ML17258B064
| ML17258B064 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 05/11/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17258B063 | List: |
| References | |
| NUDOCS 8105290484 | |
| Download: ML17258B064 (7) | |
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION A51END ENT NO.
42 TO PROVISIONAL OPERATING LICENSE NO.
DPR-18 ROCHESTER GAS AND ELECTRIC CORPORATION R.
E.
GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 INTRODUCTION By letter dated November 13, 1980, Rochester Gas and Electric Corporation (the licensee)
'proposed changes to the Technical Specifications
{TSs) appended to Provisional Operating License No.
DPR-18 for the R.
E. Ginna Nuclear Power Plant.
The changes involve the incorporation of certain of the TMI-2 Lessons Learned Category "A" requirements.
The licensee's request is in direct response to the NRC staff's letter dated July 2, 1980.
BACKGROUND INFORMATION By our letter dated September 13, 1979, we issued to all operating nuclear power plants requirements established as a result of our review of the TMI-2 accident.
Certain of these requi rements, designated Lessons Learned Category "A" requirements, were to have been completed by the licensee prior to any operation subsequent to January 1, 1980.
Our evaluation of the licensee's compliance with these Category "A", items was attached to our letter to Rochester Gas and Electric Corporation dated July,7', 1980.
In order to provide reasonable assurance that operating reactor facilities are maintained withinthe:limits determined acceptable following the implementation of the TMI-2 Lessons Learned Category "A" items, we requested that licensees amend their TS to incorporate additional Limiting Conditions of Operation and Surveillance Rhquirements, as appropriate.
This request was transmitted to all licensees on July 2, 1980.
Included therein were model specifications that we had determined to be acceptable.
The licensee's application is in direct response to our request.
Each of the issues identified by the NRC staff and the licensee's response is discussed in the Evaluation below.
EVALUATION 2.1.1.
Emer enc Power Su 1
Re uirements The pressurizer water level indicators, pressurizer relief and block valves, and pressurizer heaters are important in a post-accident situation.
Adequate emergency power supplies add assurance of post-accident functioning of these components.
The licensee has the requisite emergency power supplies.
The power-operated relief valves receive DC control power from the bus supplied by the battery chargers and thus will receive power from the 8 X@5 R'90 /It'/
emergency battery in the event of an emergency.
The pressurizer water level indication, block valves, and pressurizer heaters all receive normal power from safeguards bodes which can be provided power from the onsite emergency diesel generators.
Because of this and the fact that
'the diesel generators are tested in accordance with existing Technical Specifications, the licensee saw no need to propose additional specifi-cations.
We concur and conclude that the emergency power supplies pro-vide reasonable assurance of post-accident functioning of the subject components and are thus acceptable.
2.1.3.a Direct Indication of Valve Position The licensee has provided a direct indication of power-operated relief valve (PORV) and safety valve position in the control room.
These indications are a diagnostic aid for the plant operator and provide no automatic action.
The licensee has provided TSs with a 31-day channel
- check, an 18-month channel test requirement for the PORV indicator and an 18-month channel calibration requirement for the safety valve indicator.
Although the channel test was not included as an option in the model specifications, we consider, it an acceptable alternative to the channel calibration and we conclude that the specifications are acceptable.
2.1.3.b Instrumentation for Inade uate Core Coolin The licensee has installed an instrument system to detect the effects of low reactor coolant level and inadequate core cooling.
These instruments, subcooling meters, receive and process data from existing plant instrumentation.
We previously reviewed this system in our Safety Evaluation dated July 7, 1980.
The licensee submitted TSs with a 31-day channel check and an 18-month channel calibration requirement and actions to be taken in the event of component inoperability.
We conclude the TSs are acceptable as they meet our,July 2, 1980 model TS criteria.
- However, we note that there is continuing discussion regarding the accuracy and reliability of reactor vessel water level instrumentation.
Rochester Gas and Electric Corporation has declined to install such instrumentation until suitable equipment has been developed.
This subject is considered to remain an open item.
2.1.4 Diverse Containment Isolation The licensee has modified the containment isolation system so that diverse parameters will be sensed to ensure automatic isolation of non-essential systems under postulated accident conditions.
These parameters are high containment pressure and low pressurizer pressure.
We have reviewed this system in our Lessons Learned Category "A" Safety Evaluation dated July 7, 1980 The modification is such that it does not result in the auto-matic loss of containment isolation after the containment isolation signal's reset.
Reopening of containment isolation would require deliberate operator action.
The TSs submitted by the licensee list each affected containment ilolation valve and provide for the appropriate surveillance
and actions in the event of component inoperability; therefore, we conclude that the TSs are acceptable.
We note, however, that specific reference to the testing of valves prior to their return to servi ce after maintenance, replacement, or repair is not included in th'e Technical Specifications.
This require-ment is, however, included in Appendix C of the Ginna guality Assurance Manual.
We consider this acceptable 2.1.7.a Auto Initiation of Auxiliar Feedwater S stems The licensee has provided for the automatic initiation of auxiliary (emergency) feedwater flow on loss of normal feedwater flow.
The auto-initiation signals used by the licensee include low-low steam generator water level and loss of 4kv voltage
We have previously reviewed the design and install ation of this system as part of our Lessons Learned Category "A" program.
The circuits are designed to be testable and the design retains the capability of manual actuation from the control room even in the event of failure of the auto-initiating c1rcuitry.
The TSs submitted by the licensee list the appropriate components, describe the tests and provide for proper test frequency.
The TSs contain appropriate actions in the event of component inoperability; therefore, we conclude that the TSs are acceptable.
2.1.7.b Auxiliar Emer enc Feedwater Flow Indication The licensee has installed auxiliary (emergency) feedwater flow indication that meets our testability and vital power requirements.
We reviewed this system in our Safety Evaluation dated July 7, 1980.
The licensee has proposed a
TS with. 31-day channel check and 18-month channel calibration requirements.
We find this TS acceptable as it meets the criteria of our July 2, 1980 model TS criteria.
2.2.1.b Shift Technical Advisor STA Our request indicated. that the TSs related to minimum shift manning should be revised to reflect the augmentation of an STA.
The licensee's application would add one STA to each shift to perform the function of accident assessment.
The individual performing this function will have a bachelor' degree or equivalent in a scientific or engineering discipline or will be a qualified but non-degreed Senior Reactor Operator.
We have found this satisfactory for the short-term, although in the long-term, incumbents in this position must have bachelor's degrees Part. of the STA duties are related to the operating experience review function, and Rochester Gas and Electric Corporation has added the position of Technical Assistant for Operational Assessment, to whom the STA will report in this regard.
We have found that these additions will enhance the safety of plant operation and are thus acceptable.
OTHER SPECIFIC COMMENTS REGARDING TECHNICAL SPECIFICATION CHANGES:
In our review;;of the proposed specifications, we determined that additional items not specifically mentioned above deserve comment.
These are:
Specification 2.3.1.3 has been modified to incorporate pressurizer level measurement requirements stated in IE Bulletin 79-'21.
We have found this acceptable.
(2)
(3)
(4)
(5)
Specification 3.5.3 regarding Engineered Safety Feature Actuation Instrumentation, does not incorporate response time surveillance as included in the model TS.
The licensee does not have Standard Technical Specifications and the Ginna Technical Specifications do not include response times except as otherwise incorporated in this action for the AFW and Containment Isolation Systems.
We have concluded that the lack of these requirements is satisfactory.
Although the required operator actions for items 3.c and 3.e of Table 3.5-.2 do not conform to the model TS, we consider them satisfactory.
The model TS imply that the charm'hl is totally inoperable, whereas Rochester Gas and Electric takes into account the fact that the channel can be "tripped," thus requiring only a signal from the operable channel to scram the plant.
Although the model TS for item 3.e of Table 3.5-2 would require the start signal for the AFW pumps at any time when power is greater than 0%,
RGSE has determined that the actual value should be SX.
This is because at less than 5X power, AFW flow is being controlled to maintain steam generator level (main feed water is not in service) and automatic initiation of full AFW flow would be deleterious.
We concur in this assessment and conclude that the modification is acceptable.
As a point of clarification, containment purge and vent valves are included in the Containment Ventilation Isolation unit of Table
- 3. 5-3.
RG8E has not included channel functional tests for the PORVs because of the potential for serious error during such testing at the Ginna plant.
It must first be recognized that the safe position of these valves is shut; RG&E maintains that channel functional testing of the valve is therefore unnecessary.
- Also, in order to perform the test, the operator must manually control the pressurizer.
the control bank of heaters must be removed from
- service, and manual control of pressurizer spray is necessary.
We concur that the benefit to be gained from such an evolution does not outweigh the risk involved and thus that the exception is justified and acceptable.
(7)
Items 4.c and 4.d of Technical Specification Table i.5-5 aie not included in the surveillance requirements of Technical Specification Table 4.1-1.
RGSE has agreed to submit, by March 16, 1981, applicable changes to the specifications
In the interim, the surveillance is being performed as required by plant procedures.
We find this acceptable.
(8)
RGSE has determined there are no bypass functions to be included in Specification 4.4.6
( Containment Isolation Response) and 4.8.9 (AFW initiation).
(9)
Changes to Figure 6.2-2 include those made to reflect the current Ginna Station organization and include the addition of, an Office Supervisor, a Materials coordinator, and the plant stockroom.
None of these changes are expected to have an impact on the safe operation of the plant and we therefore find them acceptable.
(10)
Paragraph 4.3.3.3, transmitted originally by NRC Order dated April 20, 1981, has been modified to correct a typographical
'mission of a qualifying phrase.
EVALUATION TO SUPPORT LICENSE CONDITIONS 2.1.4 Inte rit of S stems Outside Containment Our letter dated July 2, 1980, indicated'hat. the license should be, amended by adding a license condition related to a Systems Integri'ty Measurements Program.
Such a condition would require the licensee to effect an appropriate program to eliminate or prevent the release of significant amounts of radioactivity to the environment via leakage from engineered safety systems and auxiliary systems, which are located outside reactor containment.
By letter dated November 13, 1980, the licensee agreed to adopt such a license condition; accordingly we have included this condition in the license.
2.1.8.c Iodine Monitorin Our letter dated July 2, 1980, indicated that the license should be amended by'adding a license condition related to iodine monitoring.
Such a condition would require the licensee to effect a program which would ensure the capability to determine the airborne iodine concentration in areas requi ring personnel access under accident conditions.
By letter dated November 13, 1980,"
the licensee agreed to adopt such a license condition; accordingly, we have included this condition in the license.
IV.
ENVIRONMENTAL CONS I OERATION We have determined that the amendment does not authorize a change in effluent types or total amounts nor an in'crease in power level and will not result in any significant environmental impact.
Haveing made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact state..ent, or negative declaration and environmental impact appraisal need not be prepared in connection wi+4 the issuance o
he amendment.
V.
CON CLOS ION Me have concluded, based on the considerations discussed
- above, that,'1) because this amendment does not involve a significant fnc&aso in the probability or consequences of accidents.prev'iously 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 safegy..of the..publ.ic will not be endangered by operation in the proposed
- manner, and (3) such activities will be conducted in compliance with the Commission'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:
May 11, 1981
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