ML20237A579
| ML20237A579 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 08/11/1995 |
| From: | Anthony G, Brobson J, Padleokas A CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20237A565 | List: |
| References | |
| EER-92-0483, EER-92-0483-R01, EER-92-483, EER-92-483-R1, NUDOCS 9808140190 | |
| Download: ML20237A579 (31) | |
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EER Number 924483 Rev.1 0-list System F'de Number Reference Y (Al 2096 HPCI Document (s) See Attached l
THie: E41 V159 Testirg Method mange kh hb; hf hh "fh hhhhb b lk k
l Se6ende Roseted?
No EQ Affected?
No MOV per ENP-667 No Fonow NED sesamic Des gn Guides:
Foaow PLP 02: Al 71 EO Review; Consutt NED Mech. or Component C
NED management ecorovst ENP-34.1. Form 3 ef requeed Eng. to address GL 89-10 I
cp Use As Is?
No Parte Upgrade?
No Short Term Structursi No l
Follow PMC 15.6 bitogrity?
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Expiration Date s J cn Permanent Repub?
No Tenperary Repair?
No Temporary Mod 6 cation?
No n
PLP-08 if 1S1:
Expiraten Date Follow PLP-22
- .N Form 6 for Dwgcoc changes PLP48 if ISI: Notify Temp Cond Coord Notify Temp Cond Coord 60.69 Reqidred?
Yes FSAR Affected?
No Operabitty Assessment?
No Complete Al-109 Safety Review; PNSC Two toch revwws vf Class A; Complete within 0104 time frame;
& RCI-2.1 if unrevwwed safety RCl4.1 form: Corp. NSRG epprovat T/S management approval l*
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Docuraent(6$ See Attached Titse: E41-V1EE9 Teodng Method Change k*
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FoRow PLP C2: Al-71 EQ Review; Consutt NED Mech. or Component NED management approvet ENP 34.1, Form 3 if requesd Eng. to address GL 8910 cp Use As Ie?
No Ports Upgrade?
No Short Term Structural No Fonow PMC 15.6 Inteytty?
.O Expiration Date CD Permanent Repair?
No Temporary Repair?
No Temporary Mod 6cetion?
No m
PLP-08 If ISI:
Empiroten Date Follow PLP-22
- h Form 6 for Dwg/ Doc chenpas PLPM if ISI: Notify Temp Cond Coord Notify Temp Cond Coord 50.59 Reedred?
Yes FSAR Affected?
No Operatety Assessment?
No Complete Al 109 Safety Rowew: PNSC Two toch reviews if Oass A:
Complete within OlN time freme;
& RCI-2.1 is unrevieW safety RCl4.1 form: Corp. NSRG approval T/S management opproval E?hhlhkkf hN bY YN ibb Y.Y b
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EER 92-0483 Rsv 1 l
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PRO 8LEM DESCRIPTION During the 1992 forced outages, PT-20.12 was performed unsat on both the Unit 1 l
l and Unit 2 E41-V159 HPCIinjection check valvss. This PT uses the mechanical l
exerciser to verify the valve will open and close with no binding as required to meet l
ASME Section XI requirements. A review of the actual configuration and the test i
indicated that the test met the ASME Section XI requirements but the procedure, as written, would not fully verify that the valve is functional. Because of the desire to I
have the surveillance test verify valve function a more positive fashion, a test relying
)
on partially disassembly was required to be used as allowed by GL 89-04 in Rev 0 of this EER.
Rev 1 to this EER documents findings of subsequent check valve disassembles and l
identifies differences between the check valve design and the assumptions made in l
writing OPT-20.12 which lead to the unsatisfactory PT results in 1992.
1 EVALUATION / DISPOSITION j
The E41-V159 is a Anchor Darling 14" 900# carbon steel exercisable swing check valve located in the MSIV pit. This valve is shown on FP-82680. The upstream half of the valve is exposed to high temperature feedwater at approximately 1000 psig during normal plant operation. It has a lever pinned to an actuator shaft that provides the ability to manually exercise the valve disc. The actuator shaft is 1.5" in diameter l
st the stuffing box. This actuator shaft also serves to function as a hinge pin on one l
side of the valve. On the other side of the valve an indicator shaft is installed. This indicator sh aft is pinned to the swing arm providing an external position indication of l
swing arm rotation. The indicator shaft also serves to function as a hinge pin on the i
side of the valve opposite to the actuator shaft. The indicator shaft is 0.5" in diameter where it penetrates the indicator side stuffing box.
PT-20.12 currently requires that the valve can be opened with s 134 ft-lb and that the valve returns from the open to the closed position with no assistance. Although initial post-modification testing showed that the valve would return to its closed position without assistance, packing adjustments over time have increased the i
packing drag on both the indicator shaft and actuator shaft to where torque must be applied to the actuator handle to restore the disc to the closed position. Provided that no binding exists between the actuator' shaft and swing arm, packing drag on the actuator shaft is insignificant to valve performance. Packing drag on the indicator shaft, however has a direct affect on the freedom of movement of the valve disc and needs to be considered when discussing failure modes.
ASME Section XI IWV-3522 discusses the method derived for obtaining the acceptance criteria for acceptable opening torque. A nominal value of 200% of the torque measured to open the valve when the valve was new and in good operating condition is applied for the acceptance criteria.
EER 92-0483 Rav 1 Page 3 of 12 Because of the various problems encountered during the performance of this PT, the test method was carefully reviewed. At this time it was identified that the PT did not test for a potential valve failure mode. Since the actuator shaft rotates with the swing arm while the vcive is allowed to close, the procedure did not assure that a combination of indicator shaft fri; tion and potential binding between the actuator shaft and the swing arm does not exceed the moment created by the weight on the disc. Excessive friction could potentially hold the disc open. Although testing for this failure mode is not required by Section XI of the code, procedural enhancements to OPT-20.12 and/ or disassembly and inspection in accordance with Generic Letter 89-04 can detect this failure mode. The Generic Letter allows partial disassembly inspections to be performed once per cycle in lieu of the testing described in ASME Section XI.
The significance of swing arm to hinge pin binding has been evaluated with respect to the opening function of the valve. The discharge pressure of HPCI in the min flow mode while in Auto control has been observed at 2: 1500 psig on numerous past auto initiations. The effect of elsvation reduces the dP across the E41-V159 by approximately 25 psig. Therefore, the available dP to open the valve is approximately 1g00 - 1125 - 25 = 350 psig. The exposed area of the disc is approximately 113 in. Therefore, the available force is approximately 39000 lb generating > 30,000 ft-Ib of torque at the hinge point. Due to the physicallimitations of the actuator, the maximum applied force for this test is much less that the available force to open the valve. Therefore, if the valve can be opened at all with the actuator lever, it is an adequate demonstration of the ability of the valve to open.
Since these valves have been installed, during some of the plant runs between the l
tests, HPCI has injected into the vessel. Since the valves have been found in the closed position at the beginning of the tests, this indicates that the valves have opened and closed reliably. Additionally the satisfactory performance of the test after maintenance on the actuator shaft packing, have demonstrated that there has been no significant binding of the indicator shaft and no disc / swing arm to body type of interference.
Disassembly inspections performed on 1 E41-V159 during B110R1 and 2 E41-V159 during B211R1 found the valvesin an acceptable state. OPT 20.12 was performed coincidentally with these disassembles to provide assurance that the proposed test methodology will detect failure modes for the valve.
As per the above discussion, OPT-20.12" test methodology requires revision. The following sequence of events should be proceduralized in order that the identified failure mode can be detected.
1.
Verify no dP exists across the valve disc. (System drained) 2.
Rotate actuator handle counterclockwise by hand until actuator shaft engages swing arm. During this rotation, observe the indicator to ensure that the indicator remains at approximately the sarne poset>on. This observation will l
assure that the actuator shaft is not bound to the swing arm. (Some slight movement of the indicator may be observed due to mechanical " play", however a one to one movement of octuator handle to ind4cator is unacceptable.)
3.
Attach a right handed torque wrt:nch to the nut at the end of the actuator shaft f and rotate the actuator shaft clockwise. Measure ervJ record the torque required to overcome the pockmg fnetson on the actuster shaft.
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b EER 92-0483 Rav 1 Pags 4 of 12 4.
Attach a left-handed torque wrench to the nut at the end of the actuator shaft and rotate the actuator shaft counterclockwise. Measure and record the torque required to open the valve and compare this value to the acceptance criteria.
5.
Release the actuator shaft. If the valve closes, the test is complete.
6.
If the valve does not closes without assistance, apply a torque in the clockwise direction less than or equal to the torque measured in step 3. If the valve closes, the test is satisfactory.
I CORRECTIVE ACTIONS I
Revise PT-20.12 to include test methodology as described.
i Revise OENP-16.7 and IST program valve tables to denote that either testing LAW I
OPT-20.12 or check valve disassembly are acceptable methods of verifying check valve exercising in accordance with ASME Section XI and applicable relief requests.
L
EER 92-0483 Rev 1 Page 5 of 12 ATTACHMENT 2 Page 55 of 69 10 CFR50.59 Program Manual Rev. 3 ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page _1_ of 1 SAFETY REVIEW COVER SHEET DOCUMENT NO. EER 92-0483 REV.NO. 1 DESCRIPTION OR TITLE: E41 Vt B9 Testina Method Chance 1.
Assigned Responsibilities:
j gg Safety Analysis Preparer:
M. IVL r AOLM IlOI SO O Lead 1st Safety Reviewer; 7-O 2nd Safety Reviewer:
Godht44 O 2.
Safety Analysis Preparer: Complete PAR TY ANALYSIS 7!)5k5 Safety Analysis Properor M
/
SIGNATURE DATE I 3.
Lead 1st Safety Reviewer: Complete Part it, item Classification.
l 4.
Lead 1st Safety Reviewer: Part 111 may be completed. If either question 1 or 2 is
- yen,* then Part IV is not require,d.
5.
Lead 1st Safety Reviewer: Deterrreine which DISCIPLINES are required for review of t"us item (including own) and mark the appropriate blockts) below.
OfSCIPLINES Reauired:
fPrint Namel Sionatirrecete (Sten 71 l l Nuclear Plant Operations l l Nucleat Engineering "T h. h@rd Tod W
7-/$48 IXl Mechanical I I Electncel
{/
l 1 Instrumentation f. Contrel l I Structural i I Metalkn 0V I l Chemistry /Radie, chemistry l ] He alth Physecs l l Admnstrative Controls l
6.
A QUALIFIED SAFETY REVIEWER will be assegned for each DISCIPLINE marked in step 5 and hes/her name printed in j
the space o ovided. Each person hsted shafi perform a SAFETY REVIEW and provide inpu into the Safety Review r
Package.
7.
The Lead 1st Safety Reviewer wel assure that a Part fit or Port IV is comp 6eted (see step 4 abovel and a Part VI sf requeed (see 9.d of Part Ill. Each person bsted in step 5 shall sgn and date next to his/her name en step 5. indicaurg comp 4 ten of a SAFETY REVtEW.
corday with Section 8.0 Pwrtarm a SAFETY REVIEW indBJThe#r O
2nd Safety Rewewer
- a'~ #
o.,e 2nd S.r,y R-e (f ' p,J J P orSe-9 PMSC avnew mwees' er *ves.* ertedi Part v ene mens reason tw ew-Y.22 ff2 e
ll IXl f I Peenemer t.PeetviEWED 5 ARTY OUESTIO*s I f hw 9 et Pwt IV erwesured *Vas*
II Ov"we M I>
cas,cip swo 3 Pogre77ofes u_______--_
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EER 92-0483 Rev 1 Page 6 of 12 ATTACHMENT 2 dage 56 of 69 10 C".oO.69 Program Manual Rev. 3 I
ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page_2_of 6 PART 1: SAFETY ANALYSIS (See instructions in Section 8.4.1) j (Attach additional sheets as necessary.)
DOCUMENT NO. EER 92-0483 REV. NO. 1 DESCRIPTION OF CHANGE:
/
The revision to this EER documents test methodology changss to OPT-20.12 so that the procedure tests the function of the valve so that the determined failure modes of the valve are challenged.
ANALYSIS:
Changes to the test methodology will assure that the valve can perform its safety functions as desenbed in the FSAR. The test methodology meets requirements of applicable codes and standards. As such there is no impact to plant safety.
REFERENCES:
Technical Specification Sections 4.0.5, 3/4.5.1 l
Final Safety Analysis Report Sections 5.4.12, 6.3, 7.3, and 15 ASME Section XI 1980 Edition, Winter 1981 Addenda IWV-3500 Generic Letter 89-04 " Guidelines for Inservice Testing at Nuclear Power Plants" i
OPT-11 1.2.3 Check Valve Disassembly and Inspection OPT 20.12 E41-V159 Operability Test i
EER 94-0116 2-E41-V159 Inspection Evaluation and Permanent Repair j
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w.t og Rev. 3 Page 78 of 88 j
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J EER 92-0483 Rev 1 Page 7 of 12 ATTACHMENT 2 Page 57 of 69 10 CFR50.59 Program Manual Rev. 3 ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page 3 of 1 PART 11: ITEM CLASSIFICATION DOCUMENT NO.
EER 97-0483 REV.NO. 1 Yes No 1.
Does this item represent:
a.
A change to the f acility as described in the SAFETY ANALYSIS II IX)
REPORT 7 b.
A change to the procedures as described in the SAFETY ANALYSIS I1
[XI REPORT?
c.
A test or experiment not described i1 the SAFETY ANALYSIS II IX)
REPORT?
l 2.
Does this item involve a change to the individual plant Operating License or to II IX) l its Technical Specifications?
3.
Does this item require a revision to the FSAR?
I1 IX) 4.
Does this item involve a change to the Off Site Dose Calculation Manual?
!1 IXl 5.
Does this itern constitute a change to the Process Control Program?
I1 IXl 6.
Does this item involve a major change to a Radwaste Treatment System?
I1 IX) 7.
Does this item involve a change to the Technical Specification Equipment List
[]
IX)
(BSEP and SHNPP only)?
8.
Does this item impact the NPDES Permit (all 3 sites) or constitute an il IX)
- unreviewed environmental question * (SHNPP Environmental Plan, Section 3.11 or a "significant environmental impact * (BSEPl?
9.
Does this item involve a change to a previously accepted:
a.
Quality Assurance Program
[]
IX) j b.
Security Plan (including Trairnng, Qualification, and Contingency iI IX)
Plans)?
c.
Il IXl d.
Independent Spent Fuel Storage Installation license? (If *yes," refer to II IX]
Section 8.4.2. *Ouestion 9.* for special considerations. Complete Part VIin accordance with Section 8.4.6)
SEE SECTION 8.4.2 FOR INSTRUCTIONS FOR EACH *YES* ANSWER.
REFERENCES. List FSAR and Techrocal *Micaten references used to answer questions 19 above.
kkntity speofic refe<eace sectens veed toe any *Ves* answer.
See Pane 2
_ l fe w. 3 Page 79 of 88 om tog w
EER 92-0483 Rev 1 Page 8 of 12 ATTACHMENT 2 Page 58 of 69 10 CFR50.59 Program Manual Rev. 3 Page _4_., of J.
PART 111: UNREVIEWED SAFETY QUESTION DETERMINATION SCREEN DOCUMENT NO.
EER 92-0483 REV.NO. 1
.YJl!!i N2 1.
Is this change fully addressed by another completed
[]
[X]
UNREVIEWED SAFETY QUESTION determination? (See Sections 7.2.1, 7.2.2.5, and 7.9.1.1)
REFERENCE DOCUMENT: N/A REV.NO.
l Yes Ng 2.
For procedures, is the change a non-intent, change which gnA (check all that apply): (See Section 7.2.2.3)
[]
[X]
[]
Corrects typographical errors which do not alter the meaning or intent of the procedure; or,
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Adds or revises steps for clarification (provided they are consistent with the original purpose or applicability of the procedure); oi,
()
Changes the title of an organizational position; or,
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Changes names, addresses, or telephone numbers of persons; or,
)
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Changes the designation of an item of equipment where the l
equipment is the same as the original equipment or is an authorized replacement; or, i
()
Changes a specified tool or instrument to an equivalent substitute; or, Il Changes the format of a procedure without altering the meerwng, intent, or content; of (l
Deletes a part or aR of a procedure, the deleted portions of wher.h are whosy covered by approved plant procedures?
If the answee to setter Question 1 or Question 2 in PART lil is "Yes.* then PART IV need riot be cornpleted.
I O m oe
%. s ras. 80 of se I
l EER 92-0483 Rev 1 I
Page 9 of 12 i
i ATTACHMENT 2 Page 59 of 69 10 CFR50.69 Prograrn Manual Rev. 3 ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page 5 of 6 PART IV: UNREVIEWED SAFETY OUESTION DETERMINATION DOCUMENT NO.
92-0483 REV. NO. L Using the SAFETY ANALYSIS developed for the change, test or experiment, as well as other required references (LICENSING BASIS DOCUMENTATION, Design Drawings, Design Basis Documents, codes, etc.), the preparer of the Unreviewed Safety Question Determination must directly answer each of the following seven questions and make a determination of whether an UNREVIEWED SAFETY OUESTION exists.
A WRITTEN BASIS IS REQUIRED FOR EACH ANSWER Xf2 Nn 1.
May the proposed activity increase the probability of
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[X) occurrence of an accident evaluated previously in the SAFETY ANALYSIS REPORT?
T_he revised t?st will be oerformed with HPCI out of service and aoorooriete clearances in place. Normas acoroved methods will be used for workina on the valve. T h_g administrative controls are adeouate to ensure procer return to service. Therefore, this test will not increase the probability of a loss of coolant event or of a HPCI inadvertent iniection.
2.
May the proposed activit/ increase the consequences of an
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[X) accident evaluated previously in the SAFETY ANALYSIS REPORT 7 The revised test will not reduce the reliability of HPCI as oer 3 below, therefore HPCI will be available to oerform the assumed mitraation functions. The desian configuration of the plant es not affected at all.
3.
May the proposed activity increase the probability of
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[X) occurrence of a malfunction of equipment important to safety evaluated previously in the SAFETY ANALYSIS REPORT?
The revised test will oerform a better verification that the valve will function as reouired.
This will amorove the overall reliability of the component and the system.
4.
May the proposed activ;ty increase the consequence of a
[]
[X]
malfunction of equipment important to safety evaluated previously in the SAFETY ANALYSIS REPORT?
The revised testwill not reduce the reliability of HPCI as per 3 abos e. therefore HPCI will be available to perform the assumed mitioatien functions. The desian configuration of the olant is not affected.
5.
May the proposed actnnty create the possibility of an
[]
IX) accr$ent of a different type than any evaluated previously m ue SAFETY ANALYSIS REPORT?
The desion confiouraten of the otant ts riot effected.
OAs.tO9 Rev.3 Page 81 of 80 1
r-EER 92 0483 Rev 1 Page 10 of 12 I
ATTACHMENT 2 Page 60 of 69 10 CFR50,59 Program Manual Rev. 3 ATTACHMENT A CP&L SAFETY REVIEW PACKAGE Page 6 of_6 PART IV: Icontinued)
DOCUMENT NO.
EER 92-0483 REV.NO. 1 Y.n N2 6.
May the proposed activity create the possibility of a
[]
IX]
malfunction of equipment important to safety of a different type than any evaluated previously in the SAFETY ANALYSIS REPORT?
The deslan configuration of the olent is not affected by this test methodoloat 7.
Does the proposed activity reduce the margin of safety as
()
(X) defined in the basis of any Technical Specification?
The revised test w;11 continue to meet the surveillance requirement of 4.0.5. It will orovide better assurance thet the surveillance testina will detect the potential f ailures of the valve.
8.
Based on the answers to questions 1 - 7, does this item
()
(X) result in an UNREVIEWED SAFETY QUESTION? If the answer to any of the questions 1-7 is "Yes," then the item is considered to constitute an UNREVIEWED SAFETY QUESTION.
9.
Is PNSC review required for any of the following reasons?
[]
(X)
If, in answering question 1 or 3 *No," it was determined that the probability increase was small relative to the uncertainties; or, in anawaring question 2 or 4 *No," it was determined that the doses increased, but the dose was stillless than the NRC ACCEPTANCE LIMIT; or, in answering Question 7 *No,* a parameter would be closer to the NRC ACCEPTANCE LIMIT, but the end result was still within the NRC ACCEPTANCE LIMIT: then PNSC review is required.
REFERENCES:
~
See Pace 2 l
This Unreviewed Safety Question Determination is for the following DISCIPLINE (s): (Additional Part IV i
forms may be included as appropriate.)
l
[ ] Nuclear Plant Ope ations 11 Structural i I Nuclear Engineering
( l Metallurgy DG Mechanical i I Chemistry / Radiochemistry i I Electrical l I Health Physics l
( ) instrumentation & Control l I Administrative Controls OAs. tog Rev.L Page 82 of 88 l
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4 EER 92-0483 Rev 1 l
I Page 11 of 12 t
}
t EER No.
92 0483 Revision 1 p.g.11 Docunwnt Number Type (Technical Manuel, Specmcation, DBO, Group Respons% dor EDBS, Procedure, ete.1 Revision OPT-20.12 Revise Test Methodology BESS NSSS Mechsnical (Procedure OENP 16.7 Denote OPT-20.12 is en acceptable tort Action Request PAR) method in lieu of check velve disassembly outwnitted IAW CAP 403 per OPT.11,1.2.3
[(((f % Af,gjj, b
- Nb IST Vehre Tables trormerly CENP-17)
Denote IST program change I
i 1
o.E NP.12 40 Equivalent to Rev 34
__e
s, EER 92-0483 Rev 0 Page 12 of 12 ENGINEERING EVALUATION REPORT l
ENVIRONMENTAL QUALIFICATION IMPACT FORM (EER-EQlF)
Will the evaluation, on either a temporary or permanent basis:
[ ] Yes [XI No
I ] Yes [Xl No
- 3. Authorize the repair of EQ equipment / common components with other than avalified like-in-kind equipment / components parts?
{ j Yes [Xl No
- 4. Affect the existing ins Tilation or interface (of EQ equipment / common component applications, as may be designated in EDBS and/or in the quahfication data package (including changing the type of interface /
installation)?
[ ] Yes [X] No
- 5. Justify the (quality class) upgrade of equipment / common components gr component parts which could be utilized in EQ applications?
[ ] Yes [XI No
- 6. (Re) Define qualification parameters (e.g., normal or LOCA/HELB env:ron-mental conditions, postaccident operating time requirements, essential passive / active postaccident operating requirements, qualified life assumptions /results, etc.) for specific EQ equipment?
[ ] Yes [X] No
- 7. Provide an EQ-related justihcation for continued operation (as required per PLP-02, Section 4.4.3.3 gr 4.4.4)?
[ ] Yes [X] No
~
- 8. Provide the resolution of a qualification problem (as required per PLP-02, Section 4.4.4)?
I ) Yes [XI No Notes:
- 1. If all no, then no further EQ consideration is required. Mark the EE.R Traveler accordingly as required by ENP-12 and include this completed EER-EQlF within the EER package. An EQ Tochrdcol Review is p_qi required.
- 2. If an_y ves, an EQ impact assessment (per Section 5.3) must be performed during the evaluation process. Mark the EER Traveler accordingly and include this completed EER EQlF within the EER package. An EQ technical review is required.
7 Equivalent to BSEP/Vol. XXIENP-34.1 20 Rev. 4
ENCLOSURE 4 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 INSERVICE TESTING PROGRAM PLAN FOR THIRD TEN-YEAR INTERVAL (NRC TAC NOS. mal 115 AND MA1116)
COLD S11UTDOWN JUSTIFICATION CSJ-15 l
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Coldshutdow n. justification: CSJ-15 Page1
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SYSTEM:
i Nuclear Stearn Supply (D-02521, Sh.1 A&B, D-25021, Sh.1 A&B)
COMPONENTS:
1-B21-F022A thru l-B21-F022D and 1-B21-F028A thru l-B21-F028D 2-B21-F022A thru 2-B21-F022D and 2-B21-F028A thru 2-B21-F028D 1
CATEGORY:
)
A CIASS:
1
)
TEST REQUIREMENT:
Valves with fail-safe actuators shall be tested by observing the operation of the actuator upon loss of actuating power in accordance with the frequency of para. 4.2.1.1 (nominally every 3 months) (Part 10, Paragraph 4.2.1.6).
BASIS:
As described in Cold Shutdown Justification CSJ-01, these valves can only be exercised (full-stroke) during cold shutdown periods. During normal valve stroking, the fail-safe feature related to loss of electric power is verified; however, the fail-safe performance of the valves on loss of operating air pressure is not typically tested. To do so requires realignment of the main steam line isolation valve operating air supply system and,in the case of the B21-F022 valves, access to the drywell. The extent of the effort needed to
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perform this testing is beyond the normal scope of activities performed during short outages and would consume plant resources needed elsewhere for higher priority activities.
l AI. TERNATE TESTING:
During cold shutdown periods, the electrical fail-safe feature of these valves will be observed in conjunction with testing perfomied per CSJ-01, and at each refuel outage cach valve will be observed to operate properly in the fail-safe mode upon loss of the operating air supply pressure.
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..i ENCLOSURE 5 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 INSERVICE TESTING PROGRAM PLAN FOR THIRD TEN-YEAR INTERVAL (NRC TAC NOS. mal 115 AND mal 116)
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i VALVE RELIEF REOUEST VRR-12
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l Helief Request: VRR-12 Page1 SYSTEM:
Nuclear Steam Supply (D-02521 Sh. IC, D-25021Sh. IC)
COMPOSENTS:
1-1321-F010A and 1-821-F010B 2 B21-F010A and 2 821-F010B CATEQ!)RY:
AC CLAMi:
e d
LfHfJl)JJ1REMENT:
Check valves shall be exercised nominally every 3 months (Part 10, Para. 4.3.2.1).
A retest shewing acceptable performance shall be run following any required corrective action bef ore thu valve is returned to service (Part 10, Para. 4.3.2.6).
BAS!jiEOE RELIEF:
In accordance with 10CFR50.55a(f)(5)(iii), Carolina Power & Light (CP&L) is requesting approval to implement an impractical relief request, These valves open to provide flowpaths for HPCI and RCIC Cow into the reactor vessel as well as normal reactor feedwater makeup.
These are simple check valves, with no external means of exercising nor for determining disk position; thus, the only practical method of exercising these valves to their open position and confirming full open operation per the guidance of NRC Generic Letter 89-04 and NUREG-1482 is with flow from the reactor feedwater system, or from the HPCI or RCIC systems themselves. The HPCI accident flow requirement is 4250 gpm, and RCIC accident flow requirement is 400 gpm. Injecting water directly from either the HPCI or RCIC systems to the reactor is impractical during plant operation due to the possibility of creating an unacceptable reactor vessel water level transient, thermal shock to reactor vessel nozzles, a reactivity excursion, or upsetting reactor water chemistry. Under normal shutdown conditions, steam is unavailable to operate the HPCI and RCIC turbines and there r
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is a potential for over-pressurizing the reactor vessel. Thus, the only practical way of f
exercisir4 these valves is with reactor feedwater flow during power (steaming) operation.
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Itcliefitequest Vitit-12 Page 2 During normal plant operation, the feedwater flow is approximately 12.500 gpm per loop.
Normal plant operation exceeds 12,500 ppm, which is greater than the maximum accident l
flow of either llPCI or itCIC through these check valves. The reactor feedwater system arrangement is such that flow indication can be obtained for each of the individual feedwater loops. Thus, flow measurement through each check valve can be made to verify proper opening of the subject check valve.
i AI.TliitNATE TliSTING:
Exercising of these valves open will only be performed to the extent that adequate reactor
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feedwater flow is available. Full accident flow through each feedwater injection leg will be confirmed by monitoring A-loop and B-loop flow through feedwater flow venturis 1/2-C32-FE-N001 A/B during power operation. Where maintenance or corrective action has been performed on a valve during a shutdown period, the subject valve will not be flow tested (i.e.. opened) prior to being placed in service.
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ENCLOSURE 6 i
i BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62
' INSERVICE TESTING PROGRAM PLAN FOR THIRD TEN-YEAR INTERVAL (NRC TAC NOS. MA1115 AND mal 116) i l
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VALVE RELIEF REOUEST VRR-13 l
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Nuclear Steam Supply (D-02521, Sh. l A&B, D-25021, Sh. l A&B)
COMPONENTS:
1-821-F022A thru l-B21-F022D and 1-B21-F028A thru 1-B21-F028D 2-B21-F022A thru 2-B21-F022D and 2-B21-F028A thru 2-B21-F028D CATEGORY:
A CLASS:
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l TEST REQUIREMENT:
Other power-operated valves with reference stroke times less than or equal to 10 seconds hall exhibit no more than 150 percent change in stroke time when compared to the reference value (Part 10, Para. 4.2.1.8(d)).
BASIS FOR RELIEF:
In accordance with 10CFR 50.55a(a)(3)(i), Carolina Power & Light (CP&L) is requesting approval to implement an alternative test. The stroke times of these valves are adjusted within an acceptable band of 3-5 seconds by adjusting orifice::
associated with hydraulic dashpots attached to each operator. Thus, the stroke time performance of each valve operator is more a function of the dashpot setting than the material condition of the valve.
The acceptable band of I second is restrictive enough to ensure that each of the valves remains operable within the established limits of the plant safety analyses.
Elimination of the 50 percent limit on deviation will have no significant impact on the reliability of these valves nor on the health and safety of the public.
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IteliefItcuucs" VitR-13 Pace 2-t j
ALTERNATE TESTING:
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The acceptance criteria for closure stroke time for these valves will be 3-5 l
seconds, as established by the Brunswick Steam Electric Plant Technical Speci6 cations. An arbitrary reference value will be established at four seconds, and the acceptance values will be set at three and five seconds. These values are more conservative than the values established per the acceptance criteria of Part 10, Paragraph 4.2.1.8(c).
ENCLOSURE 7 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 INSERVICE TESTING PROGR AM PLAN FOR THIRD TEN-Y EAR INTERVAL (NRC TAC NOS. mal 115 AND mal 116)
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l EXCERPT FROM NRC LETTER DATED JANUARY 4,1990, "SECOND TEN-YEAR INTERVAL INSERVICE TESTING PROGRAM -
BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 (TAC NOS. 63.93 AND 63524)"
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UNITED STATES V d7
[ f.,fh NUCLEAR REGULATORY COMMISSION yO W/
WASHINGTON, D, C. 20555 s
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Ef January 4, 1990 g
. Docket Nos.
50-325 fjgC-9 0 - CM and 50-324-Mr. Lynn W. Eury Executive Vice President Power Supply Carolina Power & Light Company Pcst Of fice Box 1551 Raleigh, North Ca ro lina 27602
Dear Mr. Eury:
SUBJECT:
SECOND TEN-YEAR INTERVAL INSERVICE TESTING PROGRAM - BRUNSWICK STEAM ELECTRIC PLANT, UtlITS 1 AND 2 (TAC NOS. 63523 AND 63524)
By letter dated 6ctober 23, 1986, Carolina Power & Light Company (CP&L) submitted the proposed Inservice Testing (IST) Program for the second ten-year cycle of operation for the Brunswick Steam Electric Plant (BSEP), Units 1 and 2.
The second ten-year interval began on July 10, 1986 and ends July 9,1996.
The staff, with assistance from its contractor, Idaho National Engineering Laboratory (EG&G), reviewed the program.
During _ the course of the review, EG&G developed a list of coments and questions which were transmitted to CP&L as an attachment to the meeting notice dated July-15, 1987.
CP&L provided a written response to the EG&G comments and questions, which was further. discussed during a working meeting with your staff, and the hRC staff and NRC contractors.
This meeting was held on July 21-22, 1987 at the BSEP site.
Additional staff concerns and open items en the IST program were identified at the meeting.
To address staff concerns CP&L provided a revised program by letter dated Novenber 24, 1987, and provided additional changes by letters dated l
March 28,1988, November 2,1988, and Septenber 8,1989.
The BSEP, Units I and 2, second ten year IST program, as submitted and amended, was evaluated for complience with the Code of Federal Regulations, I
10 CFR 50.55a(g), which requires certain Class 1, 2 and 3 pumps and valves in water-cooled nuclear reactor facilities to meet the inservice testing requirements stated in the LSME Boiler and Pressure Vessel Code,Section XI; specifically, subsection IWP, Inservice Testing of Pumps in Nuclear Power Plants, and Subsection IWV, Inservice Testing of Valves in Nuclear Power Plants.
Each facility is recpired to establish a program for the inservice-testing of pumps and valves which is updated every ten years to meet the requirements in the latest approved edition and addenda to Section XI of the ASME Code.
The staff, with technical assistance from EG&G, has reviewed and evaluated the program originally submitted and the revised program and additional changes provided by CP&L letters, including the requests for relief from those require-ments that CP&L determined to be impractical to perform at BSEP, Units I and 2.
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The alternate exercising frequency required by Technical Specifications has been previously approved by the NRC staff to reduce wear on the control rod drive mechanisms and to reduce the number of rapid reactivity transients to which the reactor core is exposed.
Based on the determination that the Code requirements are impractical and that the licensee's proposal provides a reasonable alternative to the Code requirements, relief should be pe nted j
as requested.
I 4.3 Nuclear Steam Sucoly System 4.3.1 Cateaory B/C Valves 4.3.1.1 Relief Reauest.
The licensee has requested relief from the exert:ising requirements of Section XI, Paragraph IWV-3412, for the automatic depressurization system (ADS) valves, B21-F013A thru H, J, K, and L, and proposed to exercise these valves once each 18 months with reactor supplied steam.
4.3.1.1.1 Licensee's Basis for Reauestina Relief--Opening an ADS valve during normal operation would place the plant in a " mini-LOCA" condition if the ADS valves were to fail in the open position.
The amount of steam injected into the suppression pool could cause a rise in suppression pool temperature beyond the Technical Specification operating limits.
It is impractical to measure stroke times for the ADS valves, since the stroke times are on the order of 100mS.
Steam flow measurements and/or turbine bypass valves position will verify that the ADS valves have performed their function in less than five seconds. Time "zero" for this stroke time measurement corresponds to the instant the ADS hand switch is aligned in the "open" position.
I Alternate Testina:
Each valve will be exercised at least once per 18 months when the reactor is operating at. sufficient power to bypass a quantity of steam through the turbine, bypass valve (s) equal to or greater I
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than the capacity of an ADS valve.
Since the turbine bypass valves respond automatically to RPV dome pressure, the actuation of an ADS valve wiil result in rapid closure of the turbine bypass valves.
Conversely, closing the ADS valve will be accommodated by rapid opening of the turbine bypass valves.
A change in turbine bypass valve position can be directly associated with a certain steam flow rate. This flow rate would be equal to the quantity of steam discharged by the ADS.
No stroke time measurements will be performed.
An abrupt change in turbine bypass valve position or steam line flow (per Technical Specification 4.5.2.b) within five seconds will be adequate to demonstrate valve operability.
4.3.1.1.2 Evaluation--The AOS valves act both as power operated valves in response to a manual or automatic control signal and as safety relief valves.
As a result, these valves should be tested to both the Category 8 and C requirements.
Full-stroke exercising these valves quarterly during power operations is ircpractical as this greatly increases the risk of creating a mini loss-of-coolant accident.
NUREG-0626 " Generic Evaluation of Feedwater Transients and Small Break Loss of-Coolant Accidents 1
in GE-Designed Operating Plants and Near Term Operating License Applications" recommends reduction of challenges to relief valves to lessen the risk of Small Break LOCA (see also NUREG-0737,Section II.K.3.16).
To full-stroke exercise these valves requires reactor steam pressure and is not practical during cold shutdowns when the reactor pressure is low. Accurate stroke times for these valves cannot be obtained since their stroke times are on the order of 100 milliseconds and there is no direct position indication.
The licensee's proposal to verify these valves' operability once each 18 months with the reactor at power by passing reactor j
steam t'hrough the valves and to verify the valve strokes in less than five seconds utilizing turbine bypass valve position provides an acceptable level l
of quality and safety and a reasonable alternative to the Code requirements.
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4 Based on the determination that the Code requirements are impractical, that the licensee's proposal provides a reasonable alternative to the Code j
requirements, relief should be granted as requested.
4.4 Core Soray System 4.4.1 Cateoory A/C Valves
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4.4.1.1 Relief Recuest, The licensee has requested relief from the exercising (open) frequency requirements of Section XI, Paragraphs IWV-3412 and 3522, for the core spray injection isolation valves, E21-F006A and F0068, and proposed to exercise these valves during each refueling outage.
4.4.1.1.1 Licensee's Basis for Reauestino Relief--These valves
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open to allow core spray injection into the reactor vessel.
Testing of these valves requires initiating core spray and injection j
into the reactor vessel.
Core spray is a low pressure system, therefore, for system protection the inboard and outboard isolation valves (E21-F004A&B and E21-F005A&B) are interlocked to allow only one valve to open at a time when the reactor i
vessel's pressure is greater than 410 psi.
During reactor operation, the reactor vessel is at approximately 1000
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psi.
Introducing nonpreheated water into the reactor vessel requires the U
vessel shell temperature to be less than 200 F to limit possibility of thermal shock, which could cause reactor vessel nozzle cracking. During normal cold shutdown, the reactor vessel shell temperature does not get 0
below 200 F.
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ENCLOSURE 8 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 INSERVICE TESTING PROGRAM PLAN FOR THIRD TEN-YEAR INTERVAL (NRC TAC NOS. MAlll5 AND MAlll6) l VALVE RELIEF REOUEST VRR-11 l
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l Relief Reauest: VRR-11 Page 1 of 3 Rev.O Vacuum Breaker Test Frequency l
SYSTEM:
Containment Atmospheric Control (CAC) (D-25015, Sheet 1B and D-02515, Sheet 1B)
CLASS:
2 COMPONENTS:
1-CAC-X20A and 1-CAC-X20B 2-CAC X20A and 2-CAC-X20B CATEGORY:
A/C TEST REQUIREMENTS:
Subsection IWV of the American Society of Mechanical Engineers (ASME) Code,Section XI, 1
requires valve testing be performed in accordance with the requirements contained in the ASME Code for Operations and Maintenance of Nuclear Power Plants (OM Code), Part 10, " Inservice T esting of Valves in Light-Water Reactor Power Plants." Paragraph 4.3.1 of the OM Code, Part 10, states that safety and relief valves shall meet the inservice test requirements of the OM Code, Part 1,1987 Edition.
The OM Code, Part 1, Paragraph 1.3.4.3, " Primary Containment Vacuum Relief Valves,"
requires the following:
1.
Within every six month period, operability tests be performed unless historical data indicates a requirement for more frequent testing.
2.
Leak tests be performed every two years unless historical data indicates a requirement for more frequent testing.
The scope of the operability tests is further defined in OM Code, Part 1, Paragraph 3.3.2.3,
" Vacuum Relief Valves," which states:
1.
The valves be actuated to verify open and close capability, set pressure, and performance of any pressure and position sensing accessories.
2.
Determination of compliance with the Owner's seat tightness criteria.
Relief Reauest: VRR-11 Page 2 of 3 Rev 0 Vacuum Breaker Test Frequency BASIS FOR RELIEF:
Carolina Power & Light (CP&L) Company is requesting approval from the impractical requirements of the OM Code, Part 1, Paragraphs 1.3.4.3 and 3.3.2.3 with respect to the set pressu e testing of the primary containment vacuum relief valves every six months. This request
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for reliefis being made in accordance with 10 CFR 50.55a(f)(5)(iii).
- Maintenance on the subject valves would require a Local Leak Rate Test (LLRT), in accordance
- with 10 CFR Part 50, Appendix J, to verify satisfactory leak tightness subsequent to the maintenance activities. The adjacent inboard containment isolation valves (i.e., CAC-V16 and CAC-V17) do not perform a reliable seal to use as an LLRT boundary valve.
The CACN16 and CAC-V17 valves are Posi-Seal butterfly valves with offset discs and stems.
Although the valves are considered to be capable of by-directional sealing, maximum sealing is achieved when the valve disc seal ring is located on the lower pressure side of the valve stem. In 1
this " preferred" orientation, the seat ring retaining ring is also located on the outboard face of the valve and the valve packing is subjected to containment side pressure only. The valves were originally installed in this orientation. LLRTs were performed by pressurizing the volume L
between the inboard and outboard valve (i.e., Primary Containment Vacuum Breaker and CAC-V16/V17). This mode of testing excluded the inboard valve packing from the LLRT.
In 1985, in response to NRC Inspection Report 50-325,324/85-31, inboard valve (i.e.,
CAC-V16/V17) orientation was reversed in order to place the valve packing within the LLRT L
boundary. As a result, the inboard valves were installed with the retaining ring located on the
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containment side (inboard) valve face. In this orientation, the valves no longer offer maximum disc sealing in the loss-of-coolant accident (LOCA) direction, and difficulties were encountered l'
in achieving satisfactory ILRT results, in spite of satisfactory LLRT results. The valves were l0 subsequently reversed again, and are in their current orientation with the retaining ring located on the outboard face of the valve. When the valves were reversed, a flanged connection was added
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. to the torus to allow testing in the LOCA direction.
Post-maintenance testing associated with the Primary Containment Vacuum Breakers would p
require an LLRT. Performance of the LLRT would require shutdown the unit and de-inerting in order to enter the torus for test flange installation. To perform the LLRT between the Primary Containment Vacuum Breaker and the CAC-V16/V17 would require design and installation of a valve capable of sealing in both directions against the LLRT.
i In addition, in accordance with Technical Specification 3.6.1.5.3, a functional test is performed on these valves every 92 days and the full open setpoint test is performed every 24 months in accordance with Technical Specification 3.6.1.5.4. The Technical Specification Bases for the 24 month setpoint test frequency states "The 24 month frequency has been demonstrated i
neceptable, based upon operating experience, and is further justified because of other Surveillance performed more frequently that convey proper functioning status of each vacuum breaker."
i Relief Reauest: VRR-11 Page 3 of 3 Rev.O Vacuum Breaker Test Frequency ALTERNATE TESTING:
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- Each of these primary containmem vacuum relief valves will be set pressure tested each refueling in accordance with Technical Specification 3.6.1.5.4 and functionally tested quarterly in accordance with Technical Specification 3.6.1.5.3.
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ENCLOSURE 9 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324/ LICENSE NOS. DPR-71 AND DPR-62 INSERVICE TESTING PROGRAM FOR THIRD TEN-YEAR INTERVAL (NRC TAC NOS. MAlll5 AND MAlll6)
SUMMARY
OF RELIEF REOUESTS FOR THIRD INTERVAL INSERVICE TESTING PROGRAM.
BRUNSWICK STEAM ELECTRIC PLANT. UNIT NOS.1 AND 2 l
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