ML19209A268
| ML19209A268 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 09/19/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19209A259 | List: |
| References | |
| NUDOCS 7910030253 | |
| Download: ML19209A268 (23) | |
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3 UNITED STATES E'
'g NUCLEAR REGULATORY COMMISSION 3,,7 4
E WASHINGTON, D. C. 20555 S
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 64 TO PROVISIONhl OPERATING LICENSE NO. DPR-21 NORTHEAST NUCLEAR ENERGY COMPANY, ET AL MILLSTONE NUCLEAR POWER STATION, UNIT N0. 1 DOCKET NO. 50-245 INTRODUCTION By application dated May 27, 1977, which supersedes application dated February 28, 1977, the Northeast Nuclear Energy Company (the licensee) requested changes tn the Technical Specifications appended to Provisional Operating License No. DPR-21 for-the Millstone Nuclear Power Station, Unit No.1.
Periodic updating information for the valve and pump testing program was submitted by letter dated February 28, 1979. The proposed changes would replace the current inservice inspection and testing technical specifications w th an inservice inspection and testing program i
that meets the requirements of 10 ;FR 50.55a.
I.
Class I Components A.
Reactor Vessel 1.
Request relief from volumetric examination of i5e welds listed below.
These welds are it ated in the reactor vessel beltline region, vessel shell, and lower head.
(Items Bl.1 and Bl.2, Examination Categories B A and B-8, respectively, from Table IWB-2600.)
VCBA-1 BHBB-3 VLBA-1 BHBB-4 VLBA-2 BHBB-5 VLBA-3 BHBB-6 VLAA-1 BHAB-1 VLAA-2 BHAB-2 VLAA-3 BHAB-3 VCBB-1 BHAB-4 VAAA-1 BHAB-5 BHCB-1 BHAB-6 BHCB-2 BHAB-7 BHBB-1 BHAB-8 BHBB-2 Code Requirement Volumetric examination of the shell longitudinal and circumferential welds.
Examination shall cover at least 10% of the length of each longitudinal weld and 5% of the length of each circumferential weld during the inspection interval. Examinations may be performed at or near the end of the inspection interval.
t 91003 2.63 o
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2-Basis for Requesting Relief The reactor vessel is insulated with permanent reflective insulation and surrounded by the concrete biological shield.
The annular space between the inside diameter of the biological shield and the outside diameter of the insulation is a nominal 6-1/2 inches.
Thus, access for removal of the insulation panels is extremely limited and this inaccessibility precludes direct examination of these welds from the outsiue surface. The interior surface of the reactor vessel is stainless steel clad and the vessel's internals, shroud, and jet pumps would make an internal volumetric examination of most of these welds impractical.
Evaluation Imposition of the Code requirements would subject the licensee to extreme hardships in necessitating removal of portions of the concrete biological shield and the permanently installed insulation to perform the required examination of the welds listed from the vessel outside surface.
The reactor vessel is presently monitored for raoiation damage in the beltline region.
This vessel surveillance program i-. in compliance with ASTM E185-66 and has been evaluated with respect to the require-ments of ASTM E185-73.
We have determined that the surveillance program meets the essential requirements of ASTM E185-73 and therefore conforms to the intent of 10 CFR Part 50, Appendix H.
This program will provide sufficient data to monitor radiation damage to the vessel beltline materials throughout the vessel's service life.
In addition, the vessel was designed and fabricated in accordance with the rules of Section III of the 1965 Edition of the ASME Boiler and Pressure Vessel Code. We have evaluated the vassel's fracture tough-ness properties and find that they meet the principle requirements set forth in 10 CFR Part 50, Appendix G.
Utilizing the results of the surveillance program to monitor material damage from neutron irradiation and the guidelines in Regulatory Guide 1.99 to establish operating limitations will insure that the reactor vessel will be operated in accordance with 10 CFR Part du, Appendix G requirements.
This provides acceptable margins of safety to prevent brittle fracture of the vessel during any conditions of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which it may be subjected during the remaining service life.
Other methods of volumetric examination with the exist *ng limitations which will produce meaningful results have not been fully developed at this time.
However, acoustic emission, which is considered to be in the developmental stage, is being studied by the licensee as an examination technique for these welds. The licensee has committed to adopt acoustic emission and include it in the inservice inspection 1087 195
. program for strveillance of these welds when a system is demonstrated to be practical and endorsed by the NRC.
The licensee has also proposed to visually inspect the areas of these welds from the reactor vessel inside surface to the extent practical utilizing a remote television camera during the inspections required by Examination Categories B-N-1 and B-N-2.
Other vessel welds of the same examination category which are higher stressed than those listed and thus more susceptible to inservice flaw growth will be examined in accordance with the requirements of Section XI.
The licensee must follow the augmented program outlined below when examining these welds:
(a) Examine volumetrically at least 100% of accessible length of each longitudinal weld and at least 100% of the accessible length of each circumferential weld, from either inside or outside the vessel.
(b) Visually inspect to the extent practical, and from the vessel inside surface, the areas of the welds required to be examined.
(c) In the event that a Code unacceptable flaw is detected, 100%
volumetric examination shall be performed on the welds listed.
We conclude that the vessel design, ongoing surveillance program of the reactor vessel materials in the beltline region, and the augmented examination requirements are adequate for providing an acceptable level of safety and assurance that the vessel structural integrity will not be compromised during the inspection period.
2.
Request relief from examination of integrally-welded vessel supports. (Item Bl.12, Examination Category B-H, from Table IWB-2600.)
Code Requirement Volumetric examination of at least 10% of the weld to the vessel during each inspection interval.
Basis for Requesting Relief Access to the reactor vessel support skirt weld is impeded by the reactor vessel insulation structural support steel and the concrete biological shield.
The area is one of high radiation levels (2R field at a distance of six to seven feet from the weld) which would endanger the health of examination personnel if access were available.
1087 196 e
4 Remote examination techniques have not been developed to acccmplish the examination requirements.
Evaluation Access to this weld is limited.
However, because of the structural importance of this weld, the licensee must try to surf ace examine at least 10% of the weld during this inspection period.
Surface examination will be the most practical examination technique considering the limitations and radiation and will provide atteptable results in determining the structural integrity of this weld.
3.
Request relief from volumetric examination of control rod drive housings per IWB-1220(b)(1).
(Item B1.18, Examination Category B-0, from Table IWB-2600.)
Code Requirement Volumetric examination of welds in 10% of the peripheral control rod drive housings unless the component is exempt under one of the cendi-tions listed in IWB-1220.
Basis for Requesting Relief The component is exempt from volumetric examination under IWB-1220(b)(1).
Evaluation The licensee has shown that under the postulated condition of a loss of coolant from the control rod drive housings the reactor can De shut down and cooled down in an orderly manner by the reactor coolant makeup system.
Thus, the Code requirements have been satisfied.
B.
Piping Pressure Boundary 1.
Request to substitute surface examination for portions of the welds listed below.
(Item B4.5, Examination Category B-J. )
CCBJ-2 FWAJ-13 CCBJ-3 FWAJ-28 FWBJ-21 FWAJ-29 FWBJ-24 MSDJ-7 FWBD-1 MSDJ-11 FWBD-2 MSCJ-12 FWBJ-31 MSCJ-14 FWAJ-12 MSCD-1 MSBD-1 SCAJ-PB-1 1087 197
. RVBJ-l ICAJ-l RVAJ-1 ICAJ-4 MSAJ-10 ICAJ-5 MSAJ-11 ICAJ-6 CSAJ-1 ICAJ-11 CSAJ-2 ICAJ-12 CSAJ-6 ICAJ-14 CSAJ-7 ICAJ-15 CSAJ-9 ICBJ-3 CSAJ-10 ICBJ-4 CSAJ-11 ICBJ-12 CSAJ-13 ICBJ-14 CSAJ-16 CUAJ-PB-1 CSBJ-3 RMAJ-RRB CSBJ-6 RMAJ-3 CSBJ-7 RCAJ-PB-1 CSBJ-9 RCBJ-PB-2 CSBJ-10 RCBD-1 CSBJ-11 RCBJ-1 CSBJ-12 RCBJ-26 CSBJ-14 RRJF-1 CSBJ-16 RRFD-1 SCAJ-CU-1 RRGD-1 SCAJ-2 RRJD-1 SCAJ-3 RRBJ-3 SCAJ-4' RRCJ-3 CROJ-2 RREJ-3 CRDJ-3 RRFJ-3 CRDJ-4 RRDJ-4 CRDJ-5 RRAJ-4 CROJ-6 CRDJ-9 CRDJ-7 CRDJ-11 CRDJ-8 CROJ-12 Code Requirement Volumetric examination of all of the area of 25% of the circumferential joints including the adjoining one-foot sections of longitudinal joints and 25% of the pipe branch connection joints during the inspec-tion interval.
Basis for Requesting Relief These welds are pipe-to-cast valve body joints, pipe-to-fitting joints, or welds obstructed by hangers or other pipes.
Volumetric examination of the entire weld cannot be performed because of U/T incompatibilities with the cast materials, fittings, valves or weld geometry, or because of the inaccessibility of the weld to volumetric examination equipment.
1087 198
. Evaluation Practical alternative techniques for volumetrically examining the entire areas of the welds listed which will produce meaningful results are nnt presently available.
The licensee has proposed to use surface examination on areas of the welds which cannot be volumetrically examined.
The combination of ultrasonic and surface examinations is judged adequate in providing a level of safety and assurance of the piping boundary integrity.
2.
Request to substitute surface examination for the required volumetric examination of integrally-welded pipe supports.
(Item B4.9, Examina-tion Category B-K-1, f rom Table IWB-2600. )
Code Requirement Volumetric examination of 25% of the integrally-welded supports during each inspection interval.
Basis for Requesting Relief These welds are either partial penetration or fillet welds by design.
Volumetric examination of these types of welds by the ultrasonic testing method would not produce meaningful results.
Evaluation Radiographic examination of these welds would be difficult to perform and interpret, costly, and burdensome for the licensee with little added assurance of safety. The licensee has committed to subject these welds to surface examination and to volumetrically examine the base metal.
Based on the loading conditions of these types of welds, flaws would most likely generate at the weld surface and thus be detectable by surface examination. Ultrasonic examination of the base metal would provide assurance that flaws in the base metal do not exist.
The examination techniques to be employed by the licensee are therefore considered acceptable in providing assurance that the pipe supports integrity will be maintained during the inspection period.
C.
Valve Pressure Bo,undary 1.
Request relier from visual examination of the internal surfaces of the valves listed.
(Item B6.7, Examination Category B-M-2.)
1-FW-llA 1-FW-11B l-RR-1 A 1-RR-2A
~.
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. 1-RR-1B l-RR-2B l-RR-4A 1-RR-48 Code Requirement One valve in each group of valves of the same constructional design, e.g., globe, gate or check valve, manufacturing method and manufacturer that performs similar functions in the system shall be visually examined internally during each inspection interval.
Basis for Requesting Relief These valves are located in piping which penetrates the reactor pressure vessel and cannot be isolated for disassembly and visual examination.
To accomplish the required examination woulu entail drainage of the reactor vessel as well as removal of the core.
Evaluation These valves are subjected to the volumetric examination requirements of Category B-M-1 and the system leakage and hydrostatic pressure tests required by IWA-5000. Other valves in lines which penetrate the reactor pressure vessel experience similar thermal-hydraulic conditions.
These valves were fabricated from the same materials as the valves listed and will be inspected internally during this inspec-tion period.
In addition, the licensee is required to inspect the internal surfaces of the valves listed if unacceptable conditions exist in those valves inspected or when drainage of the reactcr vessel is necessary for other purposes.
It is judged that the condi-tion of the valves inspected will be representative of the condition of the valves listed and that these examinations will provide assurance that the valves structural integrity will be acceptable for safe operation during this inspection period.
II.
Class 2 Components A.
Pressure Vessels 1.
Request to substitute surface examination for the required volumetric examination of the LPCI heat exchangers circumferential shell welds.
(Item C1.1, Examination Category C-A.)
Code Requirement Volumetric examination of at least 20% of each circumferential weld, uniformly distributed among three areas around the vessel circumference over the service lifetime of the component.
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e Basis for Requesting Relief The gecmetry of the circumferential shell welds is such that volumetric examination of these welds utilizing ultrasonic examination techniques cannot be performed adequately to produce meaningful results.
Egaluation Because of the LPCI heat exchangers design, radiography is also an impractical niethod for performing the required volumetric examination.
The licensee has proposed to use surface examination in lieu of volumetric examination. We require that surface examination of at least 20% of each circumferential shell weld be performed during this inspection period and that the examinations be performed on the portion of the welds in the proximity of the nozzles.
Because of the heat exchanger design and the weld configuration, we have determined that surface examination of the welds in these areas is acceptable for assuring the structural integrity of the heat exchangers.
2.
Request relief from volumetric examination of the LPCI heat exchangers nozzle velds, CCA-C-B-1, CCB-C-B-1, CCA-C-B-2 and CCB-C-B-2.
(Item C1.2, Examination Category C-B.)
Code Requirement Volumetric examination of 100% of the nozzle-to-vessel attachment weld over the service lifetime of the component.
Basis for Requesting Relief These welds are covered by reinforcing collars and cannot be examined by ultrasonic techniques.
Evaluation Based on drawings and details of these welds supplied by the licensee, the collar-to-nozzle and nozzle-to-shell welds are integral welds and therefore are nozzle-to-shell welds.
The licensee has proposed surface examination of these welds. We require that ultrasonic examination of these welds be utilized to the extent practical supple-mented by the surface examination. This combinatio of examination techniques is judged acceptable for flaw detection in the concerned area.
3.
Request relief from examination of the isolation condenser nozzle welds, 1C4C-B-1, ICAC-B-2, ILBC-B-1 and ICBC-B-2.
(Item C1.2, Exami-nation Category C-B.)
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-9 Code Requirement Volumetric examination of 100% of the nozzle-to-vessel attachment weld over the service lifetime of the component.
Basis for Requesting Relief These welds are covered by encapsulation sleeves which were installed as a result of the high energy pipe break study.
Evaluation The licensee is presently committed to an augmented inspection program for the isolation condenser. Under this progra n, the condenser shell-side temperature and water level are continuously monitored, eddy-current examinat:on of the tubes is performed every second scheduled refueling outage, and a primary side pressure test is Surface examinations must be perfortned performed annually.
on these welds from the nozzles I.D. in conjunction with the scheduled eddy-current examinations. With the augmented inspection programs, it is considered that adequate measures are being taken to provide assurance that failure of these welds is unlikely to occur during the inspection period and that a flaw will be detected early enough in its developmental stage to prevent a significant safety hazard.
B.
Piping 1.
Request relief from examination of the jsolation enndenwr system piping longitudinal weld joints in fitting and branch connection welds.
(Items C2.2 and C2.3, Examination Category C-F.)
Code Requirement Volumetric examination of 100% of the weld over the service lifetime.
Basis for Requesting Relief Access to these welds is prevented by high energy pipe break encapsu-lation sleeves.
Evaluation The licensee is presently committed to an augmented inspection program for the isolation condenser under which these welds are subjected to a pressure test annually.
During the pressure test, the areas of these welds will be visually checked for leakage.
This augmented inspection is considered adequate as an interim measure for assuring the integrity of the welds.
v 1087 202
. 2.
Request to substitute surface examination for volumetric examination of those welds in the main steam piping and branch connections which cannot be examined volumetrically.
(Items C2.1 and C2.3, Examination Category C-F.)
Code Requirement Volumetric examination of 100% of the weld.
Basis for Requesting Relief The piping system is designed such that some volumetric examinations are limited due to geometric configuration or accessibility. These limitations generally exist at pipe-to-fitting or valve-to-fitting welds where volumetric examination can be performed from the pipe side only.
Evaluation The licensee has committed to examine these welds volumetrically to the extent practical and to use surface examination on the portion of the weld which cannot be tested adequately by ultrasonics.
Since the steam line will Le subjected to the system pressure test in addition to the examinations to which the licensee is comitted, the combination of these examinations will be sufficiently reliable to detect flaws and provide assurance that the steam line's. structural integrity will be maintained during the inspection period.
III. System Pressure Tests 1.
Request relief from testing portions of the systems listed at the code required test pressure:
System Class 1 Boundary Class 2 Boundary Feedwater 1-FW-llA/B HP Heater Discharge Isolation Valves Standby Liquid Control 1-SL-8 1 - S L-6 Core Spray 1-CS-6A/B l-CS-5A/B LPCI l-LP-11A/B l-LP-10A/B Reactor Cleanup Return 1-CU-29 CRD Return 301-98 Stop Upstream of 301-95 Code Requirement
.The pressure retaining components shall be subjected to a hydrostatic test at 1.10 times the system operating pressure at least once toward the end of each inspection interval.
1087 203
. Basis for Requesting Relief
.\\
The location of check valves in several systems that penetrate the primary containment preclude the Class 1 pressure test boundary from extending outward beyond the first of such valves, usually located inside containment even though the class change boundary is outside containment.
In these cases, the Class I leakage and pressure test boundary would be the inside check valve.
Conversely, pressure tests of Class 2 systems, which are outside containment, would have to be bounded at a stop valve which may or may not be the Class 1/ Class 2 boundsry.
. Evaluation Because of the design of those systems listed, isolation of Class 1 and Class 2 systems at the boundary cannot be accomplished.
To prevent overpressurization of the Class 1 portions of thase systems (1,080 psig for Class 1 ve-sus 2,375 psig for Class 2), the portions of these systems which cannot be tested to the Code required pressure must be visually inspected during system operation to the extent practicable. These portions of the piping systems are examined volumetrically under Category B-J.
The examinations and alternate testing will be acceptable in assuring the systems' integrity for safe operation of the facility during this inspection period.
2.
Request to hydrostatically test the service water, fuel pool cooling and the reactor building closed cooling water systems at pressures which are 'ower than the Code required as shown below:
Class 3 Syst g Test Pressure Code Test Pressure Service Water 45 psig 165 psig Fuel Pool Cooling 190 psig 220 psig R8CCW 80 psig 165 psig Code Requirement The pressure retaining components shall be subjected to a hydrostatic test at 1.10 times the system design pressure at least once toward the end of each inspection interval.
Basis for Requesting Relief These systems are required for continuous cooling of vital system components and cannot be removed from service for the period of time required for the hydrostatic test.
1087 204
. Evaluation The licensee has proposed testing these systems for leakage by utiliz-s*
ing the pump head and throttled flow for continuous cooling of the The pressures at which the systems will be vital system components.
tested are greater than normal operating pressure and are considered adequate for providing assurance that they are structurally acceptable.
Request to test the isolation condenser system at 1,375 psig instead 3.
of the Code required 1,563 psig.
Code Requirement The pressure retaining components shall be subjected to a hydrostatic test 1.25 times the system design pressure at least once toward the end of each inspection interval.
Basis for Requesting Relief To be compatible with current surveillance testing requirements.
Evaluation This requirement does not impose an undue burden or hardship on the licensee nor cause an unsafe condition to the facility. The licensee must satisfy the Code requirement and change the current surveillance testing requirements to be compatible with those of the Code.
4.
Request relief to IWA-2120(c) regarding the Authorized Nuclear Inspector's witnessing or audittag test results.
Code Requirement IWA-2120(a):
"It is the duty of the Inspector to witness or otherwise verify all examinations and pressure tests required by this division for Class 1 components and for Class 2 components where required.
The Inspector shall also make any additional investigations necess6ry to verify that all applicable requirements have been met."
IWA-2120(d):
"It is the duty of the Inspector to assure himself that the examinations 4.d tests required for Class 3 components and systems (IWD-1000) have br conducted and the results recorded."
Basis for Requesting Relief The licensee and Nort' ast Utilities Service Company Quality Assurance Personnel will be witnessing these tests or auditing the test results.
108/
205
. ~
Evaluatinn IWA-2130(b) states specifically that the Insp?ctor shall not be an employee of the Owner or his agent. The licensee must meet the requirement of the Code concerning the Authorized Nuclear Inspector.
IV.
Testing of the Pumps 1.
Request relief from measurement of bearing temperatures of the LPCI, core spray, condensate, service water, and emergency service water pumps.
Code Requirement Searing temperatures shall be measured during at least once inservice test each year.
Basis for Requesting Relief These pumps are vertical design pumps with bearings located in the motor driver and pump casing.
The bearings are inaccessible for temperature mers trements and the pumps would require extensive and costly modific.Vi ns to accomplish the requirement.
Evaluation The licensee has provided detailed drawings of these pumps which show that a possible alternate measurement, the external cooling water temperatures, which could be used as an indicator of the bearing temperature trend cannot be performed practicably because of the cooling water circuit design.
Vibration amplitude measurements are taken and recorded monthly.
Because of the frequency of measurement of this parameter and the Code requirement to compare this parameter to reference values, the vibration amplitude measurement is a suitable indicator of bearing degradation and the degradation will be realized sooner by vibration amplitude measurenents taken monthly than by yearly bearing temperature measurements.
2.
Request relief from the required flow measurenent of the reactor feed, condensate, and condensate booster pumps.
Code Requirement Monthly measurement of flow rate.
1087 206
. Basis for Requesting Relief These pumps do not have flow instrumentation installed for individual
('-
pump flow measurement. They supply water to common headers where instrumentation is set up to measure header or system flow rates.
Evaluation The licensee has committed to measure individual pump motor current, system flow rate, and pressure to detect hydraulic changes in these The measurements of these parameters are adequate in detecting pumps.
changes of the hydraulic characteristics of these pumps and thus satisfies the intent of the Code.
3.
Request to test the emergency condensate transfer and the emergency service water pumps quarterly instead of monthly.
Code Requirement An inservice test shall be run on each pump, nominally each month during normal plant operation.
Basis for Requesting Relief Plant surveillance procedure 625.1 requires the emergency condensate transfer pump M-728 to be tested quarterly.
Evaluation The licensee has not shown that compliance with the Code requirement is impractical for the facility.
Because of the added assurance of the operational readiness of these pumps by monthly testing, we conclude that the Code requirement must be satisfied.
4.
Request to test the reactor shutdown cooling pumps during shutdown instead of monthly.
Code Requirement An inservice test shall be run on each pump, nominally each month during normal plant operation.
Basis for Requesting Relief These pumps are used only during plant shutdown and cannot be operated unless the suction pressure is above four psig and the reactor water temperature is below 300 F.
1087 207
m
. Evaluation Because of the system design and operation, these pumps can only be tested during shutdown. The system has redundant capability and we have detennined that the frequency of testing is sufficient to assure the safe operation of the reactor shutdown systems. We have, therefore, granted the request.
5.
Request relief to IWA-2120(c) regarding the Authorized Nuclnr Inspector's surveillance of operating tests.
Code Requirement IWA-2120(c):
"It is the duty of the Inspector to assure himself that the inservice tests required on pumps and valves (IWP and IWV) have been completed and the results recorded."
Basis for Requesting Relief The licensee interprets this section of the Code to mean that the Authorized Nuclear Inspector may elect to witness any or all opera-tional readiness tests and the preparations thereof.
Because operating tests may be conducted at any hour, seven days a week, and the Author-ized Nuclear Inspector has no training or experience in plant opera-tions, the licensee requests a waiver from the requirement.
Evaluation Tests, under normal conditions, which are performed by a facility are scheduled and are not haphazardly done. The intent of IWA-1220(c) is not to interrupt the scheduling of a test but to assure the Authorized Inspector that tee required tests on pumps have been completed and the results recorded.
If the Authorized Inspector wishes to observe the performance of a test, he can be notified well in advance as to the date and time scheduled for such test. This requirement does not place an undue burden or nardship on the licensee. Therefore, this requirement must be satisfied.
1087 208
. V.
Valve Testing Program A. General 1.
Licensee Documents and Piping and Instrumentation Drawings By letter dated February 28, 1977, the licensee submitted proposed changes to the Millstone Unit 1 Technical Specifications to incorporate the pro-visions of 10 CFR 50.55(a) as revised on February 12, 1976 (41 FR6256). The licensee's letter dated February 28, 1979, submitted updated infomation related to pump and valve testing. Our review of the valve testing portion of the proposed test program was limited to an evaluation of those safety related valves associated with the safety systems defined in the "NRC Staff Guidance for Preparing Pump and Valve Testing Program" sent to the licensee on January 5, 1078. Throughout our review, staff questions regardirig the adequacy of the program were resolved with the licensee.
Fomal responses to these questions together with commitments to modifications to the proposea To test program were documented by the licensee in a revision to the plan.
conduct our review we used the licensee boundary diagrams submitted by letter dated February 28, 1979.
2.
Leak Testing of Valves which Perform a Pressure Isolation Function There are sevt ral afety systems connected to the reactor coolant pressure N undary i.nat have design pressures that are below the reactor.~r k t system operating pressure. There are redundant isolation valves forming the interface between these high and low pressure systems to prevent the low pressure systems from being subjected to pressures which exceed their design limits. In this role, the valves are perfoming a pressure isolation function.
It is our view that the isolation redundancy provided by these valves regarding their pressure isolation function is important.
We consider it necessary to provide assurance that the '.ondition of each of these valves is adequate to maintain this redundant isolation and system integrity. For this reason, we conclude that some methods, such as leak testing, should be used to assure that their
- condition is sufficient to maintain this pressure isolation function.
1087 209
. 1 In the event that leak testing is selected as the appropriate method for achieving this objective, we believe that the following valves should be categorized as A or AC and leak tested in accordance with IWV-3420 of Section XI of the applicable edition of the ASME Code. These valves are:
1-LP-ll A 1-CS-5A 1-LP-llB l-CS-5B 1-LP-10A 1-50-5 1-LP-10B l-50-1 1-CS-6A 1-SD-2A 1-CS-6B l-50-2B We have discussed this matter and identified the valves listed above with the licensee. The licensee has agreed to review leak testing these valves in accordance with IRV-3420 of the applicable edition of the ASME Code and to categorize these valves with the If the licensee determines that leak appropriate designation.
testing is not necessary because there are other methods that they have and will use to determine each valve's condition, they shall provide to the NRC for evaluation, on a valve by valve basis, the details of the method used that clearly demonstrates the condition of each valve.
Subsection IWV-3410(a) of the Section XI Code (which discusses full 3.
stroke and partial stroke requirements) requires that Code Category A and B valves be exercised once every three months, with exceptions as defined in IWV-3410(b)(1), (e) and (f).
IWV-3520(a) (which dis-cusses full stroke and partial stroke requirements) requires that Code Category C valves be exercised once every three months, with exceptions as defined in IWV-3520(b). In the above cases of exceptions, the Code permits the valves to be tested at cold shutdown where:
(a) It is not practical to exercise the valves to the position required to fLifill their function or to the partial position during power operation.
(b) It is not practical to observe the operation of the valves (with fail-safe actuators) upon loss of actuator power.
The staff stated its position to the licensee that check valves, whose 'afety function is to open, are expected to be full-stroked.
If only limited operation is possible (and it has been demonstrated by the licensee and agreed to by the staff), the check valve shall be partial stroked. Since disk position is not always observable, the NRC staff concluded that verification of the plant's safety analvsic design flow rate through the check valve would be an adequate demon-stration of the full-stroke requirement. Any flow rate less than design will be considered part-stroke exercising unless it can be 1087 210
=
. shown that the check valve's disk position at the lower flow rate would be equivalent to or greater than that of the design flow rate through the valve. The licensee agreed to conduct his flow tests to satisfy the above position.
The licensee has stated that none of the Category A or B power operated valves listed below can be part-stroked because of the design logic of the operating circuits. These circuits are such that when an open or close signal is received the valve must complete a full stroke before the relay is released to allow the valve to stroke in the other direction. We find that the licensee's relief request from part-stroking is warranted and should be granted because the required function of the valves involves only full open or full closed positions.
4.
Inservice valve testing at cold shutdown is acceptable when the following conditions are met. The licensee is to commence testing as soon t-the cold shutdown condition is achieved but not later than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after shutdown and continue until canplete or plant is ready to return to power. Completion of all valve testing is not a prerequisite to return to power. Any testing not canpleted at one cold shutdown shall be performed during any subsequent cold shutdowns that may occur before refueling to meet the Code specified testing frequency.
For planned cold shutdowns, where the licensee will complete all the valves identified in his IST program for testing in the cold shutdown mode, exceptions to the above conditions may be taken.
It is noted that the staff differentiates for valve testing purposes between the cold shutdown mode and the refueling mode. That is, for testing purposes the refueling mode is not considered as a planned cold shutdown.
5.
The Code states that, in the case of cold shutdowns, valve testing need not be performed more often than once every three months for Category A and B valves and once every nine months for Category C valves.
It is our position that the code is inconsistent and that Category C valves shall be tested on the same schedule as Cateaorv A and B valves. The licensee has agreed to modify any procedures as necessary on cold shutdown to read, "In the case of frequent cold shutdowns, valve testing will not be performed more often than once every three (3) months for Category A, B and C valves."
6.
Changes to the Technical Specification In our November 1976 letter to the Northeast Nuclear Eneroy Co., we provided an attachment entitled "NRC Guidelines for Excluding Exercising (Cycling) Tests of Certain Valves During Plant Oparation.'
The attachnent stated that when one train of a redundant system such as in the Emergency Core Cooling System (ECCS) is inoperable, 1087 211
- -i
. i
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nonredundant valves in the remaining train should not be cycled
...o since their failure would cause a loss of total system function.
For example, during power operation in some plants, there are
,gs stated minimum requirements for systems which make up the ECCS which allow certain limiting conditions for operation to exist at any one time and if the system is not restored to meet the require-ments within the time period specified in a plant's Technical Specifications, the reactor is required to be put in some other Furthermore, prior to initiating repairs all valves and mode.
interlocks in the system that provide a duplicate function are required to be tested to demonstrate operability imediately and For such plants periodically thereafter during power operation.
this situation would be contrary to the NRC guideline as stated in the document mentioned above.
The licensee has agreed to review the Millstone 1 Technical Specifications and to consider the need to propose Technical Specification chsnges which would have the effect of precluding such testing.
If, af ter completing this review, the licensee determines that the Technical Specifications should not be changed because the guidelines are not applicable or cannot be followed, the licensee will submit to the NRC the reasons that led to their determination for each poten-tially affected valve. The licensee's submittal shall identify the potentially affected sections of the Technical Specifications, and the valves in question.
7.
Safety Related Valves Our review was limited to those Class 1, 2 and 3 valves of Section XI of the ASME Code that wer-safety related.
Safety related valves are defined as those that ars e.wied to mitigate the consequences of an accident and/or to shut down the reactor and to maintain the reactor in a shutdown condition.
It should be noted that the licensee may have included nonsafety r21ated valves in their Inservice Test Program as a decision on the licensee's part to expand the scope of their Inservice Test Program.
8.
Testing of Valves Request relief to IWA-2120(c) regarding the Authorized Nuclear Inspector's surveillance of operating tests.
Code Requirement IWA-2120(c):
"It is the duty of the Inspector to assure himself that the inservice tests required on pumps and valves (IWP and IW) have been completed and the results recorded."
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. Basis for Requestina Relief The licensee interprets this section of the Code to mean that the Authorized Nuclear Inspector may elect to witness any or all opera-tional readiness tests and the preparations thereof.
Because operating tests may be conducted at any hour, seven days a week, and the Author-ized Nuclear Inspector has no training or experience in plant opera-tions, the licensee requests a waiver from the requirement.
Evaluation Testing of some valves during normal plant operations cannot be performed because of safety considerations.
These valves are being required to be tested during other modes of plant operation and strict scheduling of tests for these valves cannot be maintained by the licensee.
In such cases, the licensee must request relief from the requirement of having an Authorized Nuclear Inspector present to observe the performance of tia tests. However, the licensee is not relieved from recording test results and retaining this information for inspection by the AJthorized Nuclear Inspector.
B.
Core Spray and Low Pressure Coolant Injection System i
1.
Category C Valves
(
1.1 Relief Requested I
Exclude the following check valves from the ASME Code Section XI requirements for quarterly operability test.
1-CS-6A and 1-CS-6B (Pump discharge to vessel checks) 1-LP-11A (Inboard check in the LPCI system)
ASME Code Requirement Refer to paragraph V. A.3.
Licensee Basis for Requested Relief There is no design provision for manually exercising these valves and stroking with system flow requires that water be pumped into the reactor vessel. This is not possible at power, because of pressure differences and thermal-hydraulic considerations.
NRC Evaluation Valves 1-LP-11 A,1-CS-6A and 6B cannot be exericised without de-pressurizing the reactor because the LPCI and core spray ;ystem pumps are not able to overcome reactor coolant pressure.
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=
. Cold shutdown testing is impracticable because of the difficulty in controlling reactor water level. The LPCI pump output is 5000 gal / min. This would cause the reactor water level to change at a rate of 25 inches / minute. The core spray pump output is 3600 gal / min which would mean an 181nch/ min change in reactor water level. With these rates it is difficult to maintain the water level within the cold shutdown operating brad. For these reasons and for the reasons set forth below, we find that full stroke exercising at refueling outages is adequate and that the relief should be granted.
It is noted that these valves are considered passive and redundant.
Passive as used here means any component whose unavailability upon demand is less than or equal to 10-4/ demand. Redundant as used here means the existence of more than one valve for perfoming a given function.
The Category C valves listed above are either passive and/or redundant valves. The optimum test interval for operability testing passive and/or redundant valves was determined by the staff using actual valve failure rate data and standard probabilistic techniques, to be in the range of 3 to 27 months. Refueling intervals, which have been proposed as the exercise interval for the valves above occur every 12 to 18 months which is within the optimum range for operability testing of these valves.
Furthermore, the ASME Code, which requires testing be done quarterly and which has been adopted in 10 CFR 50.55a, also allows testing at cold shutdowns if quarterly testing is impractical.
Cold shutdowns can occur at intervals up to refueling outages. Therefore, changing the test interval from quarterly to refueling does not differ signifi-cantly from the Code pemitted change from quarterly to cold shutdown testing.
Based on the considerations discussed above the staff concludes that the alternate testing frequencies proposed above would give the reasonable assurance of valve operability intended by the Code and that the relief thus granted will not endanger life or property or the common defense and security.
C.
Standby Liquid Control System 1.0 Category A or AC Valves 1.1 Relief Requested Category AC valves 1-SL-7 and 1-SL-8 (combined discharge outside and inside drywell respectively) from the three month ASME Section XI Code exercising requirement and instead exercise at refueling outages.
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s
. Licensee Bases for Requesting Relief Refer to paragraph V. A.3.
Licensee Basis for Requested Relief There is no provision for manual operation of these valves. Hydraulic stroking requires pumping into the reactor vessel and the operation of a squib valve. There are no test connections at present by which an individual valve leak rate test could be perfomed.
NRC Evaluation We agree with the licensee basis for not stroking these valves during power operatior. (quarterly) or during cold shutdowns. For this reason and for the reasons set forth below, we find that full stroking exer-cising at refueling outagas is adequate and that the requested relief should be granted.
It is noted that these valves are considered passive and redundant.
Passive as used here means any compgnent whose unavailability upon demand is less than or equal to 10- / demand. Redundant as used here means the existence of more than one valve for performing a given function.
The Category AC valves listed above are either passive and/or re-7 dundant valves. The optimum test interval for operability testing N passive and/or redundant valves was detemined by the staff using actual valve failure rate data and standard probabilistic techniques,,
to be in the range of 3 months to 27 months. Refueling intervals, which have been proposed as the exercise interval for the valves listed above occur every 12 to 18 months which is within the optimum range for operability testing of these valves.
Fur:!.ennore, the ASME Code, which requires that testing be done quarterly and which has been adopted in 10 CFR 50.55a, also allows testing at cold shutdowns if quarterly testing is impractical. Cold shutdowns can occur at intervals up to refueling outages. Therefore, changing the test interval from quarterly to refueling does rot differ signifi-cantly from the Code permitted change from quarterly to cold shutdown testing.
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e VI.
Sumary - Inservice Inspection and Testina We find that the revised Inservice Inspection and Testing Program for the Millstone Nuclear Power Station Unit 1 meets the requirements of 10 CFR 50.55a(g). Therefore, the proposed changes are acceptable.
We have granted relief from specific ASME Section XI Code requirements for this inspection period which ends December 28, 1980.
Based on the foregoing, we find the relief requested is authorized by law, will not endanger life or property or the comon defense and security and is in the public interest considering the burden on the licensee that could result if the relief were not granted.
ENVIRONMENTAL CONSIDERATION We have determined that this amendment and granting of the relief do not authorize a change in effluent types or total amounts nor an increase in power iuel and will not result in any sinnificant environ-mental impact. Having made this determination, we have further concluded that the amendment and relief involve actions which are insignificant from the standpoint of environmental impact and, pursuant to 10 CFR s51.5(d)(4), that an environmental impact statement, or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these actions.
CONCLUSION We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manne; and (3) such activities will be conducted in compliance with the Cc. mission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date:
Septenber 19, 1979 1087 216