is not clear how the corresponding CCDP estimate for the Table 2-3 qualitative basis evaluations was obtained from Table 3.1 and 3.2. Please describe how this analysis is done and provide some example calculations.

Answer In Table 4-1, when Table 2-3 would apply for the consequence analysis (Initiating Event with Mitigating System Impact), actual PRA runs were used (to account for dependencies). Those runs are presented in Table 3-1. Four analyzed cases are given below:

Lost Initiating Event Mitigating CCDP ~ Consequence System (From PRA Run) Rank TFWMS HPCI 4E-5 MEDIUM (MSIV Closure with a Loss of RCIC 5E-5 MEDIUM FW)

TMS HPCI 4E-6 MEDIUM (MSIV Closure with FW RCIC 4E-6 MEDIUM Available) i i

The CCDPs for the above cases are obtained by summing the CCDPs for all safety functions in Table 3-1 (including Reactivity Control). If Table 2-3 were used, the rank q

would be:

TFWMS and Loss Of HPCI(or RCIC):

Backup trains for "High Pressure Makeup" from Fippe 3-1, are RCIC (HPCI) and "Depressurization" (2 of 4 SRVs). From Table 3-2, RC C (HPCI) is credited as 0.5 train l and ADS depressurization (Transient) is credited as 1.5 trains, resulting in a total of 2

- backup trains. Based on Table 2-3, and available backup trains, this would result in a

" Medium" consequence category. The IE Category for TFWMS, from Table 2-1, is also i

" Medium."

TMS and Loss of HPCI (or RCIC):

In this case, FW is also available as a "High Pressure Makeup" mitigating train. Based 1 on Table 3-2, FW is credited as one train, resulting in a total of three backup trains.

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Based on Table 2-3, that would result in a " Low" consequence category. Ilowever, the l IE Category for TMS, from Table 2-1, is " Medium." Therefore, the " Medium" category is assigned to this event.

Since both cases were " boundary" cases, the actual PRA quantification was performed.

As shown above, the CCDP quantified from the PRA resulted in the same consequence rank as the application of Table 2-3. This, once again, confirms that the approach presented in Tables 2-1,2-2, and 2-3 is reasonable and robust.

Ouestion 8-RAI G/11/98)

The discussion on containment performance in Sections 2.2 and 3.2 of the submittal is based upon the EPRI methodology rather than being an N-560 requirement. The analysis in Section 3.2 regarding containment performance for the high pressure makeup function is confusing. Further clarification of the basis for the containment performance analysis is requested.

Answer Code Case N-560 does not provide explicit guidance on containment performance. The guidance and requirements used in the analysis are similar to the EPRI methodology. The requirements established !n this analysis have two elements; either one can lead to an increase in consequence category from that determined by CCDP based on core damage:

1. If the containment is bypassed (e.r , e are .k outside containment and the inside isolation valve failed to cicO '- me CCDP estimate, the limits of the "high," " medium," and " low" consequence ranks are decreased by an order of magnitude. The new limits will be:

HIGH: CCDP (with Containment Bypass) > IE-3 MEDIUM: CCDP (with Containment Bypass) 1E-5 to 1E-7 (range)

LOW: CCDP (with Containment Bypass) < 1E-7 For example, if the CCDP was 2E-5 (Medium consequence) and the containment was bypassed,"High"is used for the consequence category. In the case of breaks inside containment, there is always a barrier with at least a conditional probability of 0.1, which is credited for containment isolation.

2. A combination of early core melt and early failure of the containment structure is also included. This case is approximately evaluated assuming the same limits as above:

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_ - _ _ _ _ . _ - _ . _ . _ _ _ . - _ _ _ _ _ - . _ _ _ . _ - _ - - _ _ _ _ _ _ _ _ _ . _ _ . - - - ~ . _ - . - - - _ _ _ _ _ - .

3 P-LERF ~ CCDP

CLERF HIGH: P-LERF > IE-5 LOW: P-LERF $ 1E-7, ,

where, LERF-Large Early Release Frequency P-LERF-Probability of LERF CLERF-Conditional LERF The intent is to show an order of magnitude margin or require an increase in the consequence category. Since BWRs can have a conditional large early release frequency (CLERF) due to structural failures on the order of 0.1 or higher, specific cases were evaluated. Conditional LERF depends on the safety function that failed. For example, i core damage due to reactivity control failure (ATWS) and vapor suppression failure (LLOCA) has a relatively higher conditional probability oflarge early containment -

structural failure. Section 3.2 reviews taese functions to determine relative importance -

and potential impact on consequence category assignments.

Section 3.2 and Tables 2-1 and 3-1 show that there is also a margin in the initiating event CCDPs for the high pressure makeup function. For example, medium LOCA (MLOCA in Table 2-1)is assigned a " Medium" consequence rank in Table 2-1 with little margin for LERF (CCDP = 8.4E-5), ifit is assumed that the conditional LERF is larger than 0.1. However, as shown in Table 3-1 for the high pressure makeup function, there is clearly an order of magnitude margin even without considering conditional LERF, because CCDP due to HP MU failure is 4.5E-6.

As shown in Table 3-1, TFWMS with RCIC failure does not provide a sufficient margin (CCDP (due to HP MU failure) = 4.5E-5) without assessing conditional LERF; it also has the least margin relative to other scenarios in this table. The total margin relative to the criteria for elevating the consequence from " Medium" to "High" is 0.45 (CCDP/High Limit for CCDP or 4.5E-5/IE 0.45). Both the "Early" conditional probability (0.26) and "Early/High" conditional probability (0.11) of release categories were considered. Based on this assessment, a total conditional probability of LERF is judged to be ~0.1 and the consequence was not increased from " Medium" to "High."

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L__________________-__ --_ . _ _ >

Ouestion 9-RAI(3/11/98) l l You state that the indirect effects of pipe ruptures inside the Drywell are not explicitly l considered since the equipment inside the Drywell is environmentally qualified for the LOCA steam and flooding environment which is within the design basis. This is correct, but the proper operation of the pressure suppression and heat removal systems is required to maintain the design basis conditions. In particular, the Drywell to wet well vent lines, the Drywell Wetwell vacuum breakers, the ADS piping, and the containment atmospheric vents, are to varying degrees relied upon to mitigate LOCAs and maintain conditions within the Drywell within the design basis. Please include an evaluation of welds within these systems. Improved inspection of these items may be useful in offsetting any potential increases in risk due to the reduced number ofinspection of Class I welds.

Answer In our response to the first set of RAls, the application of ASME code case N560 was judged to provide an improvement in plant safety when credit was taken for the more appropriate N560 inspections. ASME code case N560 applies to examination category B-J welds (excluding socket welds), therefore the evaluation of other components is outside the scope of this application.

The subject components of this RAI (i.e. ADS piping, Drywell to Wetwell vent lines,,

Drywell vacuum breakers and containment atmosphere vents) are currently inspected to Class 2 requirements per 10CFR50.55a and as such provide an acceptable level of safety with or without the application of ASME code case N560.

Finally, application of ASME code case N560 is not expected to increase the frequency of LOCA events and as such will not increase the likelihood of challenging the above components.

Question 10- RAI(3/11/98)

Section 3.1 of the Vermont Yankee August 6,1997 submittal, states that the PRA prepared to support the Vermont Yankee Individual Plant Examination (IPE) was used to support the PRA portions of the submittal. Section 2.4 of the VY IPE states that the IPE reflects the plant configuration as of December 1,1993. The stafinotes that the recent maintenance rule program inspection report for Vermont Yankee (Inspection Report Number 50-271/97-81) states that the Core Damage Frequency (CDF) value in the VY

, PRA has increased from the IPE value of 4.2E-6 to a current CDF value of 5.0E-6.

Please verify that all plant configuration and operational changes that have occurred since December 1,1993, have been reviewed and a determination made that no PRA model changes were needed to support the ISI submittal, or that model changes were made as 17 i

needed to support the ISI submittal. Additionally, describe any changes between the PRA prepared for the IPE and the PRA being used to support the ISI program consequence analysis evaluation.

Answer The RI-ISI calculation is based on the original VY IPE. Since the original VYIPE, a number of minor changes have occurred that impact the PRA model and there have been

. minor physical changes to plant design and procedures. The latest procedures and design information for the VY plant were used and referenced in the RI-ISI calculations. There are three parts to this response described below: i

1. Changes to the IPE that were not in the VY IPE: A formal update has not been made to the VYIPE; an update is planned for the end of 1998 (see part 2 below). '

However, there have been sensitivity studies performed as part of the maintenance rule implementation (no changes have been made to the event trees, fault trees, success criteria, and logic rules). There are two primary parts to the sensitivity studies; (1) new unavailabilities are added in the cases where maintenance

- unavailability was not included in the IPE and (2) calculations with all unavailabilities set at their maintenance rule unavailability criteria. In both cases, unavailability was conservatively set to the criteria limit. Item (2) is a conservative sensitivity study to determine the risk significance assuming all equipment unavailability was at the criteria limit. This CDF change was relatively minor and is not applicable to the RI-ISI analysis because the PRA analysis is based on more realistic annual average unavailabilities. Item (1) is partially applicable because some equipment maintenance unavailability will be added to the IPE in a future update. However, the unavailability used in the sensitivity calculation was again set at the criteria limit rather than utilizing actual plant data. Real plant data has not been obtained nor analyzed for input to the IPE, but will be considered in the update. CDF for case (1) resulted in an increase from 4E-6/yr to SE-6/yr, which is notjudged significant. A review of actual plant data (VY Maintenance Rule Program in accordance with 10CFR50.65) indicates that most of this increase in CDF does not apply to the RI-ISI evaluation and that real unavailabilities will have even less impact on CDF.

2. Changes to the plant that have not been incorporated into the IPE model: Plant changes (EDCRs) are reviewed and saved in a book as input to the IPE update.

' This review is also conducted to provide risk input into the change decision process and ensure that significant changes to the IPE results have not occurred. A review of these changes indicates no significant impacts on the IPE, a few potential impmvements (minor), and no negative impacts.

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I 3. Changes to IPE (VYIPE) as part of RI-ISI analysis: The RI-ISI PRA calculations are based on information in the IPE: no change to the IPE was required.

Changes to the PRA computer model were required for computational reasons (e.g.,

to quantify initiating events at 1.0/yr with a low truncation), but again, the calculations are based on the PRA logic structure and data as presented in the IPE with no change required.

' In summary, the present RI-ISI selections are not judged to be impacted.

Question 11 - RAI(3/11/98) i Please clarify whether VY intends to periodically modify its ISI program based on risk insight and some VY internal process, or whether an approved risk informed ISI program will not be changed without prior NRC approval.

Answer The question of periodic updates to the RJ-ISI Section XI program has been discussed at the ASME Section XI Working Group on the Implementation of Risk-Based Examination (WGIRBE), the NEI Task Force on Risk-Informed ISI, as well as between members of the three pilot plants (i.e. Vermont Yankee, ANO-2 and Surry) that have submitted RI-ISI I

applications. N560 currently requires that the evaluation be updated as new information (e.g. risk insights, plant changes, industry information) becomes available. As part of the ASME Section XI revision process, updates to the risk-informed code cases are expected to further address this issue. It is expected that minor changes to the programs (e.g.

documentation, drawing updates) will be incorporated internally in accordance with !

individual plant procedural control requirements. Those changes that significantly impact inservice inspection programs or the basis for NRC's approval of the program (e.g.

response to Generic Letter 88-01, IE Bulletin 88-08) would require significant interaction between the industry and NRC and as such would not be changed without NRC approval.

I Ouestion 12 -RAI(3/11/98)

In your msponse to our Question 11-7, you referred to your Operational QA Program

(YOQAP). Please provide us the documented results of your independent review of the l

analysis in your submittal. This documentation is the documented review required by

! Section III, " Design Control," of your YOQAP (e.g., the review required by Section 6 of ANSI N45.2.11 as endorsed by RG 1.164). Also note that it is the licensee's responsibility to justify that the quality of the PRA is adequate for the proposed application. Attributes of PRA quality _which should be addressed are illustrated in the ,

Draft SRP Chapter 19 issued for public comment on 6/25/97. Processes to support a i licensee claim of adequate quality include structured and documented peer review, cross I

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comparison studies, and industry certification. Please describe the process you have employed (and provide all related documentation) or that you will employ to substantiate the quality of your PRA.

Answer -

There am two issues identified by this RAI. The response to these is as follows:

a. Our response to Question I!-7 of the previous submittal described the calculational process (i.e. analysis development and independent review) used to develop the documents that support the RI-ISI program. This process was carried out under our NRC-approved 10CFR50 Appendix B Quality Assurance Program

- (YOQAP-1-A) in accordance with procedures that provide for adequate verification of the quality of the technical analyses (consisting ofindependent review by personnel technically knowledgeable in the applicable engineering disciplines), and documentation of the performance of the verification. Objective evidence of this verification (signed title pages and review process checklists) is

. included in References 13 and 14, and is excerpted and attached to this submittal.

Copies of the applicable procedures and the referenced RI-ISI calculations are available for staff review at the Bolton, Massachusetts offices of Duke Engineering & Services.

b. As discussed in DG-1061, the complexity of the risk-informed application will' have a marked bearing on the required analysis as well as the required certification of the supporting PRA analysis. As stated in this document, focused, simple and straightforward applications will require less certification than complicated, larger scale applications. Our response to RAI # 11-1 of the November 1997 submittal, provides a discussion of how the Vermont Yankee N560 application meets the intent of DG-1061 and the five guiding principles delineated in DG-1061. That is,1) meets current regulations,2) maintain defense in depth,3) maintain sufficient safety margins,4) proposed increase in risk is small and 5) performance-based implementation and monitoring strategies. In addition, our responses to RAls # 11-2 (reliance on PRA numbers and quality of the PRA), II-3 (internal flooding), II-4 (defense in depth) and II-8 (internal and external reviews of the VY PRA) addressed the issues relative to 'the quality of the PRA as being adequate for the proposed application'. Considerations applied to the determination of PRA adequacy are discussed in Section 3.0, " Inputs and

' Assumptions", Sub-section 3.3," Plant Level Assumptions", of Reference 13, j excerpted and attached to this submittal. Input documents (e.g. plant design changes and other permanent plant records) used in this determination are

[

contained in several volumes of documentation that are available for staff review at the Bolton, Massachusetts offices of Duke Engineering & Services. In addition

- to the above, approximately 75 percent of the inspection locations result in a high consequence assignment (based primarily on a large LOCA). Another almost 20 1

20

percent of the inspecion locations rescit in a medium consequence assignment.

Assignment oflarge LOCAs or LOCAs outside of containment to a high consequence category (given the component failure) is consistent with the industry knowledge base.

Ouestion 13 - RAI (3/11/98)

In RAI V1.1, an estimate of the impact on core damage frequency (CDF) and large early release frequency (LERF) of the proposed change is made based on bounding estimates.

The basis for these estimates requires further clarification, e.g., provide the basis for the statement that likelihood of pressure boundary failure (PBF) for a pipe location with no degradation mechanism present (Xo ) is expected to have a value lower than 1E-8. Clarify how these estimates address the frequency of LERF.

Answer As stated in the answer to RAI VI.1, the basic likelihood of PBF for a piping location with no degradation mechanism present (Xo) is expected to have a value lower than 1E-8 (per location, per year). This value is selected based on the work performed by Karl N. Fleming and Steve R. Gosselin in assessing pipe rupture frequency based on insights from service experience (" Evaluation of Pipe Failure Potential via Degradation Mechanism Assessment," presented at ICONES Conference in May 1997 in France). In the WOG approach, PFM estimates also consistently produced numerical values lower than 1E-8 for segments with no degradation mechanism, but conservatively used 1E-8 value for those segments. Please refer to the answer to RAI III.1 of the October 1997 submittal.

The frequency of LERF is not directly addressed in the analysis. The LERF evaluation is described in detail in the answer to Question 8.

I LOCAs outside containment are also evaluated and further explained in our responses to Questions 1 and 2.

Based on the attention given to containment performance and the RI-ISI positive impact on CDF, it is expected that the impact on LERF is also positive. A discussion on the evaluation of LERF is provided in our response to Question 8.

Ouestion 14 - RAI (3/11/98)

The licensee has argued that the probability of detecting flaws will be greatly enhanced if inspections are targeted at specific degradation mechanisms. To this end, VY has proposed to provide further training to inspection personnel to increase their knowledge of the subject mechanisms, and the type of ultrasonic indications to be cognizant of when l

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i

s performing inspections under N-560. Ilowever, past experience with the intergranular stress corrosion crzking (IGSCC) phenomena would indicate that, while an increased level of casual knowledge is helpful, it is not sufficient to validate the reliability of the NDE method. It is unclear how the licensee intends to qualify the ultrasonic methods, procedures and personnel used to perform an " inspection for cause" at VY.

Answer Recommended inspection-for-cause examination volumes and methods for application of Code Case N560 are described for each specific degradation mechanism in EPRI TR-106706. The following specific degradation mechanisms were identified in the Vermont Yankee N560 application as applicable to portions of the Class 1 piping:

IGSCC, FAC and Thermal Fatigue. The following paragraphs describe the approach which will be used to qualify the ultrasonic methods, procedures and personnel for examination locations associated with each of these specific degradation mechanisms:

IGSCC - Inspections oflocations identified to be IGSCC susceptible will be performed in accordance with equipment, procedures and personnel qualified in accordance with the 1986 Edition of Section XI for the current IGSCC inspection program. Qualification of personnel is in accordance with the "three-party agreement" for IGSCC inspections.

FAC - Inspections oflocations identified to be FAC susceptible will be performed in accordance with procedures, equipment and personnel used for the current FAC

~ inspection program. These are relatively straightforward ultrasonic thickness measurements. j Thermal Fatigue - Inspections oflocations identified to be thermal fatigue susceptible will be performed using procedures similar to those described above for IGSCC susceptible locations, except that the examination volumes will be expanded to include the geometry at the end of the counterbore (as discussed in EPRI TR-l%706). Because of the nature of thermal fatigue cracks, the IGSCC procedures generally produce extremely large reflectors when cracks are present, and detectability and sizing are not an issue. In fact, there has been a problem in a few applications where thermal fatigue cracks have been discounted as geometric reflectors because the signals were so large and consistent. For this reason, special training of personnel performing thermal fatigue examinations will be conducted to ensure that all reflectors are properly characterized and reported. Actual thermal fatigue crack samples will be used in this training.

Ouestion 15-RAI G/11/98) j l !

The potential degradation mechanisms applied to VY piping components are stated to be the result ofindustry and plant-specific surveys. With respect to this statement provide I the following information:

22 i

a. It is unclear which databases have been reviewed to assess potential degradation at VY. What was the extent of the review of these databases? Are any of these databases considered to be a consensus or industry standard for describing potential degradation mechanisms in light water commercial nuclear facilities?

b. N-560 requires that examinations be performed commensurate with expected degradation mechanisms, however, no analysis is required to determine an optimum inspection frequency - an arbitrary schedule of every ten years is listed. It would appear that an N-560 degradation mechanism evaluation is incomplete with respect to initiation and growth parameters, and how these might impact inspection frequencies. Provide information as to how those parameters are being addressed at -

VY.

c. In the licensee's response to RAI, it was stated that the frequency of occurrence of new or unknown damage mechanisms are bounded by that of known failure mechanisms, based on EPRI studies. Further, that EPRI will continue to monitor service experience and, if new mechanisms are discovered, strategies can be developed to address them. However, the response did not identify any program or formal vehicle in place to impart newly acquired EPRI information pertaining to degradation mechanisms to VY. It is the licensee's responsibility to integrate new information that could affect piping inspection strategies into the VY RI-ISI program. The licensee has indicated that new information will be reviewed via a formal " Operating Experience" program, however no description of how this program correlates industry failure history, nor how a documented revision to de bases for the N-560 alternative will be performed is included in the licensee's submittal. This information needs to be provided.

d. The N-560 evaluation requires that licensee's elevate the risk ranking of piping segments where a potential for water hammer exists. The licensee's statement that

" appropriate measures" are in-place to preclude water hammer is unclear. If this determination is being made as a result of administrative controls versus engineered safety functions, then some probability for water hammer should be considered.

This may impact the ultimate risk ranking of several piping segments. The licensee should describe what appropriate measures have been put in place to eliminate -

water hammer as a potential in applicable systems.

Answer l

l 'a. A number of databases and technice.1 references were used to establish the damage i mechanism screening limits applied to the Vermont Yankee damage mechanism assessment. These . included, but are not limited to, the EPRI Fatigue Management Handbook (Reference 2), the INPO NPRDS Database (Reference 3), several NUREG reports on piping system reliability (References 4,5,6,7 and 8), industry LERs, a technical paper on fatigue failures in nuclear power plants in Japan (Reference 9), and 23 f

c

( an EPRI report on waterhammer (Reference 10). In addition, the Vermont Yankee specific degradation mechanism assessment was the subject of a senior design review by industry experts in metallurgy, corrosion, fatigue and thermal stratification issues in nuclear power plants.

In support of the N560 applications as well as the ASME Code Case N578 pilot plants, EPRI is currectly updating its failure events database which is expected to be published in 1998.

b. The basic tenet of ASME Code Case N560 is that the current ASME Section XI program with its ten-year inspection interval, as supplemented by augmented examinations (such as Generic Letter 88-01) provides an acceptable level of quality 1 and safety. The only mechanism identified as potentially being an aggressive l mechanism is FAC. The inspection interval for this mechanism is in accordance with I the existing FAC program. As such, FAC examinations are conducted based upon

]

analyses and past inspection results and are not tied to a fixed ten year inspection '

interval.

i l

l '

c. The Vermont Yankee Operating Experience Program is defined in Procedure

PP-7012. This program requires that Vermont Yankee review industry operating experience (OE) for its applicability to Vermont Yankee programs and practices. The ,

purpose of this pogram is to insure that lessons learned from industry and in-house j operating experience are evaluated for applicability to Vermont Yankee, and when necessary, translated into corrective actions to minimize the potential for similar events occurring at Vermont Yankee and to improve plant performance. This information generally falls into two categories i

J

1) Internal Operating Experience Item - A significant OE item that occurred at Vermont Yankee. Such items are processed internally via Procedure #AP-0009 and selected items are disseminated to the industry by the INPO Nuclear Network (DP 0027).

2) External Operating Experience - An OE item from sources external to Vermont Yankee such as INPO, NRC, vendors, other utilities, and industry groups such as NMAC, EPRI. The types ofinputs included within the external OE category include: INPO SOERs, SERs, SENs, OMRs and OEs, NRC inspection reports, information notices, bulletins, generic letters, administrative letters, and correspondences to Vermont Yankee, ANI letters, audits and inspection reports, '

vendor service bulletins, information notices and Part 21 notifications, General Electric service advisory letters, service information letters, rapid communication service information letters and turbine information letters. In addition, information is obtained from participation in audits, off-site seminars,

off-site training, trip reports, phones or conversations with staff from other 1

24

L utilities and vendors, as well as representation on other industry group such as EPRI, NMAC and owners groups.

Implementation of the above described program is through Procedure #AP-0038 (Operating Experience Procedure). This procedure assigns Operating Experience process ownership and defines the subject documents, their receipt, screening, evaluation and final disposition. A method for assessing the effectiveness of this process is also provided. The Operating Experience Coordinator (OEC) is the owner of and the single point of contact for the process. The OEC implements the procedure as directed by the Technical Support Manager and is responsible for procedure maintenance and receipt, initial screening and tracking of the OE items through the process using the Commitment Tracking database (CTS).

The CTS tracks operating experience issues and corrective actions from initiation to completion. This systern is controlled by the OEC and is defined in Procedure

AP-0028.

In addition to the above required program, Vermont Yankee maintains cognizance of industry activities and issues by its strong participation in programs such as the BWRVIP, BWROG, EPRI NDE Center and ASME Section XI.

d. As part of the N560 application, a plant specific review of water hammer experience was conducted. For the scope of piping under evaluation only the RHR system was identified as previously susceptible to water hammer based upon past operating practices. Prior to the N560 application effort, plant operating practices were revised to flush the system and slowly collapse any steam / vapor pocket = using the condensate (

system prior to RHR cut-in. This process essentie.Ily eliminates ne potential for vapor pockets and preheats the RHR fluid thereby precluding the potential for waterhammer.

As discussed in several NRC and industry studies (References 10,11 and 12), .

effective water hammer prevention can be accomplished by a variety of means j including improvements to operational practices.

However, the above notwithstanding, based upon the risk ranking results, six locations within the Class 1 portion of the RHR system were still selected for inspection. I I ,

I 25 l

Ouestion 16 -RAI O/11/98)

Attachment 2 of the VY submittal entitled " Mechanisms Specific Examination Volumes

. and Methods" documents the basis for the selection of examination volumes and methods that will be utilized at VY. However, no information is provided pertaining to qualification ofexamination procedures, examination equipment and personnel. In order

' to ensure an adequate examination of the selected examination volumes, the examination procedures, equipment and personnel must be properly qualified. With respect to this area provide the following information,

a. Will VY use its own examination procedures, equipment and personnel to perform ultrasonic examinations of selected examination volumes?

b.- Ifit does, are the procedures, equipment and personnel qualified in accordance with the requirements of Appendix VIII of the ASME Code,Section XI? i

c. If contractors are used to perform these examinations, are the contractors required to use procedures, equipment, and personnel qualified to Appendix VIII of Section j XI?

Answer i

a. In general, Vermont Yankee uses its own examination procedures, equipment and j i

personnel to perform ultrasonic examinations ofpiping for ASME Section XI required inservice inspections. These are supplemented by contract personnel, working to Vermont Yankee procedures trained to the Vermont Yankee program.

Contract personnel are qualified under the vendor qualification and certification program in accordance with ASME Section XI requirements.

]

b. Vermont Yankee does not at this time have a formal requirement to qualify examination procedures, equipment and personnel under Section XI, Appendix VIII. However, it has been Vermont Yankee's standard practice to use personnel and procedures qualified in accordance with the PDI Program Decription Document i to the extent practical. When the NRC formally mandates Appendix VIII or

approves a Section XI code edition containing Appendix VIII, and the Vermont Yankee program is updated to that Code Edition, Appendix VIII will be formally applied. However, this decision is considered to be independent of the application of Code Case N560, since Appendix VIII is not a requirement of the Code Case.

L Until that time, qualification of examinations will be per existing Vermont Yankee 1 procedures, with the special " inspection for cause" considerations discussed in our response to Question 14, above.

26

I l

c. Contractors working on the Vermont Yankee inservice inspection program are qualified under the vendor qualification and certification program in accordance with ASME Section XI requirements and are fwther indoctrinated under the Yankee Nuclear Services Division Procedure YA-TP-1, which addresses training and orientation of NDE contractor personnel. The training and orientation covers Yankee procedures, documentation requirements, interfaces and general methods for performing MDE sctivities. The program may also include written examinations and proficiency demonstadions for all applicable NDE methods, with oversight by the Yankee Level 111.

1 27

f '

References:

1. " Risk Informed Inservice Inspection Evaluation Procedure," EPRI TR-106706, June 1996.

I

2. "EPRI Fatigue Management Handbook," EPRI TR-104534, December 1994.

3. INPO NPRDS (Nuclear Plant Reliability Data System) Database, sorted by "cause of failure = aging / fatigue cycling".

4. NUREG/CR-4731, EGG-2469," Residual Life Assessment of Major Light Water Reactor Components - Overview," INEL, USNRC, November 1989.

l S. NUREG-0679," Pipe Cracking Experience in Light Water Reactors," L. Frank et al,

August 1980.

)

6. NUREG-0691, " Investigation and Evaluation of Cracking Incidents in Piping in {

Pressurized Water Reactors," PWR Pipe Crack Study group, USNRC, September 1990.

t 7. NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking," USNRC November 1980.

t

! 8. NUREG-0131, Revision 2 " Technical Report on Material Selection and Processing I Guideline for BWR Coolant Pressure Boundary Piping," USNRC January 1988.

9. K.Iida,"A Review of Fatigue Failures in LWR Plants in Japan," Nuclear Engineering and Design,138 (1992), pp297-312.

I

10. " Water Hammer Prevention, Mitigation, and Accommodation," EPRI NP-6766, July 1992.

11. NUREG-9027, Revision 1 " Evaluation of Water Hammer events in Nuclear Power Plants," USNRC March 1984,

12. EE Report No. AEOD/E91-01, "A Review of Water Hammer Events After 1985,"

USNRC Feruary 1991.

13. YC-386 Revision 0 and Revision 1, " Consequence Evaluation of Vermont Yankee Class 1 Piping in Support of ASME Code Case N560".

14. VYC-1795, Revision 0," Degradation Mechanism Evaluation For Class 1 Piping".

15. SAVY98-008," Vermont Yankee N560 Results".

26

Attac ament 1

(

Results ofIndependent Review j Excerpted from:

YC-386 Revision 0 and Revision 1 Consequence Evaluation of Vermont Yankee Class 1 Piping in Support of ,

ASME Code Case N560 l 4

(7 Pages) l l

I l

I

JUL -3M%G W8 49 P.G1

' )

i Page1 Calculation No. YC-386

$- FOR INFORMATION 0NW oarorNAt: PAos i er io4 eAGES Rev.1: PAGE 1 of PAGES Rev. 2: PAGE 1 of PAGES Rev.3: PAGE 1 of PAGES QA RECORD 7 RECORD TYPE NO. 13.C16.053

,,)L. YES Safety Class /P.O. NO. (if applicable) N/A NO YANKEE NUCLEAR SERVICES DIVISION CALCULATION / ANALYSIS FOR TITLE ConsqucngcEnlustiagf Vermont Yankee Class 1 Picing in Support of ASME Code rau N-560 PLANT Vermont Yankee (VY) CYCLE N/A CALCULATION NUMBFJL YC-386 .

$k'Q Q3 PREPARED BY REVIEWED BY APPROVED BY SUPERSEDES g /DATE , /DATE /DATE CALC /REV. NO.

ORIGINAL 'g#,W kr.b * '

ll}& 4y //-H -77 n- 14 -4 7

/ i i

KEYWORDS Inservice Insoection GSD. Pmbabilistic Risk Assessment iPRAL Risk Based. Risk InfonntAASMBEnstion _XL ConEeauence Analysis. Pinine. ASME Code Case. RISKMAN t

COMPUTER CODES: N/A EquiPyAO NO..: m3 FOR INFORMATION ONLY SYSTEMS: N/A

')

REFERENCES:

See Pagefalculationfage 66 PORM WB-103-1 Revision 4

l

Page 2 i Calculation No. YC-386 i EVALUATION OFCOMPUTER CODE USE CALCULATION NO. YC-386 REVISION NO. Original l

I j' List the computer codes used, and complete the following:

Approved Appropri- Out.:tanding -

per WE-108 1 ateness SPRs' Verified2 Code Name/ Version and/or Script File Yes No Yes No Yes* No RISKMAN X X X l

l l

??Pg

. / Refer to Section 4.1.4.4, Bullet 3, of this procedure.

2 Refer to Section 4.1.4.4, Bullet 2 of this procedure.

8 I To generate an SPR, refer to WE-108.

l Check yes only if the SPR is "ma,)or". Fill out information below, also l li a computer code was not verified per WE-108, or if there are outstanding SPRs, state below why it is !

appropriate.

1 Code Name/ Script File Appropriateness RISKMAN - See Calculation Page 28, Assumption 10 l {

i i

l 1

) FORM WE-103-2 l

l Revision 4 l 1

9

j Ju_-30-1993 00:50 P.03 l Page 3 a Cciculation No. YC-386 M CALCULATION / ANALYSIS REVIEW ll CALCULATION NO. YC-385 REVISION NO. Original COMMENTS RESOLUTION C*dek&a ws nme ed for somHreos eoM ,, a non , s a w,,/

mAdey seeroor/deesi m e nse //> ser i&'6 /ol.

1 Sere is /jor A / laci.s / d' no /'re 4 A fje. bL~i /$ Aduune! $"Y A SoliSfY owG istsd Avse// mon.s adora, hM A_AY 'amY= 8 We v en c!>effeef fer McAfrase e/- M Q28#$YYr'bf_ _

? new/v /no4.1.

l 1Trerekskkee afece no1's Nof/V */$ero lo hr t'f h1t AJltIf f AJpspa _

'z Mr <** sea.eaa. erSmr /r Asifro*i nrowa dnu kd $'s S l N

? in Ta / /~y e Era a4e d fn h5 $,

I Scop oc tir co4,Jr no 6as fa /) Sec.b n f u d 2 YC Y'!<d is aof e/Serr fe r/kn ks

/

l l

f Identify motbod(s)of review:

g Calculation / analysis review Alternative calculationalmethod l _ Qualification testing Resolution By: // 4 f

P/ /Date Commen ontinued on P e:

~

Concurrence with Resolution #- r M /M(~/#

Reviewer /Date

)

..' FORM WE-103-3 Revision 5

Jtt-30-1998 00:50 P.04 Pags 4 s, Calculation No. YC-386

, NED ANALYSIS PROCESS CHECKLIST Preparer Reviewer Name J. H. Moody Name S. M. Follen (please print) (please print) i

> l l

Organization J. H. Moody Consulting,Inc. Organization Yankee Atomic Electric Co.

Signature / Signature gAyM Date ((h[qf [ Date //- As-fy

/

Raouirement Preparer Reviewer l Ensu6e that the method desorfbed in the MOM,if applicable, and 1he base calculation, if one exlets, has been followed A/[A'

/

if not, ensare lead engineer / manager is consulted and N!A /V/9 ,

document variationin calculation / l Ensure that other applicable NED Procedures are implemented /47 s Ensure inpute/ assumptions are obtained from appropriate /97

., eources f l Ensure any change to an input / assumption is consistent with A/!A MA8 operating practios at the plant /

Ensure that the safety analysis conforms to appiioable //!A requirements /

Ensure that intermediate results that would require a change to 4 #-

plant operating practice are dispositioned and documented Ensure, if reporting preliminary results, that the standard A /V[#-

memorandum clearly states the results are preliminary and /

provides the status of the final analysis '

/V/#

Ensure that issues found when performing an analysis are dispositioned and documented /

Ensure a andard memorandum is written describing the analysis performed and containing the following elements:

r NE Any documents affected by the analyels are updated

{'/4 A

['

Rooommend updates be incorporated in the M!d A/

affected documents and include an action taken /'

feedback block l

L__ _ _ - . - - - - - - - - - - - _ - - - - _ - - - - _ - - - - . - - - - - - - - _ - - - - - - - - - - - - - -- - - - - - - - - - - - - - - - - - - - - - - - - - -

~~

~~ JLL-30-1933 08:51 P.05 i Page5 n 386 Mfg O NED ANALYSIS PROCESS CHECKUST (continued)

Reautrement M . Reviewer h, M* IO N Ar[4f ed /

Administrative Assistam

- Provide a Est of personnel for distribution at YNSD and the sponsor

/W Notify project and sponsor licensing groups of any NRC reporting requirements M

/

8[#

include a safety evaluation, if a plant operating prachceis affected M M!#

e Ensure that this cheeldist has been filled out and is attached to V

/ '

both the memorandum, placed in department chronolopcal files, f "(' /

and the calculation. (NOTE: The checidlet does not need to be attached to detributed copies of the memorandum.)

jp.

D.h.

l l

)

I L___ J

JLL-30-1998 OB 51 P.06 Page 6 i

.m Calculation No. YC-386

$'gj NED WE-103 REVIEW CHECKUST !

Reviewer compliance Reviewer Name 8. M. Follen Name M. Cote (Pi ne*ePrkt) (pkeneswint)

Orgenlaation Yankee Atomic Electric Co. Organization Yankee Atomic Electric Co.

Signature g7d Signature Mh@

Date //~M - f/ Date W dq}

Requirement Reviewer Compliance

_ Reviewer Ensure the title page is appropriately filled out. 4M e/ Correct number of pages.

/ QA Record filled out.

d> lMS number flRed out. #M

/ Record number filled out (13.000.001 included if microfiche or hard copy of oormuler runs are attached to the calculation).

/ Descriptive title.

/ Plant, cycle number and calculation number included. 'N/A' can

. be used for plant and cycle number.

/ Signatures and dates are included, and are in correct Q,%/ .

chronological order. Print the name and individuals' organization

~

(if other than YAEC) below the signature. The title page reviewer and approar dates do not pre date any date in the calculation except for changes containing that individual's initials and date.

/ All WE-106 computer codes and other keywords not in the title which can be used to retrieve the calculation are usted in the d keyword fleid.

Ensure the Form WE 103-2 is included and properly completed when a #W IU {

computercode is used. -

Ensure Form WE 103 3 is included, and has signatures / dates from both /W fvt5C.-

the preparer and the reviewer and that all comments have boon addressed, if no comments, use the foBowmg statement: " Reviewed in accordance with WE-103 with no comments".

Ensure review of the coloulation can be done without recourse to #4D WA the originator.

Ensure computer codes are used in accordance with WE 103 An trur .

Steps 4.1.4.4 through 4.1.4.0.

Ensure the calculation includes a title page, objective, method, inputs, assumptions, calculations, results, conclusions and references.

/9C N l

Ensure the inputs are referenced to formal documents, e.g., WE 103. The # N/A s referonos can not be a YAEC report unless formal QA records are checked

) and also referenced.

Ensure doelen input intamal and extemal concepondence is prepared and _/ W ~

N/A ,

reviewed, and is, therefore, a QA record. If there is only one signature on l the concepondoms, vertly that it is a QA record Ensure that if design spoolfications were used as input to the omiculation, the / N/A

~ ~~ ~~

TLC-M9sf 'asisF P.W l Pagc 7 l 3 Cdculation No. YC-386

-) Ranuiramant Reviewer Compliance

= i performance characteristics are verified in writing by the provider of the i w ,v ,.i@roduct s or by cognizant YAEC%lant personnel.

Ensure that input and modeling unoortainties are expliottiy addressed /# WA in the omiculation.

Ensure that the applicable input considerations fmm WE 100, Table 1 W.r MA have been incorporoled and are emp5citly addressed within the calcuistion.

Ensure individuals responsible for each portion of the coloulation are /V N h identified when multiple preparers and/or reviewers are utlized. Page initialing is optional, even in the cases where initial boxes are provided on the pages.

Ensure each page has a page number and the onioulation number and 4 W'fr ~

w ) Ah revision numberif applicable. Dates on each page are ophonal Ensure that every page of every attachment (or Appendix) contains its # A/ 16 W4 attachment (or Appendix) number. -Ne f rep * *we topY (*Mra P Ensure a conclusion is stated in a supplemental Revision. N/# N. >

Ensure corrections are addressed in one of the following approaches: [lP 4, ' *

Retyped and identified by a vertical line with revision number, it

'y apphoeblo,in the right margin; OR

Lined out, initialed and dated by preparer; OR

~

Photocopy of original to eliminate any previous conection tape, whiteout, or erasures.

Ensure enhancements and clouding arn initialed and dated. l/[# M A Confirm logbility meets WE-103, Attachment A. Spoolfic pages can be ## pd/

exempt if they are: (1) documents received from another organization who is the original QA custodian, or (2) supplemental pages included for information only. In these two cases, make sure a memo was issued to RMS per WE-002 Section 3.4.3.

Fleview of 10CFR60.46 reporting requirements has been documented for N/A analyses which assess ceric , nce with 10CFR60.46.

Ensure computer codes are validated for the computing environment. #M WA Ensure script files are included in the calculation or referenced to another /l' 4 08 WA-calculation. Also, ensure the preparer identifies how the code / script was run.

Ensure applicable outstanding Engineering Defiolency Reports (EDRs) have MP WA been reviewed for influence on tim calculation and note review in omiculation.

Ensure relevant SER conditions / limitations have been reviewed for their /

effect on this calculation and the review is noted in the calculation.

l l

Page1 Calculation No. YC-386, Revi ORIGINAL: PAGE 1 of 104 PAGES FOR INFORMATION ONLY Rev.1: exoE i of iO4 exoES I Rev. 2: PAGE 1 of PAGES Rev. 3: PAGE 1 of PAGES QA RECORD 7 RECORD TYPE NO. 13.C16.053 X_ YES Safety Class /P.O. NO. (if applicable) N/A NO YANKEE NUCLEAR SERVICES DIVISION CALCULATION / ANALYSIS FOR TITLE Consequence Evaluation of Vermont Yankee Class 1 Pininc in Support of ASME Code Case N-560 PLANT Vermont Yankee (VY) CYCLE N/A CALCULATION NUMBER YC-386 1

PREPARED BY REVIEWED BY APPROVED BY SUPERSEDES

/DATE /DATE /DATE CALC./REV. NO.

ORIGINAL J. H. Moody S. M. Follen M. F. Kennedy N/A 11/26/97 11-26-97 11-26-97 Revision 1 .1. Mo dy P. J. O'Regan YC-386/ Original l M. F. Kenng O h/f)&,"% % Y T. Ya h g '

/ -

yon ttuk.Cinw.WAs/

k KEYWORDS Inservice Inspection (ISD. Probabilistic Risk Assessment (PRA). Risk Based. Risk Informed. ASME Section XI. Consequence Analysis. Pipinc< AShE Code Case. RISKMAN

. COMPUTER CODES: E!/A

"*t"^ " ': "'^

SYSTEMS: N/A FOR INFORMATION ONLY p

REFERENCES:

See Pace Calculation Pace 66

, FORM WE-103-1 l1 Revision 4

I Page 2 Calculation No. YC-386, Revi -

EVALUATION OF COMPUTER CODE USE 1 CALCULATION NO. YC-386 REVISION NO.1 l l List the computer codes used, and complete the following:

Approved Appropri- Outstanding

" 3 3 per WE-108 ateness SPRs 2

Verified Code Name/ Version and/or L.:ript File Yes No Yes No Yes' No RISKMAN X X X 8

Refer to Section 4.1.4.4, Bullet 3, of this procedure.

2 Refer to Section 4.1.4.4, Bullet 2, of this procedure.

3 To generate an SPR, refer to WE-108.

Check yes only if the SPR is " major". Fill out information below, also If a computer code was not veiified per WE-108, or if there are outstanding SPRs, state below why it is appropriate.

,; Code Name/ Script File Appropriateness -

RISKMAN See Calculation Page 28, Assumption 10 i

?

i I I. )

i FORM WE-103-2 Revision 4 l

l l: '

3 Page 3 Calculation No. YC-386, Revi CALCULATION / ANALYSIS REVmW CALCULATION NO. YC-386 REVISION NO. _1 l COMMENTS RESOLUTION V eto A ~

s cewAcace w:P1 si m TG,,,,,.ej.

(i /EE - /0 5 tu?lh i,a r rw._ . 4-c ii.

Ideptify method (s) of review:

J g Calculation / analysis review Alternative calculational method A Qualification testing lk.

i Resolution By:

p ste Comments ntir. a I .1 (.

I: -

Concurrence with Resolution //ge:>61 Mil-en

, / ' flieviewer/Daic FORM WE-103-3 Revision 5

~

j?

l.

Page 4 Calculation No. YC-386, Revi .

NED ANALYSIS PROCESS CHECKLIST l

Preparer Reviewer Narne J. H. Moody Name P. J. O'Regan (please print) (please print) l I

Organization J. H. Moody Consulting, Inc. Organization DE&S l

Signature h i a Signature

[,f// g g y Date e}-l,;t.(,ffg Date , /-gg ,pg i

Requirement Prenarer Reviewer l Ensure that the method described in the MOM, if applicable, and A/!M _.Al A the base calculation,if one exists, has been followed '

A

/ l If not, ensure lead engineer / manager is consulted and l document variation in calculation / i Ensure that other applicable NED Procedures are implemented b TT Ensure inputs / assumptions are obtained from appropriate sources

/1A b 'R Ensure any change to an input / assumption is consistent with N

, operating practice at the pl3.nt /

i Ensure that the safety analysis conforms to applicable Al requirements # /

Ensure that Intermediate results that would require a change to plant operating practice are dispositioned and documented k /

/

Ensure, if reporting preliminary results, that the standard memorandum clearly states the results are preliminary and I /

provides the status of the final analysis c

Ensure that issues found when performing an analysis are dispositioned and documented M

/

p Ensure a standard memorandum is written describing the enalysis performed and containing the fo!!owing elements: /

bJYR Any documents affected by the analysis are g updated ,/ /

Recommend updates be incorporated in the #7 affected documents and include an action taken /

/

feedback block

l Page 5 l Calculation No. YC-386, Rev t NED ANALYSIS PROCESS CHECKLIST (continued) i Requirement Preoarer Reviewer

- If the memorandum recommends updates to ll/ A/

affected documents, copy the NED '/ /

Administrative Assistant Provide a list of personnel for distribution at YNSD and the sponsor

DON Notify project and sponsor licensing groups of any NRC reporting requirements /

M MN !

include a safety evaluation,if a plant operating ///

practice is affected ff '

Ensure that this checklist has been filled out and is attached to both the memorandum, placed in department chronological files, '

bK !

and the calculation. [ NOTE: The checklist does not need to be

l cttached to distributed copies of ;Se memorandum.)

i l j

. I

.l l

i

')

h i i s 9

1 l

t i#

Page 6 U No. YC-386, Revi ,

NED WE-103 REVIEW CHECKLIST Reviewer Compliance Reviewer Name P. J. O'Regan Name

(pleaseprint) (please print) hQR (g l Organization DE&S Organization DE&S l

Signature /, g b Signature /Jk h Date p [g,., Date 2//o N Requirement Reviewer Compliance Reviewer Ensure the title page is appropriately filled out. D '("L.

Correct number of pages.

OA Record filled out.

IMS number filled out.

l, a

Record number filled out (13.009.001 included if microfiche or hard copy of computer runs are attached to the calculation).

Descriptive title.

Plant, cycle number and calculation number included. 'N/A" can be used for plant and cycle number.

Signatures and dates are included, and are in correct j chronological order. Print the name and Individuals' organization

! (if other than YAEC) below the signature. The title page reviewer and approver dates do not pre-date any date in the calculation except for changes containing that individual's initials and date.

i

All WE 108 computer codes and other keywords not in the title Ab which can be used to retrieve the calculation are listed in the keyword field. '

.g_ ZMs Ensure the Form WE-103-2 is included and properly completed when a computer code is used.

6dh th 1

Ensure Form WE 103 3 is included, and has signatures / dates from both the preparer and the reviewer and that all comments have been addressed.

970'IL ,Ab If no comments, use the following statement:" Reviewed in accordance with WE-103 with no comments".

5 Ensure review of the calculation can be done without recourse to ("ShI N/A the originator.

ftSD'/L N-f 8 j Ensure computer codes are used in accordance with WE-103 NIIr d

Steps 4.1.4.4 through 4.1.4.6.

i j Ensure the calculation includes a title page, objective, method, inputs, assumptions, calculations, results, conclusions and references.

9JD4L /b fM'd Ensure the inputs are referenced to formal documents, e.g., WE-103. The N/A reference can not be a YAEC report unless formal OA records are checked

?

and also referenced. l t

Ensure design input internal and external correspondence is prepared and D (L- N/A i

reviewed, and is, therefore, a QA record. If there is only one signature on ;

the correspondence, verify that it is a OA record. 1 Ensure that if design specifications were used as input to the calculation, the MN N/A

Page 7 Calculation No. YC-386, Revi Beavirement Reviewer Compliance Revhn.ar i ;

performance characteristics are verified in writing by the provider of the component / product or by cognizant YAEC/ plant personnel.

. Ensure that input and modeling uncertainties are explicitly addressed V3DTL h WO N/A l in the calculation.

Ensure that the applicat?., input considerations from WE-100, Table 1 TD0'O_ N/A have been incorporated and are explicitly addressed within the calculation.

Ensue individuals responsible for each portion of the calculation are identified when multiple preparers and/or reviewers are utilized. Page N@ d initiating is optional, even in the cases where initial boxes are provided on the pages.

Ensure each page has a pag [5.:mber and the calculation number and ' 93D revision number if applic&ble. Dates on each page are optional.

Ensure that every page of every attachment (or Appendix) contains its attachment (or Appendix) number.

A /A 4 '

Ensure a conclusion is stated in a supplemental Revision. D;D GL h

Ensure corrections are addressed in one of the following approaches: u/b Mb

=

Retyped and identified by a vertical line with revision number, if applicable,in the right margin; OR

=

Lined out, initialed and dated by preparer; OR Photocopy of original to eliminato any previous correction tape, whiteout, or erasures.

j Ensure enhancements and c'ouding are initialed and dated. dA [6(/

Confirm legibility meets WE-103, Attachment A. Specific pages can be h0N2. Mb exempt if they are: (1) documents received from another organization who is the original QA custodian, or (2) supplemental pages included for informadon only. In these two cases, make sure a memo was issued to RMS per WE-002 Section 3.4.3.

! Review of 10CFR50.46 repoiting requirements has been documented for MM N/A analyses which assess conformance with 10CFR50.46.

P10g. UC-/a 3

Ensure computer codes are validated for the computing environment. -$rfM N/A Ensure script files are included in the calculation or referenced to another MlA N/A 3

calculation. Also, ensure the preparer identifies how the code / script was run.

Ensure applicable outstanding Engineering Deficiency Reports (EDRs) have 9:5[A N/A been reviewed for influence on the calculation and note review in calculation.

Ensure relevant SER conditions / limitations have been reviewed for their MA- E effect on this calculation and the review is noted in the calculation.

?

CC WPl03 6?chtett 4d A 3 has M. hSed le c)xyrde. bS ft iQltlevd. h 2 h N

\

I 1

i

l l

Attachment 2 Results ofIndependent Review Excerpted from:

VYC-1795 Revision 0 Degradation Mechanism Evaluation For Class 1 Piping (8 Pages)

,,, ORIGINAL: PAGE 1 of 130 PAGES

-0 R 'DJFORM/iTl0N ^" v Rev.1: '""'""""'""

^U11L3 Rev. 2:

PAGE 1 of PAGE 1 of PAGES PAGES Rev. 3: PAGE 1 of PAGES QA RECORD?

X YES RECORD TYPE 13.C16.036 NO Safety Class /P.O. NO. (if applicable) 4896 YANKEE NUCLEAR SERVICES DIVISION CALCULATION / ANALYSIS FOR TITLE Degradation Mechanism Evaluation for Class 1 Piping PLANT Vermont Yankee CYCLE Generic CALCULATION NUMBER VYC-1795 PREPARED BY REVIEWED BY APPROVED BY SUPERSEDES

/DATE /DATE /DATE CALC /REV.NO.

[

N/A P. J. O'Regan 6-73-/8 M. F. Kennedy Z

~

G. P. Semienko 2/20/98 REVISION 1 REVISION 2 REVISION 3 KEYWORDS In-Service Inspection, ISI, risk based, PRA, EPRI, degradation mechanism, corrosion, cavitation, thermal f atigue, chlonde cracking, welds, thermal sleeve, Structural Integnty COMPUTER CODES : N/A SYSTEMS : CS, FDW, HPCI, MS, RPV, '

i=

GU'A QI,DQ g((h hhL M

REFERENCES:

(see page number 15) j FORM WE.103-1 ,

Revision 4 l

l 2/20/98 calc VY2. DOC l

[ ]

MEMORANDUM ,

DE&S-BOLTON To RMS Date Febnaarv 23.1998 Group # SAVY98-010 From P. J O'Regan w,0,#

Subject VYC-1795 Lecibility Requirements . I.M.S.#

File # 98-010.WPD Attached is VYC-1795, Revision 0. Several pages are exempt from legibility requirements as these pages are pan of another organization (i.e. Structural Integrity Associates) who is the original QA custodian. These pages are I-16, I-25, I-39, I-48, I-55 and I-67.

bs

' ' ' ~

z- 7 ? p 8 P. .( O'Re[an Date Safety Assessment /NED

/ cab i

l l

l VYC-1795 Rev.0 Prepared By: Page: 2 )

l CALCULATION / ANALYSIS REVIEW CALCULATION NO. VYC-1759 REVISION NO. Original COMMENTS RESOLUTION

'ilN . na d- z / actlw<- v1 JA Mm 11e a,eaL fog /o A m,74 % com A Identify method (s) of review:

UD Calculation /analusis review O Alternative calculational method O Qualification testing Resolution By: ,, , ,5. ps , d!,15/ff 5

' ' 1 reiffer/Date ~ /[

Comments Continued on Page: N/A Concurrence with Resolution M[l/A 't- 2 e-19 Rdviewer/6 ate FORM WE-103-3 Revision 5 l

I l

VYC 1795 Rev.0 Prepared By: Page: 3 -

NED WE 103 REVIEW CHECKLISI .

Reviewer Compliance Reviewer Name Name please print) TaktetM. K O

Q4a w (please print) & ARll fnTC -

Organization Organization OG 4 S i

D6&S Signature Signature Date ' V '

Date I 2-2.% 96 'Ll25I9%

Compliance Requirement Reviewer Reviewer Ensure the title page is appropriately filled out.

. Correct number of pages. .

. QA Record filled out.

. IMS number filled out. 'J/A Record number filled out (13.C09.001 included if microfiche or hard copy of computer runs are attached to the calculation).

. Descriptive title.

. Plant, cycle number and calculation number included. "N/A" an be used for plant and cycle number.

. Signatures and dates are included, and are in correct chronological order. Print the name and individuals' organization (if other than YAEC) below the signature. The title page reviewer and approver dates do not pre-date any date in the calculation except for changes containing that individual's initials and date.

. All WE-108 computer codes and other keywords not in the title which can be used to retrieve the calculation are listed in the keyword field.

,K ,Y Ensure the Form WE-103-2 is included and properly completed when a computer code is used. h/A fJ/A

/

l Ensure Form WE-103-3 is included, and has signatures / dates from both l the preparer and the reviewer and that all comments have been addressed. If no comments, use the following statement: " Reviewed in i l

accordance with WE-103 with no comments", bM ITC.-

Ensure review of the calculation can be done without recourse to the originator. DE N/A Ensure computer codes are used in accordance with WE-103 Steps 4.1.4.4 through 4.1.4.6. O/A N lb b

Form NED 3.1 Rev.1 (effective 9/1/96)

Prepared By: Page: 4 VYC-1795 Rev.0 NED WE-103 REVIEW CHECKLIST (continued)

Compliance Requirement Reviewer Reviewer Ensure the calculation includes a title page, objective, method, inputs, assumptions, calculations, results, conclusions and references. R. nr 6 Ensure the inputs are referenced to formal documents, e.g., WE-103.

The reference can not be a YAEC report unless formal QA records are l

checked and also referenced, b,i?. N/A Ensure design input internal and external correspondence is prepared and reviewed, and is, therefore, a QA record. If there is only one I signature on the correspondence, verify that it is a QA record, f3,fl. N/A Ensure that if design specifications were used as input to the calculation, the performance characteristics are verified in writing by the provider of the component / product or by cognizant YAEC/ plant personnel. k4- N/A Ensure that input and modeling uncertainties are explicitly addressed in ,

P3D(2 N/A the calculation.

Ensure that the applicable input considerations from WE-100, Table 1 have been incorporated and are explicitly addressed within the calculation, RID'fL N/A Ensure individuals responsible for each portion of the calculation are identified when multiple preparers and/or reviewers aie utilized. Page initiating is optional, even in the cases where initial boxes are provided on N/A A the pages. J Ensure each page has a page number and the calculation number and revision numberif applicable. Dates on each page are optional. Rrd r1 MfC-Ensure that every page of every attachment (or Appendix) contains its ,

MEC, PJoyt attachment (or Appendix) number.

AllA Ensure a conclusion is stated in a supplemental Revision. JhfANA /

Ensure corrections are addressed in one of the following approaches:

e Retyped and identified by a vertical line with revision number, if

, applicable,in the right margin; OR

Lined out, initialed and dated by preparer; OR l

Photocopy of original to eliminate any previous correction tape, _i whiteout, or erasures. NlA 8C._.

Ensure enhancements and clouding are initiated and dated. d 14 l Form NED 3.1 l Rev.1 (effective 9/1/96) >

I l

l VYC-1795 Rev.0 Prepared By: Page: 5 l

NED WE-103 REVIEW CHECKLIST fcontinued)

Compliance Requirement Eg/iewer Reviewer 1

Confirm legibility meets WE-103, Attachment A. Specific pages can be j exempt if they are: (1) documents received from another organization {

who is the original QA custod!an, or (2) supplemental pages included for information only, in these two cases, make sure a memo was issued to ,

RMS per WE-002 Section 3.4.3. PJ0 ft _ MC -

Review of 10CFR50.46 reporting requirements has been documented for analyses which assess conformance with 10CFR50.46. f30 *ft N/A Ensure computer codes are validated for the computing environment. IVlA N/A Ensure script files are included in the calculation or referenced to another, calculation. Also, ensure the preparer identifies how the code / script was run. MIN N/A Ensure applicable outstanding Engineering Deficiency Reports (EDRs) have been reviewed for influence on the calculation and note review in calculation. PJD'0- N/A Ensure relevant SER conditions / limitations have been reviewed for their effect on this calculation and the review is noted in the calculation.

dA tfA 1

I l

Form NED 3.1 j l

Rev.1 (effective 9/1/96)

VYC-1795 Rev.0 Prepared By: Page: 6 NED ANALYSIS PROCESS CHECKLIST Preparer Reviewer Name George P. Semienko Name .

(please print) (please print) p ,g%"

Organization Safety Assessment, NED Organization qg yg 3gg

]

Signature f "

[ g Signature pgjg i Date [g Date f y,7;_gg '

Requirement Preoarer Reviewer Ensure that the method described in the MOM, if applicable, and the //M '

o//4 base calculation, if one exists, has been followed If not, ensure lead engineer / manager is consulted and M!A N/4- )

document variation in calculation Ensure that other applicable NED Procedures are implemented FNrt__

Ensure inputs / assumptions are obtained from appropriate sources f3D'rt.

Ensure any change to an input / assumption is consistent with

[ fJD'/L operating practice at the plant Ensure that the safety analysis conforms to applicable requirements 8 M/A Ensure that intermediate results that would require a change to /V N/A plant operating practice are dispositioned and documented Ensure, if reporting preliminary results, that the standard NM '

N/A memorandum clearly states the results are preliminary and provides the status of the final analysis Ensure that issues found when performing an analysis are A N/A dispositioned and documented Ensure a standard memorandum is written describing the analysis 93D R l performed and containing the following elements:

. Any documents affected by the analysis are updated NA-

Recommend updates be incorporated in the affected documents M MN and include an action taken feedback block /

( NED Procedure No. 6 Rev. 2 Date 5/7/97 L-. - --

- VYC-1795 Rev.0 Prepared By: Page: 7 NED ANALYSIS PROCESS CHECKLIST (continued)

Requirement - Preoarer Reviewer

. If the memorandum recommends updates to affected N!A A)l4 documents, copy the NED Administrative Assistant

. Provide a list of personnel for distribution at YNSD and the ( 93D k sponsor

. Notify project and sponsor licensing groups of any NRC M!M TVI4-reporting requirements e include a safety evaluation,if a plant operating practice is f/ ! N/h affected Ensure that this checklist has been filled out and is attached to both the memorandum, placed in department chronological files, and the hp(1 FJD'E calculation. [ NOTE: The checklist does not need to be attached to distributed copies of the memorandum.]

i l

I NED Procedure No. 6 Rev. 2 Date 5/7/97

Attachment 3 i

Inputs and Assumptions Excerpted from:

YC-386 Revision 1 Consequence Evaluation of Vermont Yankee Class 1 Piping,in Support of ASME Code Case N560 (7 Pages) l l

l

Page 23 Calculation No. YC-386, Rev 1 3.0 inputs and Assumptions In addition to Reference 1, numerous plant specific documents are reviewed (see Section 7 references); the following are key bputs to this analysis:

The VY PRA (References 2 and 3) is used to assess plant initiating event challenges and their frequencies, conditional core damage probability, and the importance and unavailability of systems, trains, and accident sequence types, The PRA also contains information on systems operation, dependencies, safety functions, and spatial consequences.

VY flow diagrams and isometrics (Reference 4) are utilized to identify piping locations, l

isolation valves, and detection capability. l Additional reviews, inputs, and assumptions are sur unarized in the following subsections.

3.1 PRA Review The VY PRA (Reference 2) is used to evaluate the importance ofinitiating events, systems, safety functions, and spatial locations affected by potential pipe leaks and/or failures.

Core damage frequency depicted in the VY PRA is approximately 4.3E-6/yr (Reference 2, page 1.4-1). Table 2-1 shows the contribution of accident initiator types as well as conditional core damage probability (CCDP) which provides an indication of overall mitigation capability for each initiator. Table 3-1 shows how CCDP contributes by safety function for certain initiating i events in this analysis. A simplified representation of the safety functions can be found in Figure 3-1.

4 The following summarizes the major contributors to core damage in the PRA based on a review

of the first 10 sequences (Reference 2, Table 3.4.1). Also, the contribution from safety functions and systems is discussed below and in the next section.

!'

Sequences 1 & 2 - loss of the power conversion system (initiating events TFWMS and TLP) and subsequent failure of RCIC, HPCI, and emergency depressurization are the top two sequences. The frequency of these scenarios (each is approximately 4E-7/yr) and the l initiating frequency (0.1/yr) confirms the " Medium" consequence importance of loss of feedwater initiators in Table 2-1.

Sequences 3 & 4 - the next two sequences are ATWS events; the probability of reactor

?

protection system failure (e.g., failure of control rods to insert) is about 1E-5. Sequence 3 involves failure of SLC. Thus, SLC pipe failure on demand (transient and failure of rods to

. insert) will be a " Medium" consequence with a CCDP of approximately 1E-5. Sequence 4 involves feedwater pump trip failure which is not expected to be impacted by class 1 pipe failures analyzed in this analysis.

4 l

5 w_______-___-___-____________

i i

Page 24 Calculation No. YC-386, Rev 1 Sequence 5 - involves a turbine trip initiating event and subsequent failure of feedwater, I RCIC, HPCI, and emergency depressurization. The frequency of this scenarios l (approximately 1E-7/yr) and the initiating frequency (1.5/yr) confirms the " Low" l consequence importance of transient initiators (feedwater and main condenser initially i available)in Table 2-1.

Sequence 6 - another ATWS event involving failure of ADS inhibit which is not expected to be impacted by class 1 pipe failures analyzed in this analysis.

Sequences 7 through 10 -involve support system (AC and DC) initiating events which are not expected to be impacted by class 1 pipe failures analyzed in this analysis.

The binning of core damage sequences (Reference 2, Table 3.4.2 and Section 3.1.5) was also reviewed for insights. For example, the "IIIB" bin is totally based on medium LOCAs where feedwater, HPCI, and emergency depressurization have failed. The frequency of this bin dhided by the medium LOCA initiating frequency indicates that CCDP for the high pressure injection and/or depressurization function is a " Medium" for medium LOCAs. Table 3-1 summarizes CCDP contributions by critical safety functions for key initiating events in the analysis. These safety functions are described further in the next section.

3.2 Safety Functions Each critical safety functica is considered when determining the number of available mitigating trains and/or estimating CCDP in the consequence evaluation. Note that applying CCDPs from

Table 2-1 will account for these critical safety functions, as they are included in the VY PRA.

Table 3-1 also summarizes how safety function failures contribute to the CCDPs in Table 2-1.

Figure 3-1 summarizes the VY PRA success criteria (Reference 2, Section 3.1.2) for loss of coolant accidents (LOCAs) in simplified diagrams. Based on a review of VY PRA results, including Tables 2-1 and 3-1, the following summarizes how these functions and others are treated in the consequence evaluation:

a l

Reactivity Control - this function is required immediately upon demand to protect the core. !

However, a pipe failure is judged more likely to cause a reactor trip than to prevent a reactor j

protection system (RPS) success (this function is fail safe, de-energize to actuate). This is particularly true for Class 1 piping located inside the drywell. Also,it is judged unlikely that

, a pipe failure could immediately impact recirculation pump trip (RPT) and alternate rod l~ insertion (ARI) functions simultaneous with RPS. Independent failure to SCRAM unavailability on the order of IE-5 (top event CR in Table 3-2) isjudged to envelope other

potential spatial causes. These judgments are based upon the fact that the RPS is safety l }, related and must function during design basis accidents (e.g., LOCAs inside the drywell).

This 1E-5 probability results in a medium consequence without considering any other 3

mitigating capability'or the frequency of challenging RPS. The following explains how this

function is treated in the analysis

l

I

]

Page 25 Calculation No. YC-386, Rev 1

+ If the pipe failure causes a LLOCA or MLOCA, a " Medium" CCDP is used based on Table 3-1. The IE-5 value for SCRAM failure is binned to core damage with a high likelihood of containment failure. Because this probability provides margin (0.1), the

" Medium" consequence is retained.

+ For pipe failures that cause a SLOCA or transient with feedwater and the main condenser initially available (SLOCA and T in Tables 2-1 and 3-1), a " Low" consequence is used because mitigating capabilities exist to reduce CCDP to <1E-6 with margin on j containment performance.

+ If pipe failure causes a TMS (MSIV closure) or TFWMS (loss of feedwater and main condenser), a " Medium" CCDP is assumed based on Tables 2-1, 3-1, and containment performance. There is s,ufficient margin to retain this " Medium" when considering containment perfonnance.

]~ + ' If the SLC pipe fails on demand, a " Medium" CCDP is used based on the IE-5 value and

~

the fact that SLC is unavailable to mitigate the scram failure. As with MLOCA and LLOCA, a 0.1 margin for containment performance allows the " Medium" consequence to be retained.

+ For pipe failures that occur during an independent demand on LPCI and core spray systems, a " Low" consequence is used because the frequency of challenge (e.g., core '

spray) in combination with RPS failure and mitigation failure is well <1E-6 with margin on containment performance.

l

' + Table 3-1 also includes initiators with HPCI or RCIC unavailable which were identified as impacts in Section 4. These clearly fall into the " Medium" category when containment performance is considered.

Vapor Suppression - this function is required early after a LOCA initiating event to protect containment from early over pressure failure. The probability of vapor suppression failure in i the VY PRA is 1.1E-4 (top event VS in Table 3-2). The following explains how this function

is treated in the analysis

p + If pipe failure causes a LLOCA, a "High" CCDP,is assigned since there is little time for additional mitigation. The VY PRA does not credit additional mitigation (Reference 2, Section 3.1.2.1). Although IE-4 in Table 3-1 is close to the medium CCDP, the potential Lt for early containment failure is considered important enough to retain the high I Consequence.

+ If pipe failure causes a MLOCA, a " Medium" CCDP is assigned based on the VY PRA j Q ! (Table 3-1) which credits emergency depressurization as mitigating vapor suppression 1 failure (Reference 2, Section 3.1.2.2). CCDP is about 1E-5 which provides margin to

^

. retain the medium consequence when the potential for early containment failure is considered.

+ If pipe failure causes a SLOCA, a " Low" CCDP is assigned. As shown in Table 3-1, CCDP is less than IE-6 because both emergency depressurization and drywell spray are credited as mitigating vapor suppression failure (Reference 2, Section 3.1.2.3). Although there appears to be insufficient margin for containment performance, for qualitative i i

l

Page 26 Calculation No. YC-386, Rev 1 .

I purposes, this analysis retains the " Low" consequence because relative importance, in comparison Medium and Large LOCA, is low.

+ For all other cases, a " Low" CCDP is assigned based on the VY PRA (Reference 2, Section 3.1.2.4) because the CCDP for vapor suppression is less than for SLOCA.

High Pressure Makeup and/or Depressurization - by definition, this function is not required for large LOCAs. For medium LOCAs, it is only required until the RPV depressurizes enough for low pressure injection. However, in the case of small LOCAs and transients, feedwater, RCIC, and HPCI are considered redundant to the low pressure makeup function. RPV depressurization is needed for low pressure makeup, if these high pressure makeup systems are unavailable. Based on the VY PRA (Tables 2-1 and 3-1), a CCDP in the

" Medium" range occurs fm medium LOCA (MLOCA), loss of feedwater (TFWMS), and loss of support system initi.,. tors. There is sufficient margin to retain this ensequence when containment performance is considered. Only small LOCAs (SLOCA) and transients with feedwater initially available (T and TMS) can reach the ' Low" CCDP. There is sufficient margin for containment peformance based on the discussion below.

j CCDP was estimated for certain transients with RCIC or HPCI unavailable to support the l analysis (see Table 3-1). As shown, these are a " Medium" consequence. With regard to containment performance, the margin appears insufficient to retain a " Medium" consequence for TFWMS in Table 3-1. However, based on a combination of margin in Table 3-1 (0.45) times the conditional probability of"Early" release (0.26 = [2.25E-7 + 1.65E-7]/1.5E-6 based

, on Reference 2, Table 4.6.2, Class IA sequences), the margin is close to 0.1 and the l consequence is not increased. Note the conditional probability of "Early/High" release is less than 0.1 using Reference 2 (1.65E-7/1.5E-6 times 0.45 = 0.05).

Low Pressure Makeup - there are several makeup trains, however, common cause limits the unavailability of LPCI and CS. For example, failure of the low pressure permissive in the VY PRA is >1E-4 (top event PIin Table 3-2), guaranteeing a "High" CCDP for large LOCA since there is limited time for operator response; no time is credited for large LOCA. In the case of medium LOCA, recovery of condensate is credited which results in a " Medium"

, CCDP with margin for containment performance. This function provides a " Low" CCDP for i small LOCA and general transients; there is either margin for containment perfonnance and/or the initiator is already Medium'due to another safety function. For SLOCA, the conditional probr.bility of "Early/High" release is close to 0.1 using Reference 2, Table 4.6.2, f Class ID sequences (1.34E-7/3.91E-7 times 0.4 = 0.14).

Heat Removal - The RHR system in the torus cooling mode of operation and the hardened containment vent provide reliable containment heat removal capabilities. The containment i vent is not dependent on support systems, except for controlling depressurization and 9

alternate injection when necessary. Local recovery of this equipment is possible and there is significant time available to perform these actions. With the exception oflarge and medium

, LOCAs, these capabilities provide CCDP values <1E-6. Given the time available to recover this function and the obvious domination of other functions, containment heat removal is not P

Page 27 Calculation No. YC-386, Rev 1 considered important in this analysis. Also, because loss of this function does cause core damage to occur late, containment performance is not considered.

. Containment Performance - To maintain the consequence category determined from the above functions, at least one containment barrier must be available or there must be margin in the number of available mitigating trains as described in Section 2. Otherwise, the consequence category is adjusted accordingly.

Table 3-2 summarizes unavailability for key functions, systems, and trains required to support the critical safety functions shown in Figure 3-1. Table 3-2 also provides the number of backup trains assumed in the analysis when using Tables 2-2 and 2-3. As explained in Section 2, I train s 0.01 unavailability,2 trains a 1E-4 unavailability, and etc. Also, a 0.5 train s 0.1 unavailability.

The unavailability's in Table 3-2 generally assume all support systems are available. Support system trains are generally more reliable, but may impact multiple systems. They are included in the analysis and their impact on system unavailability is considered by quantifying the PRA for those cases where Tables 2-2 and 2-3 apply.

3.3 Plant Level Assumptions l Engineering judgments are included and discussed throughout the analysis; the following are considered to be key plant level assumptions andjudgments:

1. The pipe failure can occur at anytime; three configurations are defined in Section 4. These are l normal (operating or standby), test, and accident demand. Section 4 also summarizes judgments and assumptions regarding which configurations are most important. If the pipe failure does not cause a direct initiating event, it is assumed that the pipe failure occurs during the accident demand configuration, if applicable. This assumes pipe failure occurs during the most conservative exposure time and accounts for the higher stress placed on the N operators with resultant delay in operator response.

2. Leak-before-break is not credited in the analysis. This is acknowledged as an option in the l Code Case (Reference 1) which references NUREG-1061, Volume 3.

3. Large LOCA initiators in the VY PRA assume that one LPCI injection path is unavailable g due to the initiating LOCA being assumed to occur in the recirculation discharge piping

(Refere.nce 2, Section 3.1.2). No credit was taken for the piping upstream of the discharge block valees potentially being isolated on low reactor pressure. Although this is potentially a conservative, the same low pressure permissive that closes the block valves is also required J to open the LPCIinjection valves. Since failure of the low pressure permissive or vapor suppression function are greater than IE-4, all large LOCA initiating events will have a

{ "High" CCDP whether block valve closure is credited or not. Still, it should be noted that these common cause failure modes could be over estimated and conservative.

j 4. Interfacing system LOCA (ISLOCA) initiators between the LPCI injection MOVs (MOV27A/B and MOV25A/B) are potentially isolable with the MOV25 valve. Consistent 1

1 l

l Page 28 Calculation No. YC-386, Rev 1 with the VY PRA, this was not credited in the analysis due to proximity of break, environment, interlock,'and blowdown loads (Reference 2, Section 3.1.3 and 3.2.36).

5. Assigning large LOCAs to the "High" consequence is potentially conservative. Conditional core damage probability is dominated by the probability of common cause failure of the low pressure ECCS permissive and vapor suppression failure (Tables 3-1 and 3-2). Both of these pababilities could be conservatively estimated.

l

6. Lo t, pressure ECCS is credited for steam line breaks in the steam tunnel (Reference 2, Section 3.1.3.2). This results in a " Medium" consequence. There appears to be no analysis for steam line breaks with failure to isolate, however, this judgment appears reasonable based on a review of the spatial configuration and considering that ECCS equipment is safety related. '

In the case of feedwater line breaks in the steam tunnel, little credit is given for makeup, resulting in a "High" cons ~equence. This may be conservative. l

7. The pipe size assumed in the analysis (Section 4.2) for the case where the reactor depressurizes, allowing low pressure ECCS makeup without the aid of HPCI, feedwater, or ADS isjudged to be conservative. Additional analysis may indicate that some piping leads to a medium LOCA instead of a large LOCA; this would reduce the imponance of such piping.

8. The calculation does not penalize reliability of the inside drywell MOV (HPCI, RCIC, and Reactor Water Cleanup), given failure of piping between the MOV and the drywell wall. This is based on a review ofMOV qualifications (Reference 18) and recognition that the MOVs would close in a much'shoner time period than their qualification. The likelihood that impingement from the reactor blowdown impacts valve reliability in this short time period is also considered unlikely. Even if unreliability was increased to IE-2, the present pipe f' segment rankings would not increase.

l 9. Input & Modeling Uncertainties - The methodology (Reference 1) is an order of magnitude approach to ranking the consequences of pipe failure; the absolute numerical value is j considered an order of magnitude esumate. Many uncertainties would have the same relative !

I impact on ranking and others arejudged to be small in comparison to pipe failure probability (a probability of 1.0 is assumed in this analysis. Also, see Reference 15). The pipe degradation evaluation (qualitatively addresses pipe failure probability) is needed for risk f+ ranking.

, 10. The RISKMAN computer software (Reference 14) was used to quantify accident sequences ,

and estimate conditional core damage frequency. The Code has the capability to perform data j l- analysis, systems, analysis, accident sequence analysis, spatial & extemal analysis,

_ uncertainty analysis, and importance analysis. Only the accident sequence analysis module l was used in this calculation. The relevant inputs (master frequency file) and outputs (binning

~

results) are provided in Attachment A. The VY PRA (Reference 2) event trees and rules were ;

. used. The results were , checked for reasonableness, including comparison with other Boiling l

{ Water Reactor results. Although RISKMAN has not been verified per YNSD Engineering i Instruction WE-108, it is considered acceptable for use in support of this calculation for the t following reasons: (a) this fundamental (event tree) module of RISKMAN has been and

,I continues to be used by many RISKMAN users in performance of probabilistic risk assessments, which have undergone many reviews, including Nuclear Regulatory

'3

Page 29 Calculation No. YC-386, Rev 1 I

(

Commission reviews and internal reviews; (b) specific results applicable to this calculation !

have been reviewed for reasonableness in supponing the results of this calculation. I

11. Although the VY PRA (Reference 2) is not a YNSD Engineering Instmetion WE-103 approved calculadon, it has received significant technical review, both internal and extemal.

In addition, the applicable results of the VY PRA used in this application were checked for reasonableness and compared to other Boiling Water Reactor PRA results. It is considered an acceptable input for the purposes of this calculation.

12. Other non WE-103 qualified references (References 3,9, and 11), were also reviewed for reasonableness and foed to be acceptable inputs for the purpose of this calculation.

13. YNSD Engineering Instruction WE-100, Table I was considered. This calculation does not cause or effect plant design change. This calculation may be used to suppon a future change the VY Inservice Inspection Program, at which time impacts must be considered.

14. Outstanding YNSD condition reports (CRs) and engineering discrepancy reports (EDRs) were reviewed and have no affect on this calculation.

15. There are no Nuclear Regulatory Commission Safety Evaluation Reports (SER) compliance issues sine there is no SER was written on the VY PRA or this calculation.

16. There is no applicable methods overview memorandum (MOM) that applies to this calculation.

17. The following items can be considered potential 1 imitation with regard to analysis assumptions, but they are not judged to have a significant impact:

A plant walkdown has not been conducted, however, spatial assumptions were discussed with and reviewed by the engineers responsible for intemal hazards analysis.

The seismic analysis in support of A-46 and IPEEE is not completed, however, this is not expected to impact this analysis of class 1 piping as described in Section 4.

!

Similarly, fire and flood analyses in support of IPEEE are not completed, however, this is not expected to impact this analysis of class 1 piping as described in Section 4.

l b

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BVY-98-114, Provides Response to Request for Addl Info Re Implementation of ASME Code N560 at Plant

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