ML20141H110
| ML20141H110 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 07/11/1997 |
| From: | Reid D VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20141H113 | List: |
| References | |
| BVY-97-90, NUDOCS 9707220014 | |
| Download: ML20141H110 (8) | |
Text
e o
s VERMONT YANKEE NUCLEAR POWER CORPORATION Ferry Road, Brattleboro, VT 05301-7002 ENGINEERING OFFICE
\\
580 MAIN STREET BOLTON, MA 01740 (508)779 6711 July 11,1997 BW 97-90 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
References:
(a)
License No. DPR-28 (Docket No. 50-271)
(b) 10CFR50 Appendix J, Option B (c)
Reg Guide 1.163, " Performance-Based Containment Leak Test Program", dated September 1995 (d)
NEl 94-01 " Industry Guideline for implementing Performance-Based Option of 10CFR Part 50, Appendix J", dated July 26,1995 (e)
ANSI /ANS-56.81994, " Containment System Leakage Testing Requirements" (f)
Letter; Christopher 1. Grimes (NRC) to David Modeen (NEI), Option B Technical Specification Submittal Model, dated November 2,1995 (g)
Letter, WNPC to USNRC, Proposed Change No. 46; Torus /Drywell Differential Pressure, WW 78-46, dated May 16,1978 (h)
Letter, WNPC to USNRC, Operation of Purge and Vent Valves Under Inerting Conditions, FW 82-32, dated March 26,1982 (i)
Letter, USNRC to WNPC, Operation of Purge and Vent Valves Under Inerted Conditions, NW 82-60, dated May 3,1982
Subject:
Technical Specification Proposed Change No.190,10CFR',0 Appendix J, Option B Pursuant to 10CFR50.90 of the Commission's Rules and Regulatioru, Vermont Yankee Nuclear Power Corporation (WNPC) proposes the following changes to Appendia A of the Facility Operating License
[ Reference (a)].
Aor7 {/
Proposed Chanoe This proposed change is provided to incorporate Option B of 10CFR50 Appendix J and editorial changes to TS Table 4.7.2. Specifically we propose to replace pages 147,156 through 161,168, and 279 of the Vurmont Yankee Technical Specifications with the attached pages 147,156 through 161, 168, and 279.
9707220014 970711 PDR ADOCK 05000271 P
PDR 30040 llllllNlR[Hl,1llllll
9 t
VERMONT YANKEE NUCLEAR POWER CORPORATION United Stit::s Nucitar Regulatory Commission July 11,1997 Page 2 of 8 Reason / Basis for Chance
. Option B
'In order to implement Option B to 10CFR50 Appendix J, it is necessary to make changes to Technical Specifications 3.7/4.7, CONTAINMENT SYSTEMS, and their associated BASES. In addition, a description of Vermont Yankee's Primary Containment Leakage Rate Testing Program (PCLRTP) will be added to the Administrative Contro!a Section (6.0) of the Technical Specifications. To accomplish the implementation of Option B, the prescriptive requirements relating to Option A of 10CFR50 Appendix J and components subject to Appendix J testing requirements will be relocated to the PCLRTP. The testing intervals for the containment system and for the components that penetrate the primary containment, under Appendix J, Option B, will be performance-based. Therefore, the requirements for determining the appropriate testing intental will be contained in the PCLRTP.
The proposed Technical Specification changes are consistent with the guidance provided in Reference
- (f).
Section 4.7.A.3 Technical Specification Amendment No. 2 added this paragraph for certain systems which were
- considered to be extensions of Primary Containment. The referenced Table (4.7.2.b) lists a number
. of valves that were not leak tested. The conservative wording of this section to require two Type C tested valves is being replaced by the wording from the Standard Technical Specifications and relocated to the LCO section because:
.I a.
the requirement to perform action to maintain Technical Specification operability is more appropriately defined as a Limiting Condition for Operation (LCO) instead of a Surveillance j
Requirement, b.
the system that the amendment was based on (RHR) is now being leak rate tested based on the change to the design bases for RHR and Core Spray, and c.
the listing and bases for primary containment isolation valve testing will be contained in the l
PCLRTP.
' Option B implementation VY's PCLRTP will follow Regulatory Guide (RG) 1.163 (Reference c), including the four exceptions specified under Paragraph C Regulatory Position. This endorses, with certain exceptions, an industry guide, NEl 94-01 [ Reference (d)), and Industry standard, ANSl/ANS 56.8-1994 [ Reference (e)}.
Table 4.7.2 Editorial Chance in Reference (g) Vermont Yankee proposed Technical Specification changes to establish requirements for drywell to suppression chamber differential pressure control and suppression pool water level. As part of our submittal we characterized a Pumpback System as the expected means for maintaining this differential pressure and proposed " normal" valve positions in TS Table 4.7.2 that would be typical for operation of this system. Also addressed in this submittal were means for maintaining the required differential pressure with alternate valve line ups in the event the Pumpback System were unavailable.
Subsequently, use of the Pumpback System was discontinued due to reliability problems and the alternate means of maintaining differential pressure utilized. The alternate means of maintaining
m.
"e:
4 VERMONT YANKEE NUCLEAR POWER CORPORATION Unitid States Nucisar Rigulatory Ccmmission July 11,1997 Page 3 of 8 differential pressure involved changing the normal positions of four primary containment isolation valves on the inlet to the contalnment from those listed in Table 4.7,2. The containment vent and purge valve
]
positions for several operational modes (including the one discussed above) were docketed in 1-Reference (h) in preparation for containment inerting' operation. Reference (1) concluded that the
?
' Reference (h) submittal describing operation of vent and purge valves under inerted conditions was acceptable. TS Table 4.7.2 descriptions were not changed however, and continued to list " normal" valve positions based upon use of the Pumpback System and not the more acceptable method for maintaining differential pressure with an inerted containment. This part of the change is considered editorial in nature in that it updates the Information in Table 4.7.2 to current operational practices, as j
described in Reference (g) and approved in Reference (1).
i
]
Safety Considerations i
Option B i
Regulatory Guide 1.163, " Performance-Based Containment Leak Test Program," dated September j.
1995, provides specific guidance concerning a performance-based containment leakage test program, j
acceptable leakage rate test methods, procedures and analyses that may be used to satisfy the requirements of Option B to' Appendix J of 10CFR50. These proposed Technical Specification changes are, and the PCLRTP will be, consistent with this Regulatory Guide.
l The testing requirements of Option B to Appendix J of 10CFR50, and the PCLRTP for the Vermont j.
Yankee Nuclear Power Plant, will provide reasonable assurance that leakage through the primary i
containment and components penetrating the primary containment will not exceed allowable leakage l
. rates specified in the Technical Specifications and that the integrity of the contalnment structure is maintained during its service life, i
l For these reasons, there is reasonable assurance that the changes that would be authorized by the i
proposed change can be implemented without endangering the health and safety of the public and are j
consistent with its common defense and security.
Section 4.7.A.3 Ii Closed de-activated automatic valves, closed manual valves or blind flanges which are used to satisfy the compensatory measures of 4.7.A.3 are primary containment isolation devices and will be leak tested in accordance with the PCLRTP. In addition, the Technical Specification establishes these devices as an isolation barrier that cannot be adversely affected by a single active failure. As a result, any reduction in a margin of safety will be insignificant and offset by the benefit gained with equivalent compensatory measures to ensure the primary containment boundary is maintained, which reduces
- unnecessary plant shutdown transients.
Table 4.7.2 Editorial Change This editorial change will revise the " normal" valve positions listed for four valves that are used to control containment inerting and drywell/ suppression chamber differential pressure. All four valves receive close signals from PCIS logic. The two valves currently described as normally open are 6" valve 1619-23 and 18" valve 1619-8. Both of these valves are maintained normally closed and will be so listed in the Table. The two valves currently described as normally closed are 1" valves 16-20-20 and 16-20-228. Both of these valves are maintained normally open to provide the nitrogen makeup supply for the containment overpressure and will be listed as such in the Table. The design maximum
i e
VERMONT YANKEE NUCLEAR POWER CORPORATION United Stat:s Nuctrar Regulatory Commission July 11,1997 Page 4 of 8 operating time for these 1" target rock valves will also be added to the TS Table for Information. This editorial change is more conservative in that the larger, slower closing valves will now be specified as normally closed and much smaller, faster valves will be specified as normally open. These valves are also qualified and tested in accordance with our Inservice Testing and Appendix J Programs.
This proposed change has been reviewed by the Vermont Yankee Plant Operations Review Committee and the Vermont Yankee Nuclear Safety Audit and Review Committee and found to be acceptable.
Sianificant Hazards Considerations The standards used to arrive at a determination that a request for amendment involves no significant l
hazards considerations are included in the Commission's regulations,10CFR50.92, which state that the operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accider;t previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. In addition, the Commission has provided i
guidance in the practical application of these criteria in 51FR7751, dated March 6,1986.
The following discussion addresses the proposed changes with respect to each of these three criteria and demonstrates that the proposed changes do not constitute a significant hazard consideration:
Option B 1.
Does the change involve a significant increase in the probabl!!!y or consequences of an accident previously evaluated?
The proposed chango does not involve a change to the plant design or operation. As a result, the proposed change does not affect any of the parameters or conditions that contribute to initiation of any accidents previously evaluated. Thus, the proposed change cannot increase the probability of any accident previously evaluated.
The proposed change potentially affects the leak-tight integrity of the containment structure designed to mitigate the consequences of a loss-of-coolant accident (LOCA). The function of the containment is to maintain functional integrity during and following the peak transient pressures and temperatures which result from any LOCA. The containment is designed to limit fission product leakage following the design basis LOCA. Because the proposed change does not alter the plant design or test method, only the frequency of measuring Type A, B and C leakage, the proposed change does not directly result in an increase in containment leakage.
I However, decreasing the test frequency can increase the probability that an increase in l
containment leakage could go undetected for an extended period of time. Based upon the results of the periodic containment Type A or integrated Leak Rate Tests (ILRTs) and Type B and C or Local Leak Rate Tests (LLRTs) surveillance tests, this is not expected during the remaining life of the plant. The risk resulting from the proposed changes is as follows:
Tvoe A Testino i
NUREG/CR-4330 (NRC86) found that the effeet of containment leakage on overall accident risk is small since risk is dominated by accident sequences that result in failure or bypass of the containment. It is also determined that on an expected individual dose basis, the effect of containment leakage is small.
_.y
. _ _. _. _ _ _. ~.
_._..._m..
m 13 VEr WONT YANKEE NUCLEAR POWER CORPORATION United Statis Nuclear Regulatory Commissic..
July 11,1997 Page 5 of 8 l
Industry wide,.lLRTs have only found a small fraction of the leaks that exceed current:
acceptance criteria. Only three percent of all leaks are detected by ILRTs, and therefore, by
{
extending Type A testing intervals, only three percent of allleaks have a potential for rema:ning undetected for longer periods of time. In addition, when leakage has been detected by ILRTs, the leakage rate has been only about two times the allowable leakage rate, i
1 F
observations, together with the insensitivity of reactor accident risk to the containment leakage NUREG 1493, " Performance-Based Containment Leakage Test Program", found that these i
rate, show that reducing the Type A leakage test frequency would have a minimal impact on i
public risk.
l
?'
Tvoe B and C Testina 5
. NUREG-1493 found that while Type B and C tests can identify the vast majority (greater than 95 percent) of all potentialleakage paths, performance-based alternatives are feasible without significant risk impacts. The risk model used in NUREG-1493 suggests that the number of components tested would be reduced by about 60 percent with less than a three-fold increase in the incremental risk due to containment leakage. Since, under existing requirements, leakage contributes less than 0.1 percent of overall accident risk, the overall impact is very small. NUclEG 1493 found that while the extended testing intervals for Type B and C tests led to minor i 'reases in potential offsite dose consequences the actual decrease of on-site (worker) as would be reduced in proportion to the number of Type B or C tests not perform 6 EPRI Rest Arch Project Report TR 104285, " Risk Impact Assessment of Revised Containment Leak Rate Testing intervals," also concluded that a relaxation of the test intervals for Type B and C penetrations results in a negligible increase in total plant risk.
4 Based on the above VYNPC has concluded that the proposed change will not result in a significant increase in the probability or consequences of any accident previously evaluated.
2.
Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
The proposed change does not involve a change to the plant design or operation. As a result, the proposed change do::s not affect any of the parameters or conditions that could contribute to initiation of any accidents. This change involves the reduction in Type A, B, and C test
- frequency. The methods of performing the tests are not changed. No new accident modes are created by extending the testing intervals. No safety-related equipment or safety functions are altered as a result of this change. Extending the test frequency has no influence over nor does it contribute to, the possibility of a new or different kind of accident or malfunction from those previously analyzed.
Based upon the above, VYNPC has concluded that the proposed change will not create the possibility of a new or different kind of accident from those'previously evaluated.'
3.
Does this change involve a significant reduction in a margin of safety?
As stated in the Technical Support Document (TSD) for the NRC's Option B to Appendix J rule change, NUREG 1493 concludes a reduction in the frequency of Type A testing from the current three per ten years to one per ten years leads to an Imperceptible increase in risk.
VERMONT YANKEE NUCLEAR POWER CORPORATION United Statis Nuclear Rigulatory Commission July 11,1997 i
l Page 6 of 8.
j It also conclu' es th'at a reduction in the frequency of Type B testing of electrical penetrations d
should be possible with no adverse impact on risk. A vast majority of leakage paths are identified by Type C testing of containment isolation valves and, based on the model of -
component failure with time, performance based alternatives to the current Type C testing intervals are feasible without significant risk impacts.
As a result, VYNPC has concluded that the proposed change will not result in a significant
' reduction in the margin of safety.
4.7 A.3
-1.
Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
)
The poposed change does not result in any hardware or operating procedure changes.
Closed and de-activated automatic valves, closed manual valves or blind flanges that serve as primary containment isolation valves are not assumed to be initiators of any analyzed event.
The role of these devices is to isolate containment during analyzed events, thereby limiting I
consequences. The change establishes compensatory measures using closed and de-activated automatic valves, closed manual valves or blind flanges as an isolation barrier which is equivalent to those already included in the current. Technical Specifications. The proposed change does not introduce any new failure modes, such that a single active failure could allow I
a primary containment release through an un-isolated path. Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.
j 2.
Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
This change does not result in any changes to equipment design or capabilitles or the operation of the plant. The change still ensures the primary containment boundary is maintained. Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
Does this change invoNe a significant reduction in a margin of safety?
Closed and de-activated automatic valves, closed manual valves or blind flanges which are used to satisfy the compensatory measures of 4.7.A.3 are primary containment isolation devices will be leak tested per the PCLRTP.
In addition, the Technical Specification establishes these devices as an isolation barrier that cannot be adversely affected by a single active failure. As a result, any reduction in a margin of safety will be insign!ficant and offset by '.he benefit gained with equivalent compensatory measures to ensure the primary containment boundary is maintained, which reduces unnecessary plant shutdown transients.
Table 4.7.2 Editorial Change 1.
. Does the change involve a significant increase in the probability or consequences of an accident previously evalJated?
s_ _ _ _. _ _ _._. _ _ _ _ ~... _ _ _.. _ _. _.. -.
- y
- VERMONT YANKEE NUCLEAR POWER CORPORATION United States NuclIar Rsgulatory Commission July 11,1997:
Page 7 of 8 This change updates the information presented in this Table to reflect current practice. The methods of maintaining an inerted containment and differential pressure between the drywell and suppression pool have been previously docketed. The. valves to now be shown normally closed on the. Table are large (6" and 18") purge valves and the valves to be shown as normally open to provide makeup nitrogen are both 1" in size. The probability of an accident
+
Is not significantly increased, since the subject valves are not considered to be initiators of any 2'
. accident previously evaluated. The consequences of an accident are not significantly Increased, since each of the subject vanes receives a close signal from PCIS. In addition,
- PCIS closure of the two one inch valves will terminate the associated release pathway more
.i rapidly than the existing valve lineup reflected on the Table. Thus if is concluded that this change will not involve any significant increase in the probability or consequences of an 7
accident previously evaluated.
3 2.
Does the change create the possibility of a new or different kind of accident from amy -
previously evaluated?
l1 All four valves whose listed normal positions are proposed to be changed are PCIS valves and 1
- receive the same closing signal. All are tested in accordance with our Appendix J and IST
. programs. No changes in equipment design or operation are proposed, only the listed normal positions of the subject valves. Thus, this change will not create the possibility of a new or different kind of accident from any previously evaluated.
a j
3.
Does the change involve a significant reduction in a margin of safety?
The valves to be listed as normally open are significantly smaller and faster closing than the i
purge valves currently listed as open. Thus the change in the listed normal position of these four valves provides a more conservative initial condition than is current'y depicted in Table
}
4.7.2. No changes in equipment design or operation are proposed. Thus, it is concluded that there is no significant reduction in the margin of safety.
Based on the above, Vermont Yankee concludes that the proposed change does not constitute a
]
significant hazards consideration as defined in 10CFR50.92(c).
. Environmental Considerations VYNPC has reviewed the proposed license amendment against the criteria of 10CFR51.22 for
- environmental considerations. The proposed changes do not significantly increase the types and amounts of effluents that may be released offsite, nor significantly increase Individual or cumulative occupational radiation exposures. Based on the foregoing, VYNPC concludes that the proposed changes meet the criteria delineated in 10CFR51.22(c)(9) for a. categorical exclusion from environmental review.
Schedule of Chance Regarding our proposed schedule for this amendment, we request your review and approval to support implementation of Appendix J Option B for the next refueling outage at the Vermont Yankee Nuclear
- Power Plant, currently scheduled for March,1998. The proposed changes will be incorporated into the Vermont Yankee Technical Specifications and Primary Containment Leakage Rate Testing Program as soon as practicable following receipt of your approval.
VERMONT YANKEE NUCLEAR POWER CORPORATION United States Nuclear Regulatory Commission July 11,1997 Page 8 of 8 We trust that the information provided supports our request. However, should you have any questions in this matter, please do not hesitate to contact us.
Sincerely, VERMONT YANKEE NUCLEAR OWER CORPORATION Donald A. Reid Sr. Vice President, Operations y SMigD cc:
USNRC Region I Administrator 4
O c['
USNRC Resident inspector VYNFS C
USNRC Project Manager VYNPS NOTARY g
PUBUC-p STATE OF VERMONT
)
l" OUNU.
WINDHAM COUNTY Then personally appeared before me, Donald A. Reid, who, being duly sworn, did state that he is Sr. Vice President, Operations of Vermont Yankee Nuclear Power Corporation, that he is duly authorized to execute and file the foregoing document in the name and on the behalf of Vermont Yankee Nuclear Power Corporation, and that the statements therein are true to the best of his knowledge and belief.
i Sally A. Sangstrum, Notary Public My Commission expires February 10,1999