ML17272A202
| ML17272A202 | |
| Person / Time | |
|---|---|
| Site: | Columbia, Washington Public Power Supply System  |
| Issue date: | 11/30/1978 |
| From: | Renberger D WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | Volgenau E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| References | |
| QA-78-001, QA-78-1, NUDOCS 7812070208 | |
| Download: ML17272A202 (38) | |
Text
6'5lg DOCKET NBR:
REGULAT INFORMATION DISTRIBUTIO YSTEM DOC DATE:
781130 RECIPIENT:
ORIGINATOR.
RENBERGER, D.L.
ACCESSION 'NBR: 7812070208 COPIES RECEIVED:
COMPANY:
SUBJECT:
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cnNSTRUCT ION OFFICIENCY REPORT
( 1 OCFR50
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REGULATOR DOCKET F(LE COPY Washington Public Power Supply System A JOINT OPERATING AGENCY P. O. SOX 9dd 3000 OCO. WAOHIN4TON WAY ITICHI ANO, WAXHINOTOH 99353 PHONC (509) 375 5000 November 30, 1978 QA-78-001 Nuclear Regulatory Comission Office of Inspection 8 Enforcement Washington, D.
C.
20555 Attention:
Mr. Ernest Volgenau, Director
Subject:
REPORTABLE 10CFR50.55 e
REPORTS
Dear Mr. Volgenau:
In accordance with the provisions of 10CFR50.55(e), written reports on reportable deficiencies must be made to the appropriate NRC Regional Office within 30 days after verbal notification.
Copies of reports shall be sent to the Director of Inspection and Enforcement,
- NRC, Washington, D.
C.
The attached reports are reportable deficiencies under the provisions of 10CFR50.55(e).
They have been reported to the Nuclear Regulatory Commission, Region V within the required 30 days.
Copies of these reports were not sent to the Director of Inspection and Enforcement, NRC, Washington, D.C.
The attached reports correct this situation and, in the future, your office will receive copies concurrently with the Region V office.
List of attachments:
Letter No.:
Date:
==
Description:==
G02-77-303 - Docket Number 50-397, CPPR-93 August 18, 1977 WNP-2 Overload of Standby Gas Treatment System Due to Fuel Pool Boil-Off at 212 F
Letter No.:
Date:
==
Description:==
Letter No.:
Date:
==
Description:==
Letter No.:
Date:
==
Description:==
812070 2c)3 G01-77-362, G01-77-480, G01-77-558, G01-77-722, GOl-78-62. - Docket Number 50-460, CPPR-134 July 5, 1977, August 18, 1977, September 15, 1977 November 14, 1977 and January 23, 1978 Voids in Concrete Placement No.
28-GSB WNP-1 G01-76-694, G01-76-767, G01-77-133 - Docket Numbers 50-460 and 50-513 November 5, 1976, December 17, 1976 and February 24, 1977 Seismic Analysis of General Services Buildings of WPPSS Projects Numbers 1
8 4
G02-78-151 - Docket Number 50-397, CPPR-93 May 30, 1978 Opening in Tornado Missile Barrier in Diesel Generator Building - WNP-2
N0 Mr. Ernest. Volgenau Page 2
Reportable 10CFR50.55(e)
Reports Letter No.:
G02-78-150 - Docket Number 50-397, CPPR-93 Date:
May 26, 1978
==
Description:==
Possible Overexposure of Control Room Operators to Radiation Following a LOCA - WNP-2 Letter No.:
G02-77-504, G02-78-121 - Docket Number 50-397, CPPR-93 Date:
December 27, 1977 and April 10, 1978
==
Description:==
Cracks in Structural Steel Platform at 541'levation Inside Containment - WNP-2 Letter No.:
G02-78-245 - Docket Number 50-397, CPPR-93 Date:
November 6, 1978
==
Description:==
Tilting Disc Check Valves Failure to Close with Gravity in Vertical Position - WNP-2 If you require additional information, please contact us.
Very truly yours, DLR:000:seb Attachments D. L.
RENBERGER Assistant Director, Technology
'j'~hingtQn Pubtic PQ~ver Supclv Syst m
A JOINT QP RATING AGENCY
- p. 0. 8ox 988 3000 ceo. Wisr <r oro~ whv RroM.ioo, wASRI ~ iororc 99352
&ironic (509) 375 5000 August 10, '1977 GO-77-303 i'Iuclear Regulatory ComiIission Region V
Suite 202,. Walnut Creek Plaza 1990 N. California Souleval d Walnut Creek, California 94596 Attention:
Subject:
Nr.
R.
H. Enigelken, Director WPPSS i'IUCLEAR PROJECl'0..2 DOCKET IirUIiBER 50-397, CPPR-'93
. REPORTABLE EVEI'IT - 10CFR5tl.55 e
Dear Hr. Engelken:
Your staff v)as infer'med'f a reportable deficiency under the provisions of 10CFR50.55(e) by, tg3econ on July 29, 1977.
Pttqched is our report on this problem.
If you require additional information, please feel free to contact us.
OLR:OCT:dag Attachment cc:
D. Ro, BPA (1)
Very truly yours,
~Qo~p
.D. L. REi'ISERGER Assistant Director Generation 8 Technology
ATTACHMENT 81.
REPORTABLE DEFICIENCY AND CORRECTIVE ACTION WNP-2 OVERLOAD OF STANDBY GAS TREATi4IENT SYSTEM DUE TO FUEL POOL BOIL-OFF AT 212 F
Washington Public Power Supply System Docket No. 50-397 License No. CPPR-93
'escri tion of the Deficienc The Standby Gas Treatment System (SGTS) is an Engineered Safety Feature (ESF) filter system requjred to perform safety-related functions following a design basis accident..
The SGTS, post-LOCA, is required to maintain a
negative 0.25 in. w.g. pressure in the reactor building to prevent direct
'utleakage of radioactive fission products to the environment.
This design criteria would be violated during a
LOCA subsequent to a Safe Shutdown Earthquake (SSE).
The SGTS capacity would be exceeded due to moisture overload resulting.in. a reduction in the Reactor Build.:ng negative pressure.
Cause and Anal sis The fuel pool cooling system is Seismic Categorv II and is assumed to fail during an SSE.'he fuel poo'i water temperature will rise to 212 F, therebv gradually increasing the rate of evaporation causing the SGTS capacity to be exceeded.
The extent of overload is dependent upon the amount of spent fuel stored in the fuel pool, e.g.,
two 30Ã cores will exceed the capacity by about 32/.
The effect of this overload would be to reduce the negative pressure in the reactor building from 0.25 in. w.g.
to a negative pressure of 0.03 in. w.g.
This condition does not conform to safety analysis report design criteria and is a significant deficiency in the final design as approved and released for construction.
Safet Im lications For the reactor building to qualify as a secondary containment for the purpose of fission product control, the volume should be held at a minimum negative pressure of 0.25 in.w.g. when compared with adjacent regions.
This 'criteria is valid up to wind speeds which cause diffusion
, adequate to compensate for the increased exfiltration with respect to site boundary exposures.
Considering the 0.03 in. w.g. negative pressure, the wind speed trade-off is not justified.
Since an adequate negative pressure is not maintained, a positive pressure time period must be assumed for the reactor building which causes direct outleakage, thus allowing no credit or a significantly reduced credit for fission product control by the SGTS.
This increases the calculated site boundary and low population zone post-LOCA doses presented in the safety analysis
- report, Chapter 15.
lliNP-2 Overload of Standby Gas Treatment Page two Corrective A'ction Taken and Planned The SGTS is being modified to include cooling c'oils in the inlet to each unit to.condense the water vapor and reestablish the design conditions for the system.
In this manner, the capacity for the SGTS is not ex-ceeded and a negative 0.25 in. w.g. pressure is maintained in the reac-tor building.
The spent fuel storage rack design is currently being modified to
'accommodate high density storage.
The design change impacts the water evaporation rate from the spent fuel pool after an assumed
- SSE, and,
- thereby, influences the design criteria for. the cooling coils.
The cooling coil design will be completed in conjunction with the high density fuel storage modification.
The incorooration of the cooling coils into the SGTS will be noted in the safety analysis report as verification that the deficiency has been corrected.
This deficiency is considered to be an isolated case.
The deficiency was identi ied during a routine audit, indicating the system of checks and reviews is performing its function.
Oue to the extensive reviews in the past by our own personnel and the Architect-Engineer of design criteria, the system designs and design interfaces, the PSAR and
- FSAR,
~ no specific additional. action is required related to this deficiency.
~
i cg
-4 Vlashington Public Pa~ver Supply System A JOiNT OPERAT(NG AGENCY
/K>5 P. 0. BOX 966 3000 0XO W+SHviOVoN WAY BtatuoO. WA HWCTON 993S2 CONC(509) 946.1611 July 5, 1977 GOI-77-362 Nuclear Regulatory Commission Region V
Suite, 202, 'lalnut Creek Plaza 1990 N. California Boulevard
!!alnut Creek, California 94596 Attention:
t1r.
R.
H. Engelken, Director
Subject:
l!PPSS NUCLEAR PROJECT NO. I DOCKET NU!iBER 50-460, CPPR-134 REPORTABLE EVENT - IOCFR50.55 e
Dear f!r. Engelken:
Your staff was informed of a reportable deficiency under the provisions of IOCFR50.55(e) by telecon on June 6, 1977.
The interim r port attached provides our initial report on the prob'Iem.
A final report will be provided by August 19, 1977.
If you require additional information, please feeI free to contact us.
Very truly yours, DLR:DHH:vh Attachment D. L.
RENBERGER Assistant Director Generation 5 Technology cc:
CR Bryant, BPA
I
REPORTED DEFTCIEHCY AND CORRECTIVE ACTION FOR VOIDS IN CONCRETE PLACEl1ENT NO. 28-GSB-I INTERIM REPORT Des criotion of the Defici enc On Nay 23, 1977 voids Here discovered on the north face of Placement No. 28 when the forms were removed.
Exploratory chipping of concrete has revealed that the voids are located around she third and fourth layers of top mat reinforcing steel.
They appear to extend from 5 to 15 feet into Block No. 28 in a north-south direction and are located I to 8 feet from the west edge of the foundation mat.
Vertical height of the voids was from 8 inches to less than I inch.
All exploratory work to-date has been performed from the under-side of the top mat reinforcing steel.
Cause and Anal sis The voids in Placement No. 28-GSB-I have been attributed to improper consoli-dation of concrete in an area of extensive rebar congestion.
Required spacing of reinforcing, by design, is one No.
11 reinforcing bar spaced every 7 inches.
North-south reinforcing bars in the third and fourth layers where Iap splices are not staggered were to lap a minimum of 10 feet; however, the bars were detailed and fabricated sUch that only 8 feet 6 inches of lap existed.
An additional 25 feet 6 inch bar was added to provide required lap.
This increased the density of reinforcing steel in this area above that provided by the original design.
In spite of the increased stee'I
- density, the voids would not have occurred if the concrete had been properly consolidated.
Safet lm Iications Incomplete concrete bonding to reinforcing steel in the void areas could have affected the ability of the structure to resist design forces.
Corrective Action Taken and Planned The following actions have been taken or are planned to correct the deficiency:
Placement No. 28-CSS-1 Voids Interim Report Page 2
Exploratory chipping has been conducted, in accordance with approved procedures, to determine the extent of the voids.
A four (4) inch core boring was performed to determine if any other voids may be present in another area of high rebar congestion in Placement No. 28.
Upon inspection of the core, it has been determined that no other voids exist in areas tested.
Concrete will be removed from the top of the foundation mat to expose all voids in the top mat in the affected area.
Concrete above the void areas will also be removed.
Reinforcing steel
~ which must be cut to facilitate concrete removal will be cadweld spliced to bars in the adjacent segment.
Supplemental lap bars in the highly congested area will be eliminated where possible and main reinforcing bars will be cadweld spliced to reduce the density of reinforcing and permit space for proper placement and consolidation.
The surface of the concrete wi 11 be cleaned of loose material, coated with an approved bonding agent and concrete placed in the affected area in accordance with approved concrete repair proce-dures.
The repair will ensure that there will be no impairment to the integrity of the completed structure.
All work associated with the repair is expected to be completed approximately August I, 1977.
A complete report will be provided fifteen (15) calendar days following completion of repairs.
The following actions have been taken or are planned to reduce the potential for similar occurrences:
~
Lap splices in reinforcing steel in the corresponding area of the General Services Suilding of Hi')P-4 Huclear Project have been staggered to provide increased space or concrete placement.
e Other areas of dense reinforcing will be reviewed by the AE, and any appropriate adjustments made to provide additional space for vibrator operation to assure proper consolidation of concrete.
V/Qshington Public Powe'r Supply System A JOINT OPERATING AGENCY P. 0 BOX 966 3000 CCO, WaSRi~ern~ Wix Rir>l.*~O. WioooiOZON 9935R P~OVC(509) 946-1611 August 18, 1977 G01-77-480 Nuclear Regulatory Commission Region V
Suite 202, Walnut Creek Plaza 1990 N. California Boulevard
!talnut Creek, California 94596 Attention:
ter.
R.
H. Engelken
Subject:
WPPSS 1IUCLEAR PROJECT NO.
1 DOCKET NUttBER 50-460, CPPR-134 REPORTABLE 10CFR50.55 e
CONCRETE VOIDS
Dear ttr. Engelken:
Your staff was informed of a reportable deficiency under the provisions of 10CFR50.55(e) by telecon on June 6, 1977.
An initial report was transmitted July 5, 19?7.
A final report was to.have been provided by August 19, 1977.
- However, due primarily to contractural problems with the contractor responsible for the work (Hoffman Construction Company) the repair work associated with the deficiency has not been completed.
Work is continuing to remove defective concrete and assure that no additional voids exist.
In addition, rock pockets have been identified on'the south face of GSB No.
1 placement 8U.
The presence of these rock pockets indicates the potential for voids in this area also.
Your office will be kept informed of MPPSS progress in evaluation of the extent of the voids and of the progress towards final repair.
It is anticipated at this time, that a final report on the incident will be provided by September 16, 1977.
Very truly yours, D. L.
RENBERGER Assistant Director, Projects DLR:DHM:djs cc:
SB Barnes - UEIlC, Field EC Haren - UE8C, Field CR Bryant - BPA
Washington Public Power Supply System A JOINT OPERATING AGENCY P
O.. Box 998 3Doo Qco w*siiiiioroNwhv Rieiii.*iio,wisiiiiioro~ 99353 piioiict309> 94e.loll September 15, 1977 G01-77-558 Nuclear Regual tory Commission Region Y
Suite 202, Walnut Creek Plaza 1990 t). California Boulevard Walnut Creek, California 94596 Attention:
Hr. R. H. Engelken
Subject:
HPPSS NUCLEAR PROJECT NO.
1 DOCYET HUl1BER 50-460 - CPPR-134 REPORTABLE 10CFR50.55 e
- CONCRETE YOIDS
Dear Hr. Engelken:
Your staff was informed of a reportable deficiency under the provisions of 10CFR50.55(e) by telecon on June 6, 1977.
An initial report was transmitted July 5, 1977 and an interim status report sent on August 18, 1977.
The August 18, 1977 letter stated a final report would be issued September 16, 1977.
Since the contracto~
has not yet completed the repair, a final report cannot be issued at this time.
Work is continuing to remove defective concrete and to prepare the area for replacement of reinforcing steel which had to be removed to all.ow access to the void.
The reinforcing steel in the void area is being redesigned to relieve the congestion in the area to facilitate a successful repair.
Your office will be kept informed of 1lPPSS progress in evaluation of the extent of the voids and the progress toward final repair."
Reevaluation of the time necessary to effect the repair indicates a completion of repair and issuance of a final report by November 15, 1977.
DLR:lNH:LEN'vh
'Itery truly yours, n, L.
RENBERGER Assistant Director Generation 8 Technology
Pk.
I J
'~
Jg washington Public Power supply system A JOINT OPERATING AGENCY P 0 BOX 968 3000 QCO. WASei~CVON W*V RiCR1.irCO. V(*SHINOVON 99352 PHOHC1509) 946 1611 tlovembe~r4, 1977 G01 722 Hucl ear Regulatory Commi ssion Region V
Suite 202, Malnut Creek Plaza 1990 H. California Boulevard
'walnut Creek, California 94596 Attention:
Mr.
R.
H. Engelken
Subject:
MPPSS HUCLEAR PROJECT HO.
1 DOCKET NUMBER 50-460 - CPPR-134 REPORTABLE 10CFR50.55 e
>'COHCRETE 'lOIDS
Dear Mr. Engelken:
S'our staff was informed of a reportable defi iI n under the provisions of 10CfR50.55(e) by telecon on une 6, 1977.
An initial report was transmitted July 5, 1977 and interim status reports sent on August 18, 1977 and September 15, 1977.
The September 15, 1977 letter stated a final report would be issued Hovember 15, 1977.
Since the repair has not been completed, a final report cannot be issued. at this time.
An attempt has been made to determine the acceptability of concrete outside of the known voids using ultrasonics and core boring.
During the core boring operation another area of unsound concrete was discovered.
This new area is now under investigation and further tests are being conducted to identify any other problem areas.
Additional core borings are being taken to assure con-crete integrity along the entire west wall of GSB Ho.
1 placement 28.
Your office wi'tl be kept informed of MPPSS progress in evaluation of the extent of the voids and the progress toward final repair.
- P l1r.
R.
H. Engelken Page 2
I llovember 14, 1977 G01-77-722 The nature of the exploratory
<vork in this placement makes an accurate prediction of the completion date very difficult.
A status report Mill be issued by january 31, 1977.
Very truly yours,
~~4 l~gL~
D. L.
RENBERGER Assistant Director Generation I5 Technology DLR:DHW:LEt<:djs cc:
CR Bryant -
BPA SB Barnes - UE8C, Field EC Haren -
VEDIC, Field
UV IQ Vjashinigton Public Power Supply System A JOINT OPERATING AGENCY P 0, Gox 9dS 3000 Gco. Wisvt~ayQ~ 'LNgy RlcH+a'JO, V/ASHltlGI'OH 9935K 1'H0>c15091 945 1511
~3 G01-78-62 I'luclear Regulatory Commission Region V
Suite 202, Walnut Creek Plaza 1990 N. California Boulevard Walnut Creek, Cali fornia 94596 Attention:
1~tr.
R.
H. Engelken
~
Subject:
I!PPSS NUCLEAR PROJECTS 1lOS.
1/4 DOCKET NUhBER 50-460 - CPP+-134 REPORTABLE 10CFR50. 55 e
> COIlCRETE VOIDS
Dear ttr. Engelken:
S Your staff was informed of a re ortable deficienc under the provisions of 10CFR50.55(e) by telecon on June 6, 1977.
An initial report was trans-mitted July 5, 1977 and interim status reports sent on August 18,
- 1977, September 15, 1977 and November 14, 1977.
S of the problem, the actions taken to correct the discrepancy and the actions taken to prevent recurrence of this type of incident.
If you require additional information, please feel free to contact us.
DLR:DHW:LEN:djs Attachment 7
Very truly'yours,
/; S A~~~~~
D. L.
RENBERGER Assistant Director
.Generation 8 Technology cc:
CR Bryant - BPA SB Barnes - UE8C, Field EC Haren - UE8C, Field
+0! ~6+ ~
Attachment 1
Page 2
o The excavated area extended approximately 4 feet in depth, along the entire, vest face of Placemeht 28 and 8 feet from tne west face.
e After all known defective concrete had been removed, additional core borings were taken from the western edge of the excavated area easterly into segment 28 to verify the integrity of the remainder of the segment.
These additional core borings were examined and indicated no additional problem or suspect areas.
a Reinforcing steel removed to facilitate concrete removal was re-placed by cadweld splicing to bars in the adjacent segment or to bars within segment 28 as appropriate.
Supplemental lap bars in the highly congested area were eliminated where possible and main reinforcing bars were cadweld spliced to reduce the density of reinforcing and increase space available for placement and consolidation.
~ ~
~
o The surface of the concrete was cleaned of loose material, coated with an approved bonding agent and concrete placed in the affected area in accordance with approved concrete repair procedures.
The repair ensures that there.vill be no impairment to the integrity of the completed structures.
~
The concrete pour-back took place on January 9,
1978 in accordance with the disposition to Nonconformance Report 1-NCR-205-183.
The following actions have been taken or are planned to reduce the potential for similar occurrences:
Lap splices in reinforcing steel in the corresponding area of the General Services 8ui lding of !ANP-4 have been staggered to provide increased space for concrete placement and consolidation:
Other areas of dense reinforcing will be reviewed by the AE, and appropriate adjustments made to provide additional space for vibrator operation to assure proper consolidation of concrete.
o The Contractor involved has conducted training sessions to demon-strate proper consolidation procedures to supervision and workmen involved in concrete pTacement.
o Inspection activities by the Contractor was increased and additional surveillance effort by the Architect Engineer and the Owner was conducted.
The results from'this effort indicated no similar problems exist in other placements.
REPORTED DEFICIENCY AND CORRECTIVE ACTION FOR VOIDS IN CONCRETE PLACEMENT NO.
28-GSB-1 Descri tion of the Deficienc On Hay 23, 1977 voids were discovered on the north face of Placement No.
28 when the forms were removed.
Exploratory chipping and core boring of concrete has revealed that.the voids are located around the third and fourth layers of top mat reinforcing steel.
They extend the full width of Block Wo.
28 in a north-south direction and are located I to 8 feet from the west edge of the foundation mat.
Vertical height of the voids was from 8 inches to less than 1 inch.
Cause and Anal sis The voids in Placement iVo. GSB-1-28 have been attributed to improper con-solidation of concrete in an area of extensive rebar congestion.
Required spacing of reinforcing, by design, is one Ho.
11 reinforcing bar placed every 7 inches.
North-south reinforcing bars in the third and fourth layers where lap splices are not staggered were to lap a minimum of 10 feet;
- however, the bars were detailed and fabricated such that only 8 feet 6 inches of lap existed.
An additional 25 feet 6 inch bar was added to provide re-quired lap.
This increased the density of reinforcing steel in this area above that provided by the original design.
In spite of the increased steel
- density, the voids would not have occurred if the concrete had been properly consolidated.
Additional voids were discovered south of the areas of extensive rebar con-gestion.
These additional voids were in an area of relatively low rebar density and can only be attributed to poor consolidation practices.
Safet Im lications Incomplete concrete bonding to reinforcing steel in the void areas could have affected the ability of the structure to resist design forces.
Corrective Action Taken and Planned The following actions have been taken to correct the deficiency:
~
Exp'loratory chipping of concrete from the top west face of the foundation was performed to determine the extent of the voids and to impl'ement the repair.
Core borings were taken 'from the west face of the placement horizontally to a depth of approxi-mately 8 feet to further define the extent of the defect.
The core borings indicated the presence of defective concrete along the entire west edge of Block No. 28.
Existing reinforcing steel in all four layers of the top mat and at the west side of the placement was removed as required to provide access for chipping equipment.
All concrete above and including the defective area has been removed..
Washington Public Power Supply System A JOINT OPERATING AGENCY P. 0. 8OX 988 3000 CCO. WASW1tCOTON W*V R<aCL.A~O. W*CV>VOrON 99352 PVONC<509) 94d-ldll November 5, 1976 G01-76-694 Nuclear Regulatory Commission Region Y
Suite 202, Walnut Creek Plaza 1990 H. California Boulevard Walnut Creek, California 94596 Attention:
Nr.
R.
H. Engelken, Director
Subject:
WPPSS NUCLEAR PROJECTS NOS.
1 AHO. 4.
POTENTIALLY REPORTABLE EVENT DOCKET HUHBERS 50-460 AHO 50-513
Dear ttr. Engelken:
This is to inform you of an event which is potentially reportable under the provisions of 10CFR50.55(e).
The'event involves seismic design of the WNP-1/4 General Services Buildings.
A key assumption used in the original analysis considered all floors as rigid extensions of the exterior walls.
l<e have subsequently developed an analysis which allows the floors to flex in direct proportion to the force applied and in inverse proportion to stiffness.
Preliminary evaluation indicates there may be a potential for a change in the resultant stresses on the sub-structure.
Our AE is in -The process of performing a detailed computer analysis to determine if redesign of the substructure is required.
W.
G. Albert of your staff was verbally informed of this during his inspection of WNP-1 activities the week of October 18, 1976.
The Supply System is investigating the problem and will inform you of the results of their review.
Very truly yours, OLR:OHW:vh 0
L'cHBERGER AssistanT. Director, Generation 8 Technology
WcIshington Public Power Svpply System A JOINT OPERATING AGENCY P. 0. 8OX Qdd 3000 CCO. Wxavi~oTOrc WAY RICHLANO, WASHINO'tON QQ353 PHONC(50Ql Q4d ldll December 17, 1976 G01-76-767 Nuclear Regulatory Commission Region V-
-. Suite 202, Walnut Creek Plaza 1990 North California Boulevard Malnui Creek, CalifoTnia 94596 Attention:
tlr. R. H..Engelken, Director
Subject:
HPPSS NUCLEAR PROJECTS NOS.
1 AHD 4 REPORTABLE EVFHT - 10CFR50.55(e)
DOCKET HOS.'0-460 AHD 50-513
Reference:
G01-76-694,
- 0. L. Renberger, MPPSS, to R.
H. Engelken,
- NRC, same subject, dated November 5, 1976.
Dear Hr. Engelken:
You were informed=of a potentially reportable design deficiency in the reference 'ietter.
Subsequently.,
or, Hovemb"r 19, 1976, your staff wa.'.
inFormed by telephone that the Supply System had determined that this potential deficiency YIas reportable under the'rovision of 10CFR50.55(e) and,that there were two aspects to the problem.
Attachments to this letter are reports from our Architect-Engineer which summarize the problems and describe in a summary-form the corrective actions taken.
You will note that two problems are addressed separately to facilitate your review:
a)
The first aspect. is related to seismic design oF the General
, Services Building (GSB) for the.'horizontal earthquake, and analysis
. assumptions for floors.
,This problem is the one which we reported as affecting the substructure aIid horizontal shear loads in GSB interior walls.
b)
The second aspect of the problem is related to the vertical earth-quake and assumptions used for analysis of floor slabs.
The integrity of the structure itsel f vIas not at issue, but rather the indirect effects of seismic forces acting on equipment within the bui lding.
Mr.
R.
H. Engelken Page 2
We have reviewed our Architect-Engineer's narrative report.
Members of our technical staff have audited the analysis methods and techniques and evaluated the conclusions.
The Supply System finds the resolution of these deficiencies to be acceptable.
If you require any additional information, please feel free to contact us ~
OLR:MEW:km Attachments Yery truly yours, O.
L.
REHBERGER Assistant Director Generation
& Technology cc:
JR Schmieder - UE&C GE Thornes -
UE&C CR Bryant - BPA JB Knotts -
C&K
iVQShIngtOA PubfIC POs~i T $Upp)y $ySte~
A lOINI O? RATING rsGciVCY
/Was
~4
- p. O. Btsx gag 3Q9() tasse.
'Vlas'rewrrtsra WAY Rim swis. WasaslraiiN'a 993S2 PHo February 24, 1977 G01-77-133 t(ucl ear Regulatory Corf:mission Region Y
Suite 202, Ha 1 nu t Creek Pl aza 1990 Worth California Boulevard walnut Creek,California 94596 ij Attention:
Mr. R. H. Engelken, Director
Subject:
MPPSS NUCLEAR PROJECTS i'(OS.
1-ANO 4
+ REPORTABLE EYENT - 10CFR50.55(e)
DOCKET HOS. 50-460 ANO 50-5't3
Reference:
- Letter, R. H. Engelken to O. L. Renberger, same subject, dated January 19, 1977 Oear Mr. Engelken:
4 ihe rererenued 1etder requested
...at additionai'questians be responded to by the Supply System.
Our response is as follcws and is for,.iulated in accordance with the two items identified in the original report.
question:
(assumed'.distribu~jon.of Lateral Seismic Force~s
"!Ihat were the causes of the design deficiencies...."
Response
The original design of the General Service Building assumed that the floor slabs acted as rigid diaphragms and delivered horizontal seismic loads to the vertical walls in proportion to the shear stiffness of the walls.
During the design of the floor slabs in the superstructure of the General Service Building, it was found that the slabs themselves would not transfer these horizontal forces in a manner to support the original assumptions.
t)uclear Regulatory Commission Page 2
The validity of ihe assumpiion of rigid diaphragms was reevaluated as a
result of the supersiructure design process.
The supersiruciure design process was not, completed prior to release of ihe substructure drawings.
The redesign of the substructure was done subsequent to its original release as a result of the normal superstructure design review orocess.
Question:
"Did they represent a breakdown'in the design review process...."
Response
The seis"ic analysis incident involving horizontal shear distribution through the General Service Building siructure developed as a result of ihe designers investigation of ihe validity of analy'sis
- methods, techniques and assumptions whicp have ccmnonly been accepted within the industry, and prev',ously used f'r the design of nuclear fa'cilities.
For multi-siory struciures items corinonly assumed that-concrete floor slabs act as rigid diaphragms in transmitting horizontal earthquake forces to the structure walls and foundation.
As a result of the designers review and investigation'of this generally accepted basic analysis assumption, it was determined that, because of the uniqueness of the Mi')P-I/O General Service Building. design, the orig anal assumpt ion'was not valid, cr th'.s particular"structure.
It was determined thai more refined analysis methods and techniques should. be used io account for flexibilityof the floor slabs in the horizontal direction.
Consequently, it was decided that a three:dimensional finiie element analysis should be performed.
~ ~
The designer's evaluation of the application of finite'lement theory to seismic analysis of the General Service Building det'er'mined that a much better definition of the structural response to earthquake motion could be obtained than with the methods of seismic analysis previously employed.
i'lo breakdown in the design review process at. either UE8C or WPPSS occurred but rather a natural design evaluation took place employing more sophisticated methods and assumptions.
Question:
"'Ahat corrective action is planned~"
Huclear Pegulatory Cormission Page 3
Resoonse:
Ho correctiv act~on is required since this design was chanoed as a r suit of the designer s efforts to investigate the applicability of co, only accepted assumptions whicn have previously been used as a basis for analysis of nuclear,acilities.
Such review of assumptions by senior AE personnel are routinely conducted as a part of the design review process.
This was
'and is an ongoing process, and no corrective action is necessary.
Question:
(Yertical component of earthquake forces)
"What were the causes of the design deficiencies...?"
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~Re ense:
~
- The causes of the design changes were ihe use of assumptions in the original design which were Ia'ter found to be non-conservative."
Reevaluation of the design assumptions..<gas brought about..as a result of reviev of other S~"s issued subsequent to the design of the MNP-1/4 General Service Building.
Question:
"Oid they represent a breakdown in the design review process...?"
Resoonse:
~
~
Th'e seismic analysis incident involving the vertical response of floor slabs within the GeneralService Building developed as a result of the designer's review of other saTety analysis reports.
These 'documents indicated that floor slabs could have anplified respgnse in the.verPica1 direci:ion as opposed to the commonry used assumption that floor slabs respond the same as the supporting walls.
Further evaluation of this phenomenon revealed that lexibilityof the General Service Building floor slabs would allow some amplification in the vertical direction.
tfo breakdown in the design review process occurred either at UE8C or at WPPSS.
On the contrary, the fact that improved anal@'is techniques r(ere evaluated and implemented demonstrates that. the review process is working as intended.'
Question.
"What corrective action is planned?"
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~uclear Reaulatory Comnissian s,
pge4 ective action is p'lensed as the design was changed as a result ot the designer's actions to review other PSAR's which have been filed with ne i{Re and subsequent efforts to evaluate the impact of the application o~ more refined analysis methods on the General Service Building design.
Such review of assumptions by senior AE personnel are routinely conducted as a part of the design review process.
This was and is an ongoing process, and no corrective action is nec ssary.
pf you require any additional infoirration, please feel free to contact us Very truly yours;
- 0. L. REilBERGER Assistant Director Generation Il Technology pl p pE'A'DLS DH'A cc:
JR Schnieder, UE&C GE Thor nes, iJKg CR Sryant, SPA
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,'lay 3G, 1978 l!uclear Pegulatory Cormission Region Suite 202, Walnut Creek Plaza 190" N. California Boulevard
!,'a!nut Creek, '.ali;ornia 9-".596 AttenitiGn:
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':lPPSS NUCLEAR PROJECT ll0.
2 20Cl(ET
~l'Ji'1BER 0 397 i CPPR 93 REPORTiABLE EVEi')T - 1v'CFR50.55(e)
Dear I'r. Engelken:
Your s aff.vas informed of a reportable deficiency under the pro-vis-':ons of 10CFR50.55(e) by telephone on April 24, 1978.
Attached is our report on this probl'em.
Corrective action is being taken as described in the report.
If.'u have any ouestiors concern'g this ma tter, ol ease contact Us.
Very truly yours, co<~
D. L. REHBERGER Assistant Director Generation and Technology DLR:EAF:cph Attachment cc:
JJ Byrnes, BSR.
D. Roe, BPA
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PEPGRTABLE DE" ICIEt<CY:"~:
'<<RECTI'JE ACTIOte'PEi'lI!'tiG I i'ORiQDO liISSILE BARRIER I"( DIESEL GE!tERATOR BUILDIfiG
!1ASH'GTO!'1 PUBLIC PO!/FR SUPPLY SYSTEi~1 DOCKET.'10.
50-397 LIC llSE NO.
CFl R-93 Oes~cci tion ot the De'iciency At tne three exterior doors at ground level at the south exterior i'all of the Diesel Generator Building, reinforced concrete L-shaoed interior walls are orovided to protect the diesel generators from tornado-propelled missi'as.
These L-shaped walls vere designed
=.nd constructed, hc:.v ver, so tha<<hey do not completely block the d or opening.
A "window", or I-~neo;-sight gap aporoximately 3 inches wide by 7 feet high exists through which an externally-genel ated missi'.a could oass.
Safet Im lications The L-shaped walls do not'ully block the door openings in the exterior walls of the Diesel Generator Building.
Consequently, a tornado-propelled missile having a low trajectory in a direction
=pproximat ly 60 degrees east of !cnorth, could penetrate the exterior door and not be interceoted by the L-shaped missile barrier wall.
This could result in danage to one of the diesel generator units, i;hich, vnen combined with a single active failure of one 'o the other diesel generator units under emergency conditions, could pre-clude safe shutdown of the reactor.
n Corrective Action A structural element will be provided-at each of the three door ooenings; attached to the inside face of the exterior wall, which wiII be capable of intercepting and preventing penetr'ation by design basis tornado-generated missiles.
These structural elements will consist of steel plate enclosures anchored to the wall, and filled with plain concrete.
These will extend vertically from the floor to above the door opening,
'Nashington Public Power Supp(y Systl-iw A JOINT OPERATING AGENCY P. 0, 8OX 988 3OOO CCO. WASHINCTOH WAY RICHLANO. WASHINCTON 99353 PHONC(509) 940 ~ IC'll G02-78-150 May 26, 1978 Nuclear Regulatory Commission Region V
Suite 202 Walnut Creek Plaza 1900 t). California Boulevard tlalnut Creek, California 94596 Attention:
Mr. R.
H. Engelken, Director SIibject:
ltPPSS NUCLEAR PROJECT NO.
2 DOCKET NUMBER 50-397, CPPR-93 5
REPORTABLE DEFICIENCY - 10CFR50.55 e
Dear Mr. Engelken:
In accordance with the provisions of 10CFR50.55(e),
your staff was informed by telephone on April 24, of a reportable deficiency in-volving control room air handling unit leakage which could result in the possible over exposure of Control Room Operators to radiation following a LOCA.
Attached is our report on this deficiency.
Please contact us if you have additional questions.
Very truly yours, DLR:HLB:cph attachment D. L.
RENBERGER Assistant Director Generation and Technology cc:
JJ Byrnes, BER JJ Verderber, B8R D. Roe, BPA
PEPORT.";BLE DEFICI'ENCY AND CORRECTIVE ACTION WPPSS NUCLEAR PROJECT NO, 2
POSSIBLE OVEREXPOSURE OF CONTROL ROOM OPERATORS TO RADIATION FOLLOWING A LOCA WASHINGTON PUBLIC POWER SUPPLY SYSTEM DOCKET NO. 50-397 LICENSE NO.
CPPR-93 Des cription of Deficienc As currently designed, the air handling unit blower discharges directly from the air handling unit into the control room air supply ducting.
This arrangement results in pressures inside the air handling unit case which are lower than the pressure in the surrounding HVAC equipment rooms and promotes the possible in-leakage of unfiltered air into air handling units.
The leakage criteria contained in the control room air handling unit speci-fication was based on the guide lines given in Regulatory Guide 1.52, "Design, Testing and Maintenance Criteria for Atmosphere Cleanup Systems Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Plants" Revision 0, June, 1973.
The leakage rate indicated was 1" of rated flow for recirculating pri-mary atmosphere cleanup housings and ductwork.
Certified leak test results for the two control room air handling units were 62 and 70 cfm, respectively,.
which is approximately 0.03%%d of rated flow.
The possibility for airborne radioactive contaminants to exist within the HVAC equipment rooms results from the fact that HVAC equipment room heating, venti-lating and airconditioning is provided by critical switch gear room HVAC equip-ment.
The air supply for the critical switch gear room HVAC equipment is from the inlet plenum area which also serves the control room HVAC systems during normal operation.
This plenum inlet may be exposed to radioactive releases following accident conditions.
To reduce the thyroid dose, which is indicated as the most critical, will require limiting air handling unit in-leakage to at least 0.02%%d of total air handling unit flow..
Safet Im lication Unless the reportable deficiency is corrected, the infiltratio'n of airborne radio-active contaminants into the control room following a LOCA may possibly result in a 30-day integrated radiation dose to Control Room Operators which exceeds speci-fied limits.
Corrective Action Following identification of the reportable deficiency a number of possible modifications to the control room HVAC system and air handling unit were in-vestigated to reduce or eliminate the air handling unit in-leakage.
These modifications included (1) isolating the control room HVAC equipment in sep-arate leak tight rooms, (2) enclosing the air handling unit in an additional pressurized
- case, and (3) replacing the existing air handling unit blower with a separate blower at the air handling unit inlet.
Replacement of the air hand'ling unit blower with a separate blower unit which will blow through the ai'r handling unit has been selected as the course of action to be followed in correcting the design deficiency.
This modification provides for pressurizing the air handling unit casing and ducting thereby eliminating in-leakage of air downstream of the blower unit.
In addition to the above indicated changes, the planned modifications will require (1) revision of outside air intake ductwork and automatic damper loca-tions, (2) revision of ductwork from the emergency filter units to the air handling unit blower, (3) sealing the present return air openings through the floor slab and bottom of the air handling unit mixed air/filter section and (4).
relocating the present 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire door to a new return air slab opening in the new fan inlet plenum.
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I Mfashinaton Public Power Supply System A 101NT 0?ERATING AGENCY P. 0. SOX 988 3000 GCO. WiSH<aOTON Wir RICH<aOO. WASNiHOTOre SSSSa PHO~C (809) 945.908l G02-7-5 December 27, 1977 Nuclear Regul-atory Commission Suite 202, Malnut Creek Plaza 1990 N. California Boulevard Walnut Creek, California 94596 Attention:
f'tr.
P..
H. Engelken, Director
Subject:
llPPSS NUCLEAR PROJECT HO.
2 D CKET NUYiBERS 50-397.
PP -0
~ REPORTABLE EVENT - 10CFR50.55(e)
Dear tlr. Engelken:
Your. staff was informed of a reportable deficiency under the provisions of 10CFP50.55(e) by telecon on December 2, 1977.
Attached is our interim reoort on this problem.
Hetallurgical studies are still underway to determine the factors which caused the deficiency to occur.
!)e anticipate our welding consultant's report wi'il be completed and additional information regarding the cause of this deficiency will be transmitted to you within 30 days.
Should you have any additional questions, please contact us.
Very truly yours, D. L.
RENBERGER Assistant Director Generation and Technology Attachment DLR:RJS:df cc:
JJ Byrnes, BER (NY)
JD Hilson, BHR (Site)
AliACHHEHT gl.
R oortable Deficiency arid Correctlv Action llPPSS iluclear Project Ho.
2 Cracks in Structural Steel.Platform at 54I' Elevation Inside Containmeni Washington Publ ic Pov'er Supply System Docket tto. 50-397 License Ho.
CPPP,-93 INTERIH REPORT
~ Descriotion of the Deficienc Insp ction results have id ntifiea cracks in field welds made to heavy structural beams which makeup the 541'levation platform inside the containment drywell.
This platform spans between the sacrificial hield and the containment vessel and is primarily designed to carry pipe vihip restraint loads for large diameter and high eneray piping systems including Hain Steam and Reactor Feedwater lines.
In addition, this platform is used for supporting secondary loads including air handling equipment, recirculating pump motors, pipe hangers and,snubber supports as v'ell as structural supports for other miscellan=ous equipment in that vicinity.
Th cracks w re identified as a result of ext a
inspection requirements imposed to resolve a noriconforming condition relat d to welding procedures.
This was further expanded to include a
baseline magnetic particle examination of all field >>elds which had been made on this platform structure.
Of the G3 fie1d weld locations.
20 were acc pted, 7 were identified as major cracks, 39 were rejected as minor surface indications and 17 were inaccessable because of. equipment or other temporary obstructions.
Cause and Anal sis Several contributing factors are considered as havinig caused th k
in the S4 crac's 1
elevation platform including deficient welding and fabrication sequence procedures, incomplete inprocess inspection and possible metallurgical conditions associated vlith hydrogen embrittlement and nil ductility transition temperatures.
This matter is currently being analyzed by our metallurgical consultant.
Detailed information is pending completion of our consultant's report.
This information should be available within 30 days.
e Safet Im lication The structural mePiibe s
in which cracks were identified, are desiqned to carry the primary loads associated with the pipe whip restraints of high energy piping systems including !bin Steam and Peactor Feedwater lines.
Under pipe break loads, which. constitutes a failed load condition, the members are designed for plastic deformation.
The cracked welds could result in a local failure within the structure,
- thereby, reducing the effectiveness of a pipe whip restraint.
Corrective'ction Taken and Planned In response to the initial identification of cracks in the 541'latform, additional magnetic particle inspection was performed on all the 'field welds made to that platform structure.
In addition, the Architect Engineer/Construction l'lanager performed a visual inspection of other structural field welds made by the, Contractor inside containment.
\\
A stringent repair program has been implemented and is being directed by our Architect/Engineer.
Th Suooly System's welding consultant concurs with this repair orogram.
To assure adherance to the repair procedures, the Architect/Engineer is performing second party inspection along with specific inprocess inspection hold points requiring A/E acceptance.
The Supp'ly System is evaluating additional corrective actions affecting welding processes and inspection criteria which may be utilized to prevent recurrence of'hese conditions.
Any corrective actions will be bas-d on this review and the review made by our consultant.
The results of this review and any additional actions planned or taken will be included in our final report.
Washington Public Power Supoly System, A JOINT OPERATING AGENCY P. 0. SOX 968 3000 GCO. WASHINOSON WAY RlOHL.ANO. WiSH<NOTON SSSSa PHONe <509) 948.988l AJI i1 10 19 8
G02-78-121 Nuclear Regulatory Commission Region V
Suite 202, Walnut Creek Plaza 1900 N. California Boulevard Walnut Creek, California 94596
. Attention:
Hr. R.
H. Engleken, Director
Subject:
WPPSS NUCLEAR PROJECT NO.
2 DOCKET NUMBER 50-397, CPPR-93 g REPORTABLE EVENT - 10CFR50.55(e)
Reference:
Letter, G02-77-504, same subject, dated 12/27/77
Dear ter. Engleken:
The reference letter transmitted our interim report describing a reportable deficiency under the provisions of 10CFR50.55(e) concerning cracks in the structural steel platform at 541'levation inside containment.
Attached is our final report on this matter.
Should you have any additional questions, please contact us.
Very'ruly yours, D. L.
RENBERGER Assistant Director Generation and Technology Attachment DLR:RJS:df cc:
JJ Byrnes, S&R (New York)
ATTACHMENT Reportable Deficiency and Corrective Action WPPSS Nuclear Project, No.
2 Cracks in Structural Steel Platform at 541'levation Inside Containment Washington Public Power Supply System Docket No.
50-397'icense No.
CPPR-93 FINAL REPORT Descri tion of the Deficienc Inspection results have identified cracks in field welds made to heavy structural beams which makeup the 541'levation platform inside the
'ontainment drywell.
This platform spans between the sacrificial shield and the containment vessel and is primarily designed to carry pipe whip restraint loads for large diameter and high energy piping
'ystems including Hain Steam and Reactor Feedwater lines.
In addition, this platform is used for supporting secondary loads including air handling equipment, recirculating pump motors, pipe hangers and snubber supports as well as structural supports for other miscellaneous equipment in that vicinity.
The cracks were identified as a result of extra inspection requirements imposed to resolve a nonconforming condition related to welding procedures.
Subsequently, magnetic particle examination has been expanded to include all field welds which affect the integrity of the pipe whip support structures throughout containment and in the main steam tunnel.
PresentIy over 60 field welds on the 541'levation platform have been.repaired.
As repair efforts at this elevation near completion, inspections and necessary repairs will commence at other pipe whi p support locations.
Cause and Anal sis Several contributing factors are considered as having caused the cracks in the 541'levation platform including deficient welding and fabrication sequence procedures, incomplete inprocess inspection and possible metallurgical conditions associated with hydrogen embrittlement and nil ductility transition temperatures.
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~,
Our consultant concluded that the most likely explanation for the major cracks is that they initiated as hydrogen induced cold cracks in the HAZ regions following weld cooldown.
These hydrogen cracks were in turn enhanced by the weld shrinkage stresses associated with the restrained structural configurations and reduced toughness pro~erties of the materials which were exposed to ambient temperatures below 32 F.
A5 Safet Im lication e
The structural
- members, in which cracks were identified, are designed to carry the primary loads associated with the pipe whip restraints of high energy piping systems including Hain Steam and Reactor Feedwater lines.
Under pipe break loads, which constitutes a failed load condition, the members are designed for plastic deformation.
The cracked welds could result in a local failure within the structure,
- thereby, reducing the effectiveness of'a pipe whip restraint.
Corrective Action Taken and Planned In response to the initial identification of cracks in the 541
- platform, additional magnetic particle inspection was performed on all the field welds made to that platform structure.
In addition, magnetic particle examination is being performed on all field welds which are critical to the assembly or function of the pipe whip support structures within the dyrwell or in the main steam tunnel.
Those field welds not yet completed shall be subject to magnetic particle examination of root pass, mid-point and final weld-out.
All welds (incluhng repairs) which require magnetic particle examination, shall be inspected 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after cooldown.
Upon completion of repairs and cooldown of the 541'levation platform, it shall be v'isually inspected to assure its acceptance.
A stringent repair program has been implemented and is being directed by HPPSS/Architect/Engineer.
Included in this program are directions for sequencing weld repairs in such a manner as to minimize weld shrinkage stresses and associated distortion.
The Supply System and our welding consultant concurs with this repair program.
To assure adherance to the repair procedures, the Architect/Engineer is performing second party inspection along with specific inprocess inspection hold points requiring A/F acceptance by an A'MS gC I certified welding inspector.
All further structural welding will be performed in accordance with sequence procedures prepared by the. Contractor and approved by the Architect/
Engineer.
To assure the pipe whip support members remain well above the nil-ductility transition temperature duping cold weather, the minimum drywell temperature shall be maintained at 70 F for the duration of construction activities.
Additionally, review of the fracture toughness of the pipe whip support structures at the anticipated operating temperatures and design loads is now in progress.
i~1any of the support structures h'ave been fabricated from A516GR70 and A537CL2 materials with designated impact properties.
- However, a sampling of those supports which were fabricated from conventional A-36 materials are being further evaluated by llPPSS'onsultants to determine their ability to resist brittle fracture at operating conditions.
Should the results of this review indicate that a reportable condition
- exist, the HRC shall be so advised.
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P. o. sox 988 3000 ceo. wsswrrororr wig Rrcrrr.ANo, wAsrrrrrororr 99352 prrorrc (509) 375-5000 fg-7 November 6, 1978 Nuclear Regulatory Commission Region V
Suite 202 Malnut Creek Plaza 1900 N. California Boulevard llalnut Creek, California 94596 Attention:
Hr. R. H. Engelken, Director
Subject:
NPPSS NUCLEAR PROJECT NO.
2 DOCKET NUMBER 50-397, CPPR-93 REPORTABLE DEFICIENCY - 10CFR50.55(e
Dear >tr. Engelken:
In accordance with the provisions of 10CFR50.55(e),
your staff was informed by telephone on Octoberl0, 1978, of a reportable deficiency involving Anchor/Darling tilting disc check valves in the Residual Heat Removal System failing to close with gravity when installed in a vertical position which could potentially result in damage to the system or a delay in system response.
This problem had been previously identified in IE Circular No. 78-15.
Attached is our report on this deficiency.
If you require additional information, please feel free to contact us
~
Very truly yours, DLR:JAO:cph Attachment D. L.
RENBERGER Assistant Director Technology cc:
JJ Verderber, B&R RC Root, BER Site JJ Byrnes, 88R D. Roe, BPA E. Volgenau, NRC, !tashington, D.C.
REPORTED DEFICIENCY AND CORRECTIVE ACTION FOR TILTING DISC CHECK VALVES FAILURE TO CLOSE
'AITH GRAVITY IN VERTICAL POSITION Nature of Deficienc Anchor/Darling informed Burns and Roe by letter, dated June 18, 1978, that their tilting disc check valves of a specific pressure class and size may not close by gravity alone once they are fully opened if mounted in a vertical pipe.
The cause is that the disc center of gravity travels beyond or is directly above the disc pivot point when the valve is fully open.
This problem has been identified in IE Circular No. 78-15.
)'e have identified all these valves in safety-related systems, i.e. the Residual Heat Removal (RHR) pump discharge check valves (RHR-V-31A, B
and C) and determined that RHR-V-31B will not close by gravity when fully opened.
Safet Im lications Failure of the pump discharge check valve to close will drain the RHR/Low Pressure Coolant Injection (LPCI) discharge piping into the suppression chamber.
The water leg pump will not be able to maintain the piping full due to the flow path to the suppression chamber.
Subsequent start of the RHR pump may resu'lt in water hammer which could disable the RHR/LPCI Loop.
Also, the time required for LPCI to inject water into the reactor may be increased beyond the time assumed in the accident
- analyses, due to the extra time required to refill the discharge piping.
An additional single failure would reduce the Emergency Core Cooling Systems below minimum requirements.
I Corrective Action Taken and Planned Anchor/Darling has been contacted by MPPSS about modifying the disc by adding a weld buildup or a lug to the disc counterweight so as not to allow the disc center of gravity to travel over the disc pivot point.
The modification will be coordinated with Anchor/Darling and it is expected to be completed by'arch, 1979.
The field Quality Assurance program will inspect and verify after the fix that this valve will close by gravity when fully opened.